WorldWideScience

Sample records for piping system components

  1. Reliability of piping system components. Volume 4: The pipe failure event database

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R; Erixon, S [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B [RSA Technologies, Visat, CA (United States)

    1996-07-01

    Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A `data driven and systems oriented` analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs.

  2. Reliability of piping system components. Volume 4: The pipe failure event database

    International Nuclear Information System (INIS)

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

    1996-07-01

    Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A 'data driven and systems oriented' analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-01

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

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

    International Nuclear Information System (INIS)

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

    1995-12-01

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

  5. Seismic Design of ITER Component Cooling Water System-1 Piping

    Science.gov (United States)

    Singh, Aditya P.; Jadhav, Mahesh; Sharma, Lalit K.; Gupta, Dinesh K.; Patel, Nirav; Ranjan, Rakesh; Gohil, Guman; Patel, Hiren; Dangi, Jinendra; Kumar, Mohit; Kumar, A. G. A.

    2017-04-01

    The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event.

  6. Removal of Shippingport Station primary system components and piping

    International Nuclear Information System (INIS)

    LaGuardia, T.S.; Lipsett, S.M.

    1987-01-01

    The dismantling workscope for the Shippingport Station Decommissioning Project was divided into subtasks to permit the work to be subcontracted to the maximum extent practicable. Major subtasks were identified and described by Activity specifications which could then be grouped into logical work packages to be put out for bid. Two of the largest dismantling work packages, removal of piping and components, were grouped together and designated as Activity Specifications 4 and 5. TLG Services, Inc. and Cleveland Wrecking Company formed a Joint Venture to perform this work during a two-year period at a cost of approximately $7 million. The major portions of this dismantling workscope are described. The primary system components within this workscope consist of the stainless steel reactor coolant piping, check valves, reactor coolant pumps, steam generators, and reactor purification demineralizers and coolers. The work performed, the heavy rigging preparations and procedures, the cutting tools used, component draining/capping techniques to prevent spills, contamination containment, airborne control techniques, and lessons learned during the removal of these primary system components are described. Summaries of crew size and composition, labor hours, duration hours and radiation exposure to workers are provided and discussed briefly. The successful completion of this work is evidence of the engineering, planning, equipment, materials and labor pool available to remove large, radioactively contaminated components safely. This experience will help decommissioning planners to prepare for the removal of reactor components in future decommissioning

  7. Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

    International Nuclear Information System (INIS)

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1987-01-01

    Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

  8. Applications of the TVO piping and component analysis and monitoring system (PAMS)

    Energy Technology Data Exchange (ETDEWEB)

    Smeekes, P. (Teollisuuden Voima Oy, Olkiluoto (Finland)); Kuuluvainen, O. (Rostedt Oy, Luvia (Finland)); Torkkeli, E. (FEMdata Oy, Haukilahti (Finland))

    2010-05-15

    To make fitness, safety and lifetime related assessments for piping and components, the amount of data to be managed is getting larger and larger. At the same time it is essential that the data is reliable, up-to-date, well traceable and easy and fast to obtain. At present the main focus of PAMS is still on piping, but in the future the component related databases and applications will be more and more developed. This paper presents a piping and component database system, consisting of separate geometrical, material, loading, result and document databases as well as current and future applications of the system. By means of a user configurable interface program the user can generate indata files, run application programs and define what data to write back into the result database. The data in the result database can subsequently be used in new input files to perform postprocessing on previous results, for instance fatigue analysis. crack growth analysis or RI-ISI. The system is intended to facilitate the analyses of piping and components and generate well-documented appendices comprising significant parts of the input and output and the associated source references. (orig.)

  9. Experiments and calculations to leak openings and leak rates on typical piping components and systems

    International Nuclear Information System (INIS)

    Hoefler, A.; Grebner, H.

    1992-01-01

    Calculations of leak opening and leak rate for through cracks in piping components have been performed. The analyses are pre- or mostly post-calculations to experiments performed at the HDR facility under PWR operating conditions. Piping components under consideration are small diameter straight pipes with circumferential cracks, pipe bends with longitudinal or circumferential cracks and pipe branches with weldment cracks. The component are loaded by internal pressure and opening as well as closing bending moment. The finite element method and two-phase flow leak rate programs are used for the calculations. Results of the analyses are presented as J-integral values, crack opening displacements and areas and leak rates as well as comparisons to the experimental results. 6 refs., 16 figs., 2 tabs

  10. Reliability of piping system components. Framework for estimating failure parameters from service data

    International Nuclear Information System (INIS)

    Nyman, R.; Hegedus, D.; Tomic, B.; Lydell, B.

    1997-12-01

    This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed 'PFCA'-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies

  11. Reliability of piping system components. Framework for estimating failure parameters from service data

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hegedus, D; Tomic, B [ENCONET Consulting GesmbH, Vienna (Austria); Lydell, B [RSA Technologies, Vista, CA (United States)

    1997-12-01

    This report summarizes results and insights from the final phase of a R and D project on piping reliability sponsored by the Swedish Nuclear Power Inspectorate (SKI). The technical scope includes the development of an analysis framework for estimating piping reliability parameters from service data. The R and D has produced a large database on the operating experience with piping systems in commercial nuclear power plants worldwide. It covers the period 1970 to the present. The scope of the work emphasized pipe failures (i.e., flaws/cracks, leaks and ruptures) in light water reactors (LWRs). Pipe failures are rare events. A data reduction format was developed to ensure that homogenous data sets are prepared from scarce service data. This data reduction format distinguishes between reliability attributes and reliability influence factors. The quantitative results of the analysis of service data are in the form of conditional probabilities of pipe rupture given failures (flaws/cracks, leaks or ruptures) and frequencies of pipe failures. Finally, the R and D by SKI produced an analysis framework in support of practical applications of service data in PSA. This, multi-purpose framework, termed `PFCA`-Pipe Failure Cause and Attribute- defines minimum requirements on piping reliability analysis. The application of service data should reflect the requirements of an application. Together with raw data summaries, this analysis framework enables the development of a prior and a posterior pipe rupture probability distribution. The framework supports LOCA frequency estimation, steam line break frequency estimation, as well as the development of strategies for optimized in-service inspection strategies. 63 refs, 30 tabs, 22 figs.

  12. Structural integrity assessment of piping components

    International Nuclear Information System (INIS)

    Kushwaha, H.S.; Chattopadhyay, J.

    2008-01-01

    Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200-400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to component. The cyclic loading and system compliance effect on the load carrying capacity of piping components are investigated and new recommendations are made. (author)

  13. Compilation of references, data sources and analysis methods for LMFBR primary piping system components

    International Nuclear Information System (INIS)

    Reich, M.; Esztergar, E.P.; Ellison, E.G.; Erdogan, F.; Gray, T.G.F.; Wells, C.W.

    1977-03-01

    A survey and review program for application of fracture mechanics methods in elevated temperature design and safety analysis has been initiated in December of 1976. This is the first of a series of reports, the aim of which is to provide a critical review of the theories of fracture and the application of fracture mechanics methods to life prediction, reliability and safety analysis of piping components in nuclear plants undergoing sub-creep and elevated temperature service conditions

  14. Characterization of cooling systems based on heat pipe principle to control operation temperature of high-tech electronic components

    International Nuclear Information System (INIS)

    Dobre, Tanase; Parvulescu, Oana Cristina; Stoica, Anicuta; Iavorschi, Gustav

    2010-01-01

    The use of cooling systems based on heat pipe principle to control operation temperature of electronic components is very efficient. They have an excellent miniaturizing capacity and this fact creates adaptability for more practical situations. Starting from the observation that these cooling systems are not precisely characterized from the thermal efficiency point of view, the present paper proposes a methodology of data acquisition for their thermal characterization. An experimental set-up and a data processing algorithm are shown to describe the cooling of a heat generating electronic device using heat pipes. A Thermalright SI-97 PC cooling system is employed as a case-study to determine the heat transfer characteristics of a fins cooler.

  15. Functional capability of piping systems

    International Nuclear Information System (INIS)

    Terao, D.; Rodabaugh, E.C.

    1992-11-01

    General Design Criterion I of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations requires, in part, that structures, systems, and components important to safety be designed to withstand the effects of earthquakes without a loss of capability to perform their safety function. ne function of a piping system is to convey fluids from one location to another. The functional capability of a piping system might be lost if, for example, the cross-sectional flow area of the pipe were deformed to such an extent that the required flow through the pipe would be restricted. The objective of this report is to examine the present rules in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, and potential changes to these rules, to determine if they are adequate for ensuring the functional capability of safety-related piping systems in nuclear power plants

  16. Shippingport Station Decommissioning Project: Removal of piping and equipment and removal of primary system components

    International Nuclear Information System (INIS)

    1989-01-01

    This report is a technical synopsis of the removal of contaminated and non-contaminated piping and equipment from the Shippingport Station Decommissioning Project (SSDP). The information is provided as a part of the Technology Transfer Program to document dismantling activities in support of reactor decommissioning. 5 refs., 29 figs., 4 tabs

  17. Smart manufacturing of complex shaped pipe components

    Science.gov (United States)

    Salchak, Y. A.; Kotelnikov, A. A.; Sednev, D. A.; Borikov, V. N.

    2018-03-01

    Manufacturing industry is constantly improving. Nowadays the most relevant trend is widespread automation and optimization of the production process. This paper represents a novel approach for smart manufacturing of steel pipe valves. The system includes two main parts: mechanical treatment and quality assurance units. Mechanical treatment is performed by application of the milling machine with implementation of computerized numerical control, whilst the quality assurance unit contains three testing modules for different tasks, such as X-ray testing, optical scanning and ultrasound testing modules. The advances of each of them provide reliable results that contain information about any failures of the technological process, any deviations of geometrical parameters of the valves. The system also allows detecting defects on the surface or in the inner structure of the component.

  18. ASME code and ratcheting in piping components. Final technical report

    International Nuclear Information System (INIS)

    Hassan, T.; Matzen, V.C.

    1999-01-01

    The main objective of this research is to develop an analysis program which can accurately simulate ratcheting in piping components subjected to seismic or other cyclic loads. Ratcheting is defined as the accumulation of deformation in structures and materials with cycles. This phenomenon has been demonstrated to cause failure to piping components (known as ratcheting-fatigue failure) and is yet to be understood clearly. The design and analysis methods in the ASME Boiler and Pressure Vessel Code for ratcheting of piping components are not well accepted by the practicing engineering community. This research project attempts to understand the ratcheting-fatigue failure mechanisms and improve analysis methods for ratcheting predictions. In the first step a state-of-the-art testing facility is developed for quasi-static cyclic and seismic testing of straight and elbow piping components. A systematic testing program to study ratcheting is developed. Some tests have already been performed and the rest will be completed by summer'99. Significant progress has been made in the area of constitutive modeling. A number of sophisticated constitutive models have been evaluated in terms of their simulations for a broad class of ratcheting responses. From the knowledge gained from this evaluation study two improved models are developed. These models are demonstrated to have promise in simulating ratcheting responses in piping components. Hence, implementation of these improved models in widely used finite element programs, ANSYS and/or ABAQUS, is in progress. Upon achieving improved finite element programs for simulation of ratcheting, the ASME Code provisions for ratcheting of piping components will be reviewed and more rational methods will be suggested. Also, simplified analysis methods will be developed for operability studies of piping components and systems. Some of the future works will be performed under the auspices of the Center for Nuclear Power Plant Structures

  19. Plastics piping systems for industrial applications – Acrylonitrile-butadiene-styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) – Specifications for components and the system – Metric series

    CERN Document Server

    Deutsches Institut für Normung. Berlin

    2003-01-01

    Plastics piping systems for industrial applications – Acrylonitrile-butadiene-styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) – Specifications for components and the system – Metric series

  20. Plastics piping systems for industrial applications : acrylonitrile-butadiene- styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) : specifications for components and the system : metric series

    CERN Document Server

    International Organization for Standardization. Geneva

    2003-01-01

    Plastics piping systems for industrial applications : acrylonitrile-butadiene- styrene (ABS), unplasticized poly(vinyl chloride) (PVC-U) and chlorinated poly(vinyl chloride) (PVC-C) : specifications for components and the system : metric series

  1. PE 100 pipe systems

    CERN Document Server

    Brömstrup, Heiner

    2012-01-01

    English translation of the 3rd edition ""Rohrsysteme aus PE 100"". Because of the considerably increased performance, pipe and pipe systems made from 100 enlarge the range of applications in the sectors of gas and water supply, sewage disposal, industrial pipeline construction and in the reconstruction and redevelopment of defective pipelines (relining). This book applies in particular to engineers, technicians and foremen working in the fields of supply, disposal and industry. Subject matters of the book are all practice-relevant questions regarding the construction, operation and maintenance

  2. Mechanized ultrasonic inspection of austenitic pipe systems

    International Nuclear Information System (INIS)

    Dressler, K.; Luecking, J.; Medenbach, S.

    1999-01-01

    The contribution explains the system of standard testing methods elaborated by ABB ZAQ GmbH for inspection of austenitic plant components. The inspection tasks explained in greater detail are basic materials testing (straight pipes, bends, and pipe specials), and inspection of welds and dissimilar welds. The techniques discussed in detail are those for detection and sizing of defects. (orig./CB) [de

  3. Computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.; Fistedis, S.H.

    1977-01-01

    Integrity of piping systems is one of the main concerns of the safety issues of Liquid Metal Fast Breeder Reactors (LMFBR). Hypothetical core disruptive accidents (HCDA) and water-sodium interaction are two examples of sources of high pressure pulses that endanger the integrity of the heat transport piping systems of LMFBRs. Although plastic wall deformation attenuates pressure peaks so that only pressures slightly higher than the pipe yield pressure propagate along the system, the interaction of these pulses with the different components of the system, such as elbows, valves, heat exchangers, etc.; and with one another produce a complex system of pressure pulses that cause more plastic deformation and perhaps damage to components. A generalized piping component and a tee branching model are described. An optional tube bundle and interior rigid wall simulation model makes such a generalized component model suited for modelling of valves, reducers, expansions, and heat exchangers. The generalized component and the tee branching junction models are combined with the pipe-elbow loop model so that a more general piping system can be analyzed both hydrodynamically and structurally under the effect of simultaneous pressure pulses

  4. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  5. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    Energy Technology Data Exchange (ETDEWEB)

    Nyman, R.; Erixon, S. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B. [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B. [RSA Technologies, Visat, CA (United States)

    1996-09-01

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs.

  6. Reliability of piping system components. Volume 2: PSA LOCA data base. Review of methods for LOCA evaluation since the WASH-1400

    International Nuclear Information System (INIS)

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

    1996-09-01

    The Swedish Nuclear Power Inspectorate has undertaken a project to establish a comprehensive passive components database, validate failure rate parameter estimates and model framework for enhancement of integrating passive components failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure. Approx. 2300 failure events allowed for data exploration in Phase 2 to develop a sound basis for PSA treatment of piping system failure. In addition, a comprehensive review of the current consideration of LOCA in PSA and of all available literature in this area was undertaken. This report is devoted to identification of treatment of LOCA in PSAs. The report contains a detailed review of many programs and dozens of specific PSA studies for different reactor types. This collection and analysis of information together with information for the relational database was used to develop a matrix approach on contribution to LOCA events from different components which are part of the reactor coolant system pressure boundary. The overall conclusion of the work is that although there are some further developments in this area, there is still no significant enhancement of ways how LOCA are considered in PSAs as compared to the mid 70s, only selected studies attempted to address LOCAs in a more comprehensive way. Later phases of this project are expected to contribute to enhancement of treatment of LOCA events in PSA studies. 54 refs, 25 tabs

  7. Degradation mechanisms of small scale piping systems

    International Nuclear Information System (INIS)

    Bartonicek, J.; Koenig, G.; Blind, D.

    1996-01-01

    Operational experience shows that many degradation mechanisms can have an effect on small-scale piping systems. We can see from the analyses carried out that the degradation which has occurred is primarily linked with the fact that these piping systems were classified as being of low safety relevance. This is mainly due to such components being classified into low safety relevance category at the design stage, as well as to the low level of operational monitoring. Since in spite of the variety of designs and operational modes the degradation mechanisms detected may be attributed to the piping systems, we can make decisive statements on how to avoid such degradation mechanisms. Even small-scale piping systems may achieve guaranteed integrity in such cases by taking the appropriate action. (orig.) [de

  8. In Pipe Robot with Hybrid Locomotion System

    Directory of Open Access Journals (Sweden)

    Cristian Miclauş

    2015-06-01

    Full Text Available The first part of the paper covers aspects concerning in pipe robots and their components, such as hybrid locomotion systems and the adapting mechanisms used. The second part describes the inspection robot that was developed, which combines tracked and wheeled locomotion (hybrid locomotion. The end of the paper presents the advantages and disadvantages of the proposed robot.

  9. Experimental benchmark for piping system dynamic response analyses

    International Nuclear Information System (INIS)

    Schott, G.A.; Mallett, R.H.

    1981-01-01

    The scope and status of a piping system dynamics test program are described. A 0.20-m nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed. 3 refs

  10. Experimental benchmark for piping system dynamic-response analyses

    International Nuclear Information System (INIS)

    1981-01-01

    This paper describes the scope and status of a piping system dynamics test program. A 0.20 m(8 in.) nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Particular attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed

  11. Parametric Study on Ultimate Failure Criteria of Elbow Piping Components in Seismically Isolated NPP

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Ki, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    It is well known that the interface pipes between isolated and non-isolated structures will become the most critical in the seismically isolated NPPs. Therefore, seismic performance of such interface pipes should be evaluated comprehensively especially in terms of the seismic fragility capacity. To evaluate the seismic capacity of interface pipes in the isolated NPP, firstly, we should define the failure mode and failure criteria of critical pipe components. Hence, in this study, we performed the dynamic tests of elbow components which were installed in a seismically isolated NPP, and evaluated the ultimate failure mode and failure criteria by using the test results. To do this, we manufactured 25 critical elbow component specimens and performed cyclic loading tests under the internal pressure condition. The failure mode and failure criteria of a pipe component will be varied by the design parameters such as the internal pressure, pipe diameter, loading type, and loading amplitude. From the tests, we assessed the effects of the variation parameters onto the failure criteria. For the tests, we generated the seismic input protocol of relative displacement between the ends of elbow component. In this paper, elbow in piping system was defined as a fragile element and numerical model was updated by component test. Failure mode of piping component under seismic load was defined by the dynamic tests of ultimate pipe capacity. For the interface piping system, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system. In this study, the dynamic tests were performed for the elbow components which were installed in an actual NPPs, and the ultimate failure mode and failure criteria were also evaluated by using the test results.

  12. Heat pipes and heat pipe exchangers for heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, L L; Grakovich, L P; Kiselev, V G; Kurustalev, D K; Matveev, Yu

    1984-01-01

    Heat pipes and heat pipe exchangers are of great importance in power engineering as a means of recovering waste heat of industrial enterprises, solar energy, geothermal waters and deep soil. Heat pipes are highly effective heat transfer units for transferring thermal energy over large distance (tens of meters) with low temperature drops. Their heat transfer characteristics and reliable working for more than 10-15 yr permit the design of new systems with higher heat engineering parameters.

  13. Development of support system for nuclear power plant piping

    International Nuclear Information System (INIS)

    Horino, Satoshi

    1987-01-01

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

  14. Integrated piping structural analysis system

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Analysis methods for structure reliability of piping components

    International Nuclear Information System (INIS)

    Schimpfke, T.; Grebner, H.; Sievers, J.

    2004-01-01

    In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour (BMWA) GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The long-term objective of this development is to provide failure probabilities of passive components for probabilistic safety analysis of nuclear power plants. Up to now the code can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents some of the results of a benchmark analysis in the frame of the European project NURBIM (Nuclear Risk Based Inspection Methodologies for Passive Components). (orig.)

  16. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

    This paper deals with the method used in France for the PWR nuclear plants to derive locations and types of supports of auxiliary and secondary piping systems taking earthquake in account. The successive steps of design are described, then the seismic computation method and its particular conditions of applications for piping are presented. The different types of support (and especially seismic ones) are described and also their conditions of installation. The method used to compare functional tests results and computation results in order to control models is mentioned. Some experiments realised on site or in laboratory, in order to validate models and methods, are presented [fr

  17. Seismic analysis of nuclear piping system

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    2015-01-01

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

  19. Remote-controlled welding during replacement of components and piping

    International Nuclear Information System (INIS)

    Faeser, K.; Huemmeler, A.; Pellkofer, D.

    1986-01-01

    Only on the basis of a thorough fundamental knowledge of nuclear power stations in general and the relevant codes and regulations in particular can extended repair measures, such as the replacement of components or pipelines, be planned and prepared. The application of effective decontamination procedures and shielding measures and a high degree of mechanization of the machining and welding operations will lead to a drastic reduction of the radiation load to which the personnel is exposed. By using highly sophisticated pipe assembling and welding systems the exposure period can be minimized. At the same time a very high level of quality is being reached. The close adherence to the schedule of individual detail operations confirms and justifies the necessity of thorough planning and training of personnel. It may be assumed that in the field of nuclear engineering some pioneer work has been done that will have a stimulating effect on other areas with similar or transferable applications. (orig.) [de

  20. Radiation transmission pipe thickness measurement system

    International Nuclear Information System (INIS)

    Higashi, Yasuhiko

    2010-01-01

    Fuji Electric Systems can be measured from the outer insulation of the transmission Characteristics and radiation detection equipment had been developed that can measure pipe wall thinning in plant and running, the recruitment of another three-beam calculation method by pipe thickness measurement system was developed to measure the thickness of the pipe side. This equipment has been possible to measure the thickness of the circumferential profile of the pipe attachment by adopting automatic rotation. (author)

  1. Piping data retrieval system (PDRS): An integrated package to aid piping layout

    International Nuclear Information System (INIS)

    Vyas, K.N.; Sharma, A.; Susandhi, R.; Basu, S.

    1986-01-01

    An integrated package to aid piping layout has been developed and implemented on PDP-11/34 system at Hall 7. The package allows various equipments to be modelled, consisting of primitive equipment components. The equipment layout for the plant can then be reproduced in the form of drawings such as plan, elevation, isometric or perspective. The package has the built in function to perform hidden line removal among equipments. Once the equipment layout is finalised, the package aids in superimposing the piping as per the specified pipe routine. The report discusses the general capabilities and the major input requirements for the package. (author)

  2. System and Method for Traversing Pipes

    Science.gov (United States)

    Graf, Jodi (Inventor); Pettinger, Ross (Inventor); Azimi, Shaun (Inventor); Magruder, Darby (Inventor); Ridley, Justin (Inventor); Lapp, Anthony (Inventor)

    2017-01-01

    A system and method is provided for traversing inside one or more pipes. In an embodiment, a fluid is injected into the one or more pipes thereby promoting a fluid flow. An inspection device is deployed into the one or more pipes at least partially filled with a flowing fluid. The inspection device comprises a housing wherein the housing is designed to exploit the hydrokinetic effects associated with a fluid flow in one or more pipes as well as maneuver past a variety of pipe configurations. The inspection device may contain one or more sensors capable of performing a variety of inspection tasks.

  3. Calculation of dynamic hydraulic forces in nuclear plant piping systems

    International Nuclear Information System (INIS)

    Choi, D.K.

    1982-01-01

    A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)

  4. Innovative technology summary report: Pipe Explorertrademark system

    International Nuclear Information System (INIS)

    1996-01-01

    The Pipe Explorertrademark system, developed by Science and Engineering Associates, Inc. (SEA), under contract with the US Department of Energy (DOE) Morgantown Energy Technology Center, has been used to transport various characterizing sensors into piping systems that have been radiologically contaminated. DOE's nuclear facility decommissioning program must characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Historically, this has been attempted using hand-held survey instrumentation, surveying only the accessible exterior portions of pipe systems. Various measuring difficulties, and in some cases, the inability to measure threshold surface contamination values and worker exposure, and physical access constraints have limited the effectiveness of traditional survey approaches. The Pipe Explorertrademark system provides a viable alternative. The heart of the system is an air-tight membrane, which is initially spooled inside a canister. The end of the membrane protrudes out of the canister and attaches to the pipe being inspected. The other end of the tubular membrane is attached to the tether and characterization tools. When the canister is pressurized, the membrane inverts and deploys inside the pipe. The characterization detector and its cabling is attached to the tethered end of the membrane. As the membrane is deployed into the pipe, the detector and its cabling is towed into the pipe inside the protective membrane; measurements are taken from within the protective membrane. Once the survey measurements are completed, the process is reversed to retrieve the characterization tools

  5. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

  6. Characterization of radioactive contamination inside pipes with the Pipe Explorer{trademark} system. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, C.D.; Kendrick, D.T.; Lowry, W.; Cramer, E.

    1997-09-30

    The Department of Energy (DOE) is currently in the process of decommissioning and dismantling many of its nuclear materials processing facilities that have been in use for several decades. Site managers throughout the DOE complex must employ the safest and most cost effective means to characterize, remediate and recycle or dispose of hundreds of miles of potentially contaminated piping and duct work. The DOE discovered that standard characterization methods were inadequate for its pipes, drains, and ducts because many of the systems are buried or encased. In response to the DOE`s need for a more specialized characterization technique, Science and Engineering Associates, Inc. (SEA) developed the Pipe Explorer{trademark} system through a DOE Office of Science and Technology (OST) contract administered through the Federal Energy Technology Center (FETC). The purpose of this report is to serve as a comprehensive overview of all phases of the Pipe Explorer{trademark} development project. The report is divided into 6 sections. Section 2 of the report provides an overview of the Pipe Explorer{trademark} system, including the operating principles of using an inverting membrane to tow sensors into pipes. The basic components of the characterization system are also described. Descriptions of the various deployment systems are given in Section 3 along with descriptions of the capabilities of the deployment systems. During the course of the development project 7 types of survey instruments were demonstrated with the Pipe Explorer{trademark} and are a part of the basic toolbox of instruments available for use with the system. These survey tools are described in Section 4 along with their typical performance specifications. The 4 demonstrations of the system are described chronologically in Section 5. The report concludes with a summary of the history, status, and future of the Pipe Explorer{trademark} system in Section 6.

  7. Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system. Final report

    International Nuclear Information System (INIS)

    Cremer, C.D.; Kendrick, D.T.; Lowry, W.; Cramer, E.

    1997-01-01

    The Department of Energy (DOE) is currently in the process of decommissioning and dismantling many of its nuclear materials processing facilities that have been in use for several decades. Site managers throughout the DOE complex must employ the safest and most cost effective means to characterize, remediate and recycle or dispose of hundreds of miles of potentially contaminated piping and duct work. The DOE discovered that standard characterization methods were inadequate for its pipes, drains, and ducts because many of the systems are buried or encased. In response to the DOE's need for a more specialized characterization technique, Science and Engineering Associates, Inc. (SEA) developed the Pipe Explorer trademark system through a DOE Office of Science and Technology (OST) contract administered through the Federal Energy Technology Center (FETC). The purpose of this report is to serve as a comprehensive overview of all phases of the Pipe Explorer trademark development project. The report is divided into 6 sections. Section 2 of the report provides an overview of the Pipe Explorer trademark system, including the operating principles of using an inverting membrane to tow sensors into pipes. The basic components of the characterization system are also described. Descriptions of the various deployment systems are given in Section 3 along with descriptions of the capabilities of the deployment systems. During the course of the development project 7 types of survey instruments were demonstrated with the Pipe Explorer trademark and are a part of the basic toolbox of instruments available for use with the system. These survey tools are described in Section 4 along with their typical performance specifications. The 4 demonstrations of the system are described chronologically in Section 5. The report concludes with a summary of the history, status, and future of the Pipe Explorer trademark system in Section 6

  8. Sensitivity Analysis on Elbow Piping Components in Seismically Isolated NPP under Seismic Loading

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Hee Kun; Hahm, Dae Gi; Kim, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    In this study, the FE model is verified using specimen test results and simulation with parameter variations are conducted. Effective parameters will randomly sampled and used as input values for simulations to be applied to the fragility analysis. pipelines are representative of them because they could undergo larger displacements when they are supported on both isolated and non-isolated structures simultaneously. Especially elbows are critical components of pipes under severed loading conditions such as earthquake action because strain is accumulated on them during the repeated bending of the pipe. Therefore, seismic performance of pipe elbow components should be examined thoroughly based on the fragility analysis. Fragility assessment of interface pipe should take different sources of uncertainty into account. However, selection of important sources and repeated tests with many random input values are very time consuming and expensive, so numerical analysis is commonly used. In the present study, finite element (FE) model of elbow component will be validated using the dynamic test results of elbow components. Using the verified model, sensitivity analysis will be implemented as a preliminary process of seismic fragility of piping system. Several important input parameters are selected and how the uncertainty of them are apportioned to the uncertainty of the elbow response is to be studied. Piping elbows are critical components under cyclic loading conditions as they are subjected large displacement. In a seismically isolated NPP, seismic capacity of piping system should be evaluated with caution. Seismic fragility assessment preliminarily needs parameter sensitivity analysis about the output of interest with different input parameter values.

  9. Fluid structure interaction in piping systems

    Energy Technology Data Exchange (ETDEWEB)

    Svingen, Bjoernar

    1996-12-31

    The Dr. ing. thesis relates to an analysis of fluid structure interaction in piping systems in the frequency domain. The governing equations are the water hammer equations for the liquid, and the beam-equations for the structure. The fluid and structural equations are coupled through axial stresses and fluid continuity relations controlled by the contraction factor (Poisson coupling), and continuity and force relations at the boundaries (junction coupling). A computer program has been developed using the finite element method as a discretization technique both for the fluid and for the structure. This is made for permitting analyses of large systems including branches and loops, as well as including hydraulic piping components, and experiments are executed. Excitations are made in a frequency range from zero Hz and up to at least one thousand Hz. Frequency dependent friction is modelled as stiffness proportional Rayleigh damping both for the fluid and for the structure. With respect to the water hammer equations, stiffness proportional damping is seen as an artificial (bulk) viscosity term. A physical interpretation of this term in relation to transient/oscillating hydraulic pipe-friction is given. 77 refs., 72 figs., 4 tabs.

  10. Evaluation of the influence of seismic restraint characteristics on breeder reactor piping systems

    International Nuclear Information System (INIS)

    Mello, R.M.; Pollono, L.P.

    1979-01-01

    For the Clinch River Breeder Reactor Plant (CRBRP) heat transport system piping within the reactor containment building, dynamic analyses of the piping loops have been performed to study the effect of restraint stiffness on the dynamic behavior of the piping. In addition, analysis and testing of typical CRBRP restraint system components have been performed for the purpose of quantifying and verifying the basic characteristics of the restraints used in the piping system dynamic analysis

  11. Refurbishment of the IEAR1 primary coolant system piping supports

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

    2015-01-01

    A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

  12. Development on methods for evaluating structure reliability of piping components

    International Nuclear Information System (INIS)

    Schimpfke, T.; Grebner, H.; Peschke, J.; Sievers, J.

    2003-01-01

    In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour, GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The development is based on the experience achieved with applications of the public available US code PRAISE 3.10 (Piping Reliability Analysis Including Seismic Events), which was supplemented by additional features regarding the statistical evaluation and the crack orientation. PROST is designed to be more flexible to changes and supplementations. Up to now it can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents a parametric study on the influence by changing the method of stress intensity factor and limit load calculation and the statistical evaluation options on the leak probability of an exemplary pipe with postulated axial crack distribution. Furthermore the resulting leak probability of an exemplary pipe with postulated circumferential crack distribution is compared with the results of the modified PRAISE computer program. The intention of this investigation is to show trends. Therefore the resulting absolute values for probabilities should not be considered as realistic evaluations. (author)

  13. Assessment of cracked pipes in primary piping systems of PWR nuclear reactors

    International Nuclear Information System (INIS)

    Jong, Rudolf Peter de

    2004-01-01

    Pipes related to the Primary System of Pressurized Water Reactors (PWR) are manufactured from high toughness austenitic and low alloy ferritic steels, which are resistant to the unstable growth of defects. A crack in a piping system should cause a leakage in a considerable rate allowing its identification, before its growth could cause a catastrophic rupture of the piping. This is the LBB (Leak Before Break) concept. An essential step in applying the LBB concept consists in the analysis of the stability of a postulated through wall crack in a specific piping system. The methods for the assessment of flawed components fabricated from ductile materials require the use of Elasto-Plastic Fracture Mechanics (EPFM). Considering that the use of numerical methods to apply the concepts of EPFM may be expensive and time consuming, the existence of the so called simplified methods for the assessment of flaws in piping are still considered of great relevance. In this work, some of the simplified methods, normalized procedures and criteria for the assessment of the ductile behavior of flawed components available in literature are described and evaluated. Aspects related to the selection of the material properties necessary for the application of these methods are also discussed. In a next .step, the methods are applied to determine the instability load in some piping configurations under bending and containing circumferential through wall cracks. Geometry and material variations are considered. The instability loads, obtained for these piping as the result of the application of the selected methods, are analyzed and compared among them and with some experimental results obtained from literature. The predictions done with the methods demonstrated that they provide consistent results, with good level of accuracy with regard to the determination of maximum loads. These methods are also applied to a specific Study Case. The obtained results are then analyzed in order to give

  14. Development of pipe layout system

    International Nuclear Information System (INIS)

    Ota, Yoshimi; Yamamoto, Shigeru; Tokumasu, Shinji; Yamaguchi, Yukio; Besho, Hiromi; Sakano, Tatuo.

    1986-01-01

    In the plant design carried out so far, the process up to final drawings has been the repetition of the correction of drawings. This is because the space as the object of design is finite, and it is difficult to lay many pipes efficiently. Especially in nuclear power plants, the quantity of materials required for ensuring the safety and quality control is enormous, and only the skilled engineers having rich experience have become unable to deal with it. The model engineering using plastic models has been adopted, but still there are problems. In order to solve this problem, the development of the system for unitarily managing the various design information of plants with a computer, checking up various design with this information, automatically outputting design drawings and management data, and heightening the quality of design, synchronizing the progress, increasing the speed and saving the labor of design was carried out. This system is versatile and can be used for all plants. The emphasis in the development was placed on compact data structure, rapid picture processing and easy operation. The present status of design and the automation, the basic design of the system, the function of the system, the internal expression of models, the method of picture processing, and the results of application are reported. (Kako, I.)

  15. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

    1987-01-01

    Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-inch and a pressurized 6-inch diameter carbon steel nuclear pipe systems subjected to high-level shaking have been accomplished. The high-level shaking loads needed to cause failure were much higher than ASME Code rules would permit with present design limits. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occured in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate reasonably well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules to reduce unneeded conservatisms and to cover the ratchet-fatigue failure mode may be appropriate

  16. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

    1986-01-01

    Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-in. and a pressurized 6-in. diameter carbon steel nuclear pipe systems subjected to high level shaking have been accomplished. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occurred in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate very well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules may be appropriate to cover the ratchet-fatigue failure mode

  17. 46 CFR 58.60-7 - Industrial systems: Piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Industrial systems: Piping. 58.60-7 Section 58.60-7 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING MAIN AND AUXILIARY MACHINERY AND RELATED SYSTEMS Industrial Systems and Components on Mobile Offshore Drilling Units (MODU...

  18. Nuclear piping system damping data studies

    International Nuclear Information System (INIS)

    Ware, A.G.; Arendts, J.G.

    1985-01-01

    A programm has been conducted at the Idaho National Engineering Laboratory to study structural damping data for nuclear piping systems and to evaluate if changes in allowable damping values for structural seismic analyses are justified. The existing pipe damping data base was examined, from which a conclusion was made that there were several sets of data to support higher allowable values. The parameters which most influence pipe damping were identified and an analytical investigation demonstrated that increased damping would reduce the required number of seismic supports. A series of tests on several laboratory piping systems was used to determine the effect of various parameters such as types of supports, amplitude of vibration, frequency, insulation, and pressure on damping. A multiple regression analysis was used to statistically assess the influence of the various parameters on damping, and an international pipe damping data bank has been formed. (orig.)

  19. Internal testing of pipe systems with IRIS inspection system

    International Nuclear Information System (INIS)

    1986-01-01

    The internal piping inspection system IRIS allows inside testing of pipes with an internal diameter of NW 70 as a minimum, and of any horizontal or vertical layout of the piping system. Visual testing is done by means of an integrated CCD video system with high resolution power. Technical data are given and examples of applications, in the German and English language. (DG) [de

  20. Ratcheting study in pressurized piping components under cyclic loading at room temperature

    International Nuclear Information System (INIS)

    Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.

    2006-07-01

    The nuclear power plant piping components and systems are often subjected to reversing cyclic loading conditions due to various process transients, seismic and other events. Earlier the design of piping subjected to seismic excitation was based on the principle of plastic collapse. It is believed that during such events, fatigue-ratcheting is likely mode of failure of piping components. The 1995 ASME Boiler and Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. Experimental and analytical studies are carried out to understand this failure mechanism. The biaxial ratcheting characteristics of SA 333, Gr. 6 steel and SS 304 stainless steel at room temperature are investigated in the present work. Experiments are carried out on straight pipes subjected to internal pressure and cyclic bending load applied in a three point and four point bend test configurations. A shake table test is also carried out on a pressurized elbow by applying sinusoidal base excitation. Analytical simulation of ratcheting in the piping elements is carried out. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. (author)

  1. Piping system damping data at higher frequencies

    International Nuclear Information System (INIS)

    Ware, A.G.

    1987-01-01

    Research has been performed at the Idaho National Engineering Laboratory (INEL) for the United States Nuclear Regulatory Commission (USNRC) to determine best-estimate damping values for dynamic analyses of nuclear piping systems excited in the 20 to 100 Hz frequency range. Vibrations in this frequency range are typical of fluid-induced transients, for which no formal pipe damping guidelines exist. The available data found in the open literature and the USNRC/INEL nuclear piping damping data bank were reviewed, and a series of tests on a straight 3-in. (76-mm) piping system and a 5-in. (127-mm) system with several bends and elbows were conducted as part of this research program. These two systems were supported with typical nuclear piping supports that could be changed from test to test during the series. The resulting damping values were ≥ those of the Pressure Vessel Research Committee (PVRC) proposal for unisulated piping. Extending the PVRC damping curve from 20 to 100 Hz at 3% of critical damping would give a satisfactory representation of the test data. This position has been endorsed by the PVRC Technical Committee on Piping Systems. 14 refs

  2. Development of a simplified piping support system

    International Nuclear Information System (INIS)

    Leung, J.; Anderson, P.H.; Tang, Y.K.; Kassawara, R.P.; Tang, H.T.

    1987-01-01

    This paper presents the results of experimental and analytical studies for developing a simplified piping support system (SPSS) for nuclear power piping in place of snubbers. The basic concept of the SPSS is a passive seismic support system consisting of limit stops. Large gaps are provided to allow for free thermal expansion during normal plant operation while preventing excessive displacement during a seismic event. The results are part of a research and development program sponsored by EPRI. (orig./HP)

  3. Development of a simplified piping support system

    International Nuclear Information System (INIS)

    Leung, J.; Anderson, P.H.; Tang, Y.K.; Kassawara, R.P.; Tang, H.T.

    1987-01-01

    This paper presents the results of experimental and analytical studies for developing a simplified piping support system (SPSS) for nuclear power piping in place of snubbers. The basic concept of the SPSS is a passive seismic support system consisting of limit stops. Large gaps are provided to allow for free thermal expansion during normal plant operation while preventing excessive displacement during a seismic event. The results are part of a research and development program sponsored by the Electric Power Research Institute

  4. Pressure piping systems examination. 2. ed

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-01

    This Code is Part 13 of the IP Model Code of Safe Practice in the Petroleum Industry. Its purpose is to provide a guide to safe practices in the in-service examination and test of piping systems used in the petroleum and chemical industries. The Code gives general requirements regarding the provision and maintenance of adequate documentation, in-service examination, the control of modifications and repairs, examination frequency, protective devices and testing of piping systems. (author)

  5. 46 CFR 153.280 - Piping system design.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Piping system design. 153.280 Section 153.280 Shipping... BULK LIQUID, LIQUEFIED GAS, OR COMPRESSED GAS HAZARDOUS MATERIALS Design and Equipment Piping Systems and Cargo Handling Equipment § 153.280 Piping system design. (a) Each cargo piping system must meet...

  6. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  7. Analysis of a piping system for requalification

    International Nuclear Information System (INIS)

    Hsieh, B.J.; Tang, Yu.

    1992-01-01

    This paper discusses the global stress analysis required for the seismic/structural requalification of a reactor secondary piping system in which minor defects (flaws) were discovered during a detailed inspection. The flaws in question consisted of weld imperfections. Specifically, it was necessary to establish that the stresses at the flawed sections did not exceed the allowables and that the fatigue life remained within acceptable limits. At the same time the piping system had to be qualified for higher earthquake loads than those used in the original design. To accomplish these objectives the nominal stress distributions in the piping system under the various loads (dead load, thermal load, wind load and seismic load) were determined. First a best estimate finite element model was developed and calculations were performed using the piping analysis modules of the ANSYS Computer Code. Parameter studies were then performed to assess the effect of physically reasonable variations in material, structural, and boundary condition characteristics. The nominal stresses and forces so determined, provided input for more detailed analyses of the flawed sections. Based on the reevaluation, the piping flaws were judged to be benign, i.e., the piping safety margins were acceptable inspite of the increased seismic demand. 13 refs

  8. Stress analysis of piping systems and piping supports. Documentation

    International Nuclear Information System (INIS)

    Rusitschka, Erwin

    1999-01-01

    The presentation is focused on the Computer Aided Tools and Methods used by Siemens/KWU in the engineering activities for Nuclear Power Plant Design and Service. In the multi-disciplinary environment, KWU has developed specific tools to support As-Built Documentation as well as Service Activities. A special application based on Close Range Photogrammetry (PHOCAS) has been developed to support revamp planning even in a high level radiation environment. It comprises three completely inter-compatible expansion modules - Photo Catalog, Photo Database and 3D-Model - to generate objects which offer progressively more utilization and analysis options. To support the outage planning of NPP/CAD-based tools have been developed. The presentation gives also an overview of the broad range of skills and references in: Plant Layout and Design using 3D-CAD-Tools; evaluation of Earthquake Safety (Seismic Screening); Revamps in Existing Plants; Inter-disciplinary coordination of project engineering and execution fields; Consulting and Assistance; Conceptual Studies; Stress Analysis of Piping Systems and Piping Supports; Documentation; Training and Supports in CAD-Design, etc. All activities are performed to the greatest extent possible using proven data-processing tools. (author)

  9. Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

    International Nuclear Information System (INIS)

    Pineiro, Oscar; Fernandez, Carlos; Medrano, Mercedes; Liniers, Macarena; Botija, Jose; Alonso, Javier; Sarasola, Xabier; Damiani, Carlo

    2009-01-01

    The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: - Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. - The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. - The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. - The alignment of the pipes is a critical issue concerning the remote welding. A proper alignment

  10. Small pipe characterization system (SPCS) conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, M.O.; Ferrante, T.A.; McKay, M.D.

    1995-01-01

    Throughout the Department of Energy (DOE) complex there are many facilities that have been identified for Decontamination and Decommissioning (D&D). As processes are terminated or brought off-line, facilities are placed on the inactive list, and facility managers and site contractors are required to assure a safe and reliable decommissioning and transition of these facilities to a clean final state. Decommissioning of facilities requires extensive reliable characterization, decontamination and in some cases dismantlement. Characterization of piping systems throughout the DOE complex is becoming more and more necessary. In addition to decommissioning activities, characterization activities are performed as part of surveillance and maintenance (S&M). Because of the extent of contamination, all inactive facilities require some type of S&M. These S&M activities include visual assessment, equipment and material accounting, and maintenance. The majority of the inactive facilities have piping systems 3 inches or smaller that are inaccessible because they are contaminated, imbedded in concrete, or run through hot cells. Many of these piping systems have been inactive for a number of years and there exists no current system condition information or the historical records are poor and/or missing altogether. Many of these piping systems are placed on the contaminated list, not because of known contamination, but because of the risk of internal contamination. Many of the piping systems placed on the contamination list may not have internal contamination. Because there is a potential however, they are treated as such. The cost of D&D can be greatly reduced by identifying and removing hot spot contamination, leaving clean piping to be removed using conventional methods. Accurate characterization of these piping systems is essential before, during and after all D&D activities.

  11. Small pipe characterization system (SPCS) conceptual design

    International Nuclear Information System (INIS)

    Anderson, M.O.; Ferrante, T.A.; McKay, M.D.

    1995-01-01

    Throughout the Department of Energy (DOE) complex there are many facilities that have been identified for Decontamination and Decommissioning (D ampersand D). As processes are terminated or brought off-line, facilities are placed on the inactive list, and facility managers and site contractors are required to assure a safe and reliable decommissioning and transition of these facilities to a clean final state. Decommissioning of facilities requires extensive reliable characterization, decontamination and in some cases dismantlement. Characterization of piping systems throughout the DOE complex is becoming more and more necessary. In addition to decommissioning activities, characterization activities are performed as part of surveillance and maintenance (S ampersand M). Because of the extent of contamination, all inactive facilities require some type of S ampersand M. These S ampersand M activities include visual assessment, equipment and material accounting, and maintenance. The majority of the inactive facilities have piping systems 3 inches or smaller that are inaccessible because they are contaminated, imbedded in concrete, or run through hot cells. Many of these piping systems have been inactive for a number of years and there exists no current system condition information or the historical records are poor and/or missing altogether. Many of these piping systems are placed on the contaminated list, not because of known contamination, but because of the risk of internal contamination. Many of the piping systems placed on the contamination list may not have internal contamination. Because there is a potential however, they are treated as such. The cost of D ampersand D can be greatly reduced by identifying and removing hot spot contamination, leaving clean piping to be removed using conventional methods. Accurate characterization of these piping systems is essential before, during and after all D ampersand D activities

  12. Background of SIFs and Stress Indices for Moment Loadings of Piping Components

    International Nuclear Information System (INIS)

    Wais, E. A.; Rodabaugh, E. C.

    2005-01-01

    This report provides background information, references, and equations for twenty-four piping components (thirteen component SIFs and eleven component stress indices) that justify the values or expressions for the SIFs and indices

  13. Using Low-Frequency Phased Arrays to Detect Cracks in Cast Austenitic Piping Components

    International Nuclear Information System (INIS)

    Anderson, Michael T.; Cumblidge, Stephen E.; Doctor, Steven R.

    2005-01-01

    As part of a multi-year program funded by the United States Nuclear Regulatory Commission (US NRC) to address NDE reliability of inservice inspection (ISI) programs, recent studies conducted at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, have focused on assessing novel NDE approaches for the inspection of coarse-grained, cast stainless steel reactor components. The primary objective of this work is to provide information to the US NRC on the utility, effectiveness and reliability of ultrasonic testing (UT) and eddy current testing (ET) inspection techniques as related to the ISI of primary piping components in pressurized water reactors (PWRs). This paper describes progress, recent developments and early results from an assessment of a portion of this work relating to the ultrasonic low frequency phased array inspection technique. Westinghouse Owner's Group (WOG) cast stainless steel pipe segments with thermal and mechanical fatigue cracks, PNNL samples containing thermal fatigue cracks and several blank vintage specimens having very coarse grains that are representative of early centrifugally cast piping installed in PWRs, are being used for assessing the inspection method. The phased array approach was implemented using an R/D Tech Tomoscan III system operating at 1.0 MHz and 500 kHz, providing composite volumetric images of the samples. Several dual, transmit-receive, custom designed low-frequency arrays are employed in laboratory trials. Results from laboratory studies for assessing detection of thermal and mechanical fatigue cracks in cast stainless steel piping welds are discussed

  14. How simulation of failure risk can improve structural reliability - application to pressurized components and pipes

    OpenAIRE

    Cioclov, Dimitru Dragos

    2013-01-01

    Probabilistic methods for failure risk assessment are introduced, with reference to load carrying structures, such as pressure vessels (PV) and components of pipes systems. The definition of the failure risk associated with structural integrity is made in the context of the general approach to structural reliability. Sources of risk are summarily outlined with emphasis on variability and uncertainties (V&U) which might be encountered in the analysis. To highlight the problem, in its practical...

  15. Volatile organic components migrating from plastic pipes (HDPE, PEX and PVC) into drinking water.

    Science.gov (United States)

    Skjevrak, Ingun; Due, Anne; Gjerstad, Karl Olav; Herikstad, Hallgeir

    2003-04-01

    High-density polyethylene pipes (HDPE), crossbonded polyethylene pipes (PEX) and polyvinyl chloride (PVC) pipes for drinking water were tested with respect to migration of volatile organic components (VOC) to water. The odour of water in contact with plastic pipes was assessed according to the quantitative threshold odour number (TON) concept. A major migrating component from HDPE pipes was 2,4-di-tert-butyl-phenol (2,4-DTBP) which is a known degradation product from antioxidants such as Irgafos 168(R). In addition, a range of esters, aldehydes, ketones, aromatic hydrocarbons and terpenoids were identified as migration products from HDPE pipes. Water in contact with HDPE pipes was assessed with respect to TON, and values > or =4 were determined for five out of seven brands of HDPE pipes. The total amount of VOC released to water during three successive test periods were fairly constant for the HDPE pipes. Corresponding migration tests carried out for PEX pipes showed that VOC migrated in significant amounts into the test water, and TON >/=5 of the test water were observed in all tests. Several of the migrated VOC were not identified. Oxygenates predominated the identified VOC in the test water from PEX pipes. Migration tests of PVC pipes revealed few volatile migrants in the test samples and no significant odour of the test water.

  16. System for Cooling of Electronic Components

    Science.gov (United States)

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

    2017-01-01

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

  17. Magnetic Actuator with Multiple Vibration Components Arranged at Eccentric Positions for Use in Complex Piping

    Directory of Open Access Journals (Sweden)

    Hiroyuki Yaguchi

    2016-06-01

    Full Text Available This paper proposes a magnetic actuator using multiple vibration components to perform locomotion in a complex pipe with a 25 mm inner diameter. Due to the desire to increase the turning moment in a T-junction pipe, two vibration components were attached off-center to an acrylic plate with an eccentricity of 2 mm. The experimental results show that the magnetic actuator was able to move at 40.6 mm/s while pulling a load mass of 20 g in a pipe with an inner diameter of 25 mm. In addition, this magnetic actuator was able to move stably in U-junction and T-junction pipes. If a micro-camera is implemented in the future, the inspection of small complex pipes can be enabled. The possibility of inspection in pipes with a 25 mm inner diameter was shown by equipping the pipe with a micro-camera.

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

    International Nuclear Information System (INIS)

    Hedtjaern Swaling, Vidar; Olsson, Anders

    2011-02-01

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

  19. Design and analysis for piping systems

    International Nuclear Information System (INIS)

    Sterkel, H.-P.; Cutrim, J.H.C.

    1981-01-01

    The procedure and the typical techniques that are used in NUCLEN for the design and the calculation of the piping of Nuclear Plants. The classification system are generically described and the analysis techniques which are used for the design and verification of the piping systems, i.e. pressure design for the dimensioning of the wallthicknesses, temperature and dead weight analysis together with determination of support points, are shown. The techniques of dynamic design and analyses are described for earthquake and pressure impulse loadings. (Author) [pt

  20. Evaluation of clamp effects on LMFBR piping systems

    International Nuclear Information System (INIS)

    Jones, G.L.

    1980-01-01

    Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness

  1. Flow induced vibrations in a PWR piping system

    International Nuclear Information System (INIS)

    Seligmann, D.; Guillou, J.

    1995-11-01

    During a recurring bench test of an operating system, high amplitude vibrations have been observed on a safety piping system of a nuclear power plant. Due to the source of the pumps, these vibrations lead to wear damage and it is therefore necessary to estimate the life time of the piping system. This paper describes the methodology used to study the dynamic behaviour and to analyze the damage of a piping system submitted to internal flow. Starting from an experimental modal analysis of the piping system when not i service, we analyse the main parameters of the mechanical behaviour. Following this analysis, we obtain a mechanical model fitting the first experimental modes. On this basis, we build a vibro-acoustical model. This model takes into account the influence of the acoustical pipe length, both above and below the mechanical part, the modelling of acoustical components, the speed of sound. We did not experimentally characterize the pumps. Therefore, we use a numerical model in order to simulate the behaviour of the pumps. This model is based on the theory of the transfer matrix and takes into account the geometric and the hydraulic characteristics of the pump.The modelling of both sources (suction and discharge) connected to the pump is formed by contributions from a source corresponding to the turbulent noise at low frequency, a source at blade passage frequency. This model has been experimentally validated in a laboratory. The final results of the modelling of the complete piping system are in a complete accord with experimental measurements. (author). 3 refs., 7 figs

  2. The IPIRG-1 pipe system fracture tests: Experimental results

    International Nuclear Information System (INIS)

    Scott, P.; Olson, R.J.; Wilkowski, G.M.

    1994-01-01

    As part of the First International Piping Integrity Research Group (IPIRG-1) program, six dynamic pipe system experiments were conducted. The objective of these experiments was to generate experimental data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system subjected to combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The pipe system evaluated was an expansion loop with over 30 m (100 feet) of 16-inch nominal diameter Schedule 100 pipe. The experimental facility was equipped with special hardware to ensure that system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe system experiments. The uncracked-pipe experiment was conducted to evaluate the piping system damping and natural frequency characteristics. The cracked-pipe experiments were conducted to evaluate the fracture behavior, piping system response, and fracture stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided the tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Key results from the six pipe system experiments and material characterization efforts are presented. Detailed analyses will be published in a companion paper

  3. BOA: Pipe asbestos insulation removal robot system

    International Nuclear Information System (INIS)

    Schempf, H.; Bares, J.; Schnorr, W.

    1995-01-01

    The BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to the high labor costs and high level of radioactive contamination, making manual removal extremely costly and highly inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee

  4. Reconfigurable manufacturing execution system for pipe cutting

    Science.gov (United States)

    Yin, Y. H.; Xie, J. Y.

    2011-08-01

    This article presents a reconfigurable manufacturing execution system (RMES) filling the gap between enterprise resource planning and resource layer for pipe-cutting production with mass customisation and rapid adaptation to dynamic market, which consists of planning and scheduling layer and executive control layer. Starting from customer's task and process requirements, the cutting trajectories are planned under generalised mathematical model able to reconfigure in accordance with various intersecting types' joint, and all tasks are scheduled by nesting algorithm to maximise the utilisation rate of rough material. This RMES for pipe cutting has been effectively implemented in more than 100 companies.

  5. 33 CFR 127.1101 - Piping systems.

    Science.gov (United States)

    2010-07-01

    ...) WATERFRONT FACILITIES WATERFRONT FACILITIES HANDLING LIQUEFIED NATURAL GAS AND LIQUEFIED HAZARDOUS GAS Waterfront Facilities Handling Liquefied Hazardous Gas Design and Construction § 127.1101 Piping systems... pipeline on a pier or wharf must be located so that it is not exposed to physical damage from vehicular...

  6. Split heat pipe heat recovery system

    OpenAIRE

    E. Azad

    2008-01-01

    This paper describes a theoretical analysis of a split heat pipe heat recovery system. The analysis is based on an Effectiveness-NTU approach to deduce its heat transfer characteristics. In this study the variation of overall effectiveness of heat recovery with the number of transfer units are presented. Copyright , Manchester University Press.

  7. Manufacture of piping components for nuclear power plants

    International Nuclear Information System (INIS)

    Bartecek, R.

    1983-01-01

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

  8. Evaluation of seismic margins for an in-plant piping system

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

  9. Fatigue analysis for analytically overloaded piping components and valves in nuclear power plants

    International Nuclear Information System (INIS)

    Charalambus, B.

    1992-01-01

    Lately, in connection with life extension aspects of power plants, an increasingly accurate determination of the lifetime of components in nuclear stations is being required. In order to assess reliably current fatigue levels in piping systems, variables such as pressure, temperature, and resultant force and moment transients as well as analytical methods which take into account the real operational history must be considered. This paper presents a method for analyzing the transient heat transfer between fluid and pipe wall in order to investigate effects which until now have been assumed conservatively to be caused by a sudden jump in temperature. Further, an example is given showing that the K e factor approach in current design codes for performing simplified elastic-plastic fatigue analyses is conservative. (orig.)

  10. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    International Nuclear Information System (INIS)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik; Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae

    2015-01-01

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  11. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik [School of Materials Science and Engineering, Andong National University, Andong (Korea, Republic of); Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae [Power Engineering Research Institute, KEPCO Engineering and Construction Company, Seongnam (Korea, Republic of)

    2015-02-15

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  12. CAPD Software Development for Automatic Piping System Design: Checking Piping Pocket, Checking Valve Level and Flexibility

    International Nuclear Information System (INIS)

    Ari Satmoko; Edi Karyanta; Dedy Haryanto; Abdul Hafid; Sudarno; Kussigit Santosa; Pinitoyo, A.; Demon Handoyo

    2003-01-01

    One of several steps in industrial plant construction is preparing piping layout drawing. In this drawing, pipe and all other pieces such as instrumentation, equipment, structure should be modeled A software called CAPD was developed to replace and to behave as piping drafter or designer. CAPD was successfully developed by adding both subprogram CHKUPIPE and CHKMANV. The first subprogram can check and gives warning if there is piping pocket in the piping system. The second can identify valve position and then check whether valve can be handled by operator hand The main program CAPD was also successfully modified in order to be capable in limiting the maximum length of straight pipe. By limiting the length, piping flexibility can be increased. (author)

  13. 30 CFR 75.1905-1 - Diesel fuel piping systems.

    Science.gov (United States)

    2010-07-01

    ... facility. (g) Diesel fuel piping systems from the surface shall only be used to transport diesel fuel... storage facility. (h) The diesel fuel piping system must not be located in a borehole with electric power... entry as electric cables or power lines. Where it is necessary for piping systems to cross electric...

  14. Piping systems, containment pre-stressing and steel ventilation chimney

    International Nuclear Information System (INIS)

    Stuessi, U.

    1996-01-01

    Units 5 and 6 of NPP Kozloduy have been designed initially for seismic levels which are considered too low today. In the frame of an IAEA Coordinated Research Programme, a Swiss team has been commissioned by Natsionalna Elektricheska Kompania, Sofia, to analyse the relevant piping system, the containment prestressing and the steel ventilation chimney and to recommend upgrade measures for adequate seismic capacity where applicable. Seismic input had been specified by and agreed upon earlier by IAEA experts. The necessary investigations have been performed in 1995 and discussed with internationally recognized experts. The main results may be summarized as follows: Upgrades are necessary at different piping sy ports (additional snubbers or viscous dampers). These fixes can be done easily at low cost. The containment prestressing tendons are adequately designed for the specified load combinations. However, unfavourable construction features endanger the reliability. It is therefore strongly recommended to replace the tendons stepwise and to upgrade the existing monitoring system. Finally, the steel ventilation chimney may not withstand a seismic event, however the containment and diesel generator building will not be destroyed at possible impact by the chimney. On the other hand the roof of the main building has to be reinforced partially. It is recommended to continue the project for 1996 and 1997 to implement the upgrade measures mentioned above, to analyse the remaining piping systems and to consolidate all results obtained by different research groups of the IAEA programme with respect to piping systems including components and tanks

  15. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  16. Numerical simulation of residual stress in piping components at Framatome-ANP

    International Nuclear Information System (INIS)

    Gilies, P.; Franco, C.; Cipiere, M.-F.; Ould, P.

    2005-01-01

    Numerous manufacturing processes induce residual stresses and distortions in piping components and associated welds: quenching of cast pipings, machining and welding. In Pressurized Water Reactors, most of the components have a large thickness for sustaining pressure and distortions are a minor source of concern. This is not the case for residual stresses which may have a strong influence on several type of damage such as fatigue, corrosion, brittle fracture. In low toughness components, residual stress fields may contribute to ductile tearing initiation. These potential damages are mitigated after welding by stress relief heat treatment, which is applied in a systematic manner to ferritic components of the primary system in nuclear reactors. This treatment is not applied on austenitic piping for which the heat treatment temperature is limited due to the risk of sensitization and residual stresses are difficult to eliminate completely. Since on site measurements are costly and difficult to perform, numerical simulation appears to be an attractive tool for estimating residual stress distributions. Framatome-ANP is working on modelling manufacturing processes with that purpose in mind. This paper presents three kinds of applications illustrating efforts on welding, quenching and machining simulation. First a comparison is shown between computations and measurements of residual stress induced by welding of a dissimilar weld metal junction. Then numerical simulations of quenching of a cast stainless steel nozzle are presented. Finally quenching followed by machining and grinding of this cast component are considered in a full simulation of the manufacturing process. Computed distortions and residual stresses are compared with experimental measurements at different stages of the manufacturing process. (authors)

  17. Characterization of radioactive contamination inside pipes with the Pipe Explorer{sup trademark} system

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, C.D.; Lowry, W.; Cramer, E. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)] [and others

    1995-10-01

    The U.S. Department of Energy`s nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Historically, this has been attempted using hand held survey instrumentation, surveying only the accessible exterior portions of pipe systems. Difficulty, or inability of measuring threshold surface contamination values, worker exposure, and physical access constraints have limited the effectiveness of this approach. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer{trademark} system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane.

  18. Subprogram Calculating The Distance Between Pipe And Plane For Automatic Piping System Design

    International Nuclear Information System (INIS)

    Satmoko, Ari

    2001-01-01

    DISTLNPL subprogram was created using Auto LISP software. This subprogram is planned to complete CAPD (Computer Aided Piping Design) software being developed. The CAPD works under the following method: suggesting piping system line and evaluating whether any obstacle allows the proposed line to be constructed. DISTLNPL is able to compute the distance between pipe and any equipment having plane dimension such as wall, platform, floors, and so on. The pipe is modeled by using a line representing its axis, and the equipment is modeled using a plane limited by some lines. The obtained distance between line and plane gives information whether the pipe crosses the equipment. In the case of crashing, the subprogram will suggest an alternative point to be passed by piping system. So far, DISTLNPL has not been able to be accessed by CAPD yet. However, this subprogram promises good prospect in modeling wall, platform, and floors

  19. Investigation on field method using strain measurement on pipe surface to measure pressure pulsation in piping systems

    International Nuclear Information System (INIS)

    Maekawa, Akira; Tsuji, Takashi; Takahashi, Tsuneo; Kato, Minoru

    2013-01-01

    Accurate evaluation of the occurrence location and amplitude of pressure pulsations in piping systems can lead to efficient plant maintenance by preventing fatigue failure of piping and components because the pulsations can be one of the main causes of vibration fatigue and acoustic noise in piping. A non-destructive field method to measure pressure pulsations easily and directly was proposed to replace conventional methods such as prediction using numerical simulations and estimation using locally installed pressure gauges. The proposed method was validated experimentally by measuring pulsating flow in a mock-up piping system. As a result, it was demonstrated that the method to combine strain measurement on the outer surface of pipe with the formula for thick-walled cylinders could measure amplitudes and behavior of the pressure pulsations with a practical accuracy. Factors affecting the measurement accuracy of the proposed method were also discussed. Furthermore, the applicability of the formula for thin-walled cylinders was examined for variously shaped pipes. (author)

  20. Acoustic analysis of a piping system

    International Nuclear Information System (INIS)

    Misra, A.S.; Vijay, D.K.

    1996-01-01

    Acoustic pulsations in the Darlington Nuclear Generating Station, a 881 MW CANDU, primary heat transport piping system caused fuel bundle failures under short term operations. The problem was successfully analyzed using the steady-state acoustic analysis capability of the ABAQUS program. This paper describes in general, modelling of low amplitude acoustic pulsations in a liquid filled piping system using ABAQUS. The paper gives techniques for estimating the acoustic medium properties--bulk modulus, fluid density and acoustic damping--and modelling fluid-structure interactions at orifices and elbows. The formulations and techniques developed are benchmarked against the experiments given in 3 cited references. The benchmark analysis shows that the ABAQUS results are in excellent agreement with the experiments

  1. Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

    Energy Technology Data Exchange (ETDEWEB)

    Pineiro, Oscar; Fernandez, Carlos [TECNATOM Avda. Montes de Oca 28700 S Sebastian de los Reyes, Madrid (Spain); Medrano, Mercedes [EURATOM-CIEMAT Association for Fusion. Avda. Complutense, 22. 28040 Madrid (Spain)], E-mail: mercedes.medrano@ciemat.es; Liniers, Macarena; Botija, Jose; Alonso, Javier; Sarasola, Xabier [EURATOM-CIEMAT Association for Fusion. Avda. Complutense, 22. 28040 Madrid (Spain); Damiani, Carlo [EFDA-Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain)

    2009-06-15

    The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper: - Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes. - The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above. - The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment. - The alignment of the pipes is a critical issue concerning the remote welding

  2. Seismic testing and analysis of a prototypic nonlinear piping system

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.

    1982-11-01

    A series of seismic tests and analyses of a nonlinear Fast Flux Test Facility (FFTF) prototypic piping system are described, and measured responses are compared with analytical predictions. The test loop was representative of a typical LMFBR insulated small bore piping system and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps. Various piping support configurations were tested and analyzed to evaluate the effects of free play and other nonlinear stiffness characteristics on the piping system response

  3. On the computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.-W.; Fistedis, S.H.

    1977-01-01

    A two-dimensional coupled hydrodynamic-structural response analysis of piping systems is described. Implicit Continuous-Fluid Eulerian (ICE) technique is utilized in the hydrodynamics while a finite-element technique is used in the structural analysis. Different piping components such as elbows, valves, reducers, expansions, heat exchangers, and tees are modelled and coupled with the straight pipe model. An axisymmetric general component model that can be used in modelling valves, reducers, expansions, and heat exchangers is described. At the inlet and outlet region of such component the cross-sectional area may change suddently or gradually, or many not change at all. Among the options available in this model are deformable exterior walls, interior rigid wall simulation, and tube bundle effect. Exterior walls of pipes and components are treated as thin axisymmetric shell. A convected coordinate explicit finite-element scheme for large displacement small strain, elastic-plastic material behavior in which membrane and bending strengths are accounted for is employed. The strains are linearly related to the displacement of the element relative to its convective coordinates, and similarly, the nodal forces are linearly related to the elements stresses. The coupling of the hydrodynamics and structural problems is done in such a way that the hydrodynamics supplies the structure with a pressure loading and the structure supplying the hydrodynamics with a moving boundary condition. Because of the difficulties of handling interior walls that may occupy partial zones, the walls are assumed rigid and limited in their orientation to be parallel to the radial or axial directions, their position to zone boundaries, and their thickness to zero

  4. Pipe Crawler internal piping characterization system. Deactivation and decommissioning focus area. Innovative Technology Summary Report

    International Nuclear Information System (INIS)

    1998-02-01

    Pipe Crawler reg-sign is a pipe surveying system for performing radiological characterization and/or free release surveys of piping systems. The technology employs a family of manually advanced, wheeled platforms, or crawlers, fitted with one or more arrays of thin Geiger Mueller (GM) detectors operated from an external power supply and data processing unit. Survey readings are taken in a step-wise fashion. A video camera and tape recording system are used for video surveys of pipe interiors prior to and during radiological surveys. Pipe Crawler reg-sign has potential advantages over the baseline and other technologies in areas of cost, durability, waste minimization, and intrusiveness. Advantages include potentially reduced cost, potential reuse of the pipe system, reduced waste volume, and the ability to manage pipes in place with minimal disturbance to facility operations. Advantages over competing technologies include potentially reduced costs and the ability to perform beta-gamma surveys that are capable of passing regulatory scrutiny for free release of piping systems

  5. Fatigue evaluation of piping systems with limited vibration test data

    International Nuclear Information System (INIS)

    Huang, S.N.

    1990-11-01

    The safety-related piping in a nuclear power plant may be subjected to pump- or fluid-induced vibrations that, in general, affect only local areas of the piping systems. Pump- or fluid-induced vibrations typically are characterized by low levels of amplitudes and a high number of cycles over the lifetime of plant operation. Thus, the resulting fatigue damage to the piping systems could be an important safety concern. In general, tests and/or analyses are used to evaluate and qualify the piping systems. Test data, however, may be limited because of lack of instrumentation in critical piping locations and/or because of difficulty in obtaining data in inaccessible areas. This paper describes and summarizes a method to use limited pipe vibration test data, along with analytical harmonic response results from finite-element analyses, to assess the fatigue damage of nuclear power plant safety-related piping systems. 5 refs., 2 figs., 11 tabs

  6. Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

    International Nuclear Information System (INIS)

    Kendrick, D.T.; Cremer, C.D.; Lowry, W.; Cramer, E.

    1995-01-01

    The U.S. Department of Energy's nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer trademark system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane. Advantages of this approach include the capability of deploying through constrictions in the pipe, around 90 degrees bends, vertically up and down, and in slippery conditions. Because the detector is transported inside the membrane (which is inexpensive and disposable), it is protected from contamination, which eliminates cross-contamination. Characterization sensors that have been demonstrated with the system thus far include: gamma detectors, beta detectors, video cameras, and pipe locators. Alpha measurement capability is currently under development. A remotely operable Pipe Explorer trademark system has been developed and demonstrated for use in DOE facilities in the decommissioning stage. The system is capable of deployment in pipes as small as 2-inch-diameter and up to 250 feet long. This paper describes the technology and presents measurement results of a field demonstration conducted with the Pipe Explorer trademark system at a DOE site. These measurements identify surface activity levels of U-238 contamination as a function of location in drain lines. Cost savings to the DOE of approximately $1.5 million dollars were realized from this one demonstration

  7. Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

    International Nuclear Information System (INIS)

    Strong, B.R. Jr.; Baschiere, R.J.

    1978-01-01

    The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the 'force equivalent area' (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques. (Auth.)

  8. Inspection of secondary cooling system piping of JMTR

    International Nuclear Information System (INIS)

    Hanawa, Yoshio; Izumo, Hironobu; Fukasaku, Akitomi; Nagao, Yoshiharu; Kawamura, Hiroshi

    2008-06-01

    Piping condition was inspected form the view point of long term utilization before the renewal work of the secondary cooling system in the JMTR on FY 2008. As the result, it was confirmed that cracks, swellings and exfoliations in inner lining of the piping could be observed, and corrosion, which was reached by piping ingot, or decrease of piping thickness could hardly be observed. It was therefore confirmed that the strength or the functionality of the piping had been maintained by usual operation and maintenance. Repair of inner lining of the piping during the refurbishment of the JMTR is necessary to long term utilization of the secondary cooling system after restart of the JMTR from the view point of preventive maintenance. In addition, a periodic inspection of inner lining condition is necessary after repair of the piping. (author)

  9. Comparison and evaluation of flexible and stiff piping systems

    International Nuclear Information System (INIS)

    Hahn, W.; Tang, H.T.; Tang, Y.K.

    1983-01-01

    An experimental and numerical study was performed on a piping system, with various support configurations, to assess the difference in piping response for flexible and stiff piping systems. Questions have arisen concerning a basic design philosophy employed in present day piping designs. One basic question is, the reliability of a flexible piping system greater than that of a stiff piping system by virtue of the fact that a flexible system has fewer snubber supports. With fewer snubbers, the pipe is less susceptible to inadvertent thermal stresses introduced by snubber malfunction during normal operation. In addition to the technical issue, the matter of cost savings in flexible piping system design is a significant one. The costs associated with construction, in-service inspection and maintenance are all significantly reduced by reducing the number of snubber supports. The evaluation study, sponsored by the Electric Power Research Institute, was performed on a boiler feedwater line at Consolidated Edison's Indian Point Unit 1. In this study, the boiler feedwater line was tested and analyzed with two fundamentally different support systems. The first system was very flexible, employing rod and spring hangers, and represented the 'old' design philosophy. The pipe system was very flexible with this support system, due to the long pipe span lengths between supports and the fact that there was only one lateral support. This support did not provide much restraint since it was near an anchor. The second system employed strut and snubber supports and represented the 'modern' design philosophy. The pipe system was relatively stiff with this support system, primarily due to the increased number of supports, including lateral supports, thereby reducing the pipe span lengths between supports. The second support system was designed with removable supports to facilitate interchange of the supports with different support types (i.e., struts, mechanical snubbers and hydraulic

  10. Optimal support arrangement of piping systems using genetic algorithm

    International Nuclear Information System (INIS)

    Chiba, T.; Okado, S.; Fujii, I.; Itami, K.

    1996-01-01

    The support arrangement is one of the important factors in the design of piping systems. Much time is required to decide the arrangement of the supports. The authors applied a genetic algorithm to find the optimum support arrangement for piping systems. Examples are provided to illustrate the effectiveness of the genetic algorithm. Good results are obtained when applying the genetic algorithm to the actual designing of the piping system

  11. Evaluation of material integrity on electricity generator water steam cycles component (Main Steam Pipe)

    International Nuclear Information System (INIS)

    Sudardjo; Histori; Triyadi, Ari

    1998-01-01

    The evaluation of material integrity on electricity generator component has been done. That component was main steam pipe of Unit II Suralaya Coal Fired Power Plant. evaluation was done by replication technique. The damage was found are two porosity's, from two point samples of six points sample population. Based on cavity evaluation in steels, which proposed by Neubauer and Wedel that porosity's still at class A damage. For class A damage, its means no remedial action would be required until next major scheduled maintenance outage. That porosity's was grouped on isolated cavities and not need ti repair that main steam pipe component less than three year after replication test

  12. Seismic evaluation of piping systems using screening criteria

    International Nuclear Information System (INIS)

    Campbell, R.D.; Landers, D.F.; Minichiello, J.C.; Slagis, G.C.; Antaki, G.A.

    1994-01-01

    This document may be used by a qualified review team to identify potential sources of seismically induced failure in a piping system. Failure refers to the inability of a piping system to perform its expected function following an earthquake, as defined in Table 1. The screens may be used alone or with the Seismic Qualification Utility Group -- Generic Implementation Procedure (SQUG-GIP), depending on the piping system's required function, listed in Table 1. Features of a piping system which do not the screening criteria are called outliers. Outliers must either be resolved through further evaluations, or be considered a potential source of seismically induced failure. Outlier evaluations, which do not necessarily require the qualification of a complete piping system by stress analysis, may be based on one or more of the following: simple calculations of pipe spans, search of the test or experience data, vendor data, industry practice, etc

  13. Application of PHADEC method for the decontamination of radioactive steam piping components of Caorso plant

    International Nuclear Information System (INIS)

    Lo Frano, R.; Aquaro, D.; Fontani, E.; Pilo, F.

    2014-01-01

    Highlights: • Application of PHADEC chemical off-line methodology. • Decontamination of radioactive steam piping components of Caorso turbine building. • Experimental characterization of metallic components, e.g., by SEM analysis. • Measure of the efficiency of treatment by means of the reduction of activity and vs. the treatment time. • Minimization of secondary waste produced during decontamination activity of Caorso BWR plant. - Abstract: The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (PHosphoric Acid DEContamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e., taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the

  14. Development of seismic design method for piping system supported by elastoplastic damper. 3. Vibration test of three-dimensional piping model and its response analysis

    International Nuclear Information System (INIS)

    Namita, Yoshio; Kawahata, Jun-ichi; Ichihashi, Ichiro; Fukuda, Toshihiko.

    1995-01-01

    Component and piping systems in current nuclear power plants and chemical plants are designed to employ many supports to maintain safety and reliability against earthquakes. However, these supports are rigid and have a slight energy-dissipating effect. It is well known that applying high-damping supports to the piping system is very effective for reducing the seismic response. In this study, we investigated the design method of the elastoplastic damper [energy absorber (EAB)] and the seismic design method for a piping system supported by the EAB. Our final goal is to develop technology for applying the EAB to the piping system of an actual plant. In this paper, the vibration test results of the three-dimensional piping model are presented. From the test results, it is confirmed that EAB has a large energy-dissipating effect and is effective in reducing the seismic response of the piping system, and that the seismic design method for the piping system, which is the response spectrum mode superposition method using each modal damping and requires iterative calculation of EAB displacement, is applicable for the three-dimensional piping model. (author)

  15. Mechanical Interaction in Pressurized Pipe Systems: Experiments and Numerical Models

    Directory of Open Access Journals (Sweden)

    Mariana Simão

    2015-11-01

    Full Text Available The dynamic interaction between the unsteady flow occurrence and the resulting vibration of the pipe are analyzed based on experiments and numerical models. Waterhammer, structural dynamic and fluid–structure interaction (FSI are the main subjects dealt with in this study. Firstly, a 1D model is developed based on the method of characteristics (MOC using specific damping coefficients for initial components associated with rheological pipe material behavior, structural and fluid deformation, and type of anchored structural supports. Secondly a 3D coupled complex model based on Computational Fluid Dynamics (CFD, using a Finite Element Method (FEM, is also applied to predict and distinguish the FSI events. Herein, a specific hydrodynamic model of viscosity to replicate the operation of a valve was also developed to minimize the number of mesh elements and the complexity of the system. The importance of integrated analysis of fluid–structure interaction, especially in non-rigidity anchored pipe systems, is equally emphasized. The developed models are validated through experimental tests.

  16. Pre- and post-calculations for crack opening and leak rate experiments on piping components within the HDR-program

    International Nuclear Information System (INIS)

    Grebner, H.; Hoefler, A.; Hunger, H.

    1991-01-01

    In this paper calculations to experiments on leak opening and leak rates of piping components are presented. The experiments are performed at the HDR-facility at Karlstein/Germany and up to now straight pipes and pipe branches were considered. Numerical and experimental results are compared. (author)

  17. Pipe Explorer{trademark} surveying system. Innovative technology summary report

    Energy Technology Data Exchange (ETDEWEB)

    1999-06-01

    The US Department of Energy`s (DOE) Chicago Operations Office and the DOE`s Federal Energy Technology Center (FETC) developed a Large Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial decontamination and decommissioning (D and D) technologies in comparison with current baseline technologies. The Pipe Explorer{trademark} system was developed by Science and Engineering Associates, Inc. (SEA), Albuquerque, NM as a deployment method for transporting a variety of survey tools into pipes and ducts. Tools available for use with the system include alpha, beta and gamma radiation detectors; video cameras; and pipe locator beacons. Different versions of this technology have been demonstrated at three other sites; results of these demonstrations are provided in an earlier Innovative Technology Summary Report. As part of a D and D project, characterization radiological contamination inside piping systems is necessary before pipes can be recycled, remediated or disposed. This is usually done manually by surveying over the outside of the piping only, with limited effectiveness and risk of worker exposure. The pipe must be accessible to workers, and embedded pipes in concrete or in the ground would have to be excavated at high cost and risk of exposure to workers. The advantage of the Pipe Explorer is its ability to perform in-situ characterization of pipe internals.

  18. Pipe Explorer surveying system. Innovative technology summary report

    International Nuclear Information System (INIS)

    1999-06-01

    The US Department of Energy's (DOE) Chicago Operations Office and the DOE's Federal Energy Technology Center (FETC) developed a Large Scale Demonstration Project (LSDP) at the Chicago Pile-5 Research Reactor (CP-5) at Argonne National Laboratory-East (ANL). The objective of the LSDP is to demonstrate potentially beneficial decontamination and decommissioning (D and D) technologies in comparison with current baseline technologies. The Pipe Explorer trademark system was developed by Science and Engineering Associates, Inc. (SEA), Albuquerque, NM as a deployment method for transporting a variety of survey tools into pipes and ducts. Tools available for use with the system include alpha, beta and gamma radiation detectors; video cameras; and pipe locator beacons. Different versions of this technology have been demonstrated at three other sites; results of these demonstrations are provided in an earlier Innovative Technology Summary Report. As part of a D and D project, characterization radiological contamination inside piping systems is necessary before pipes can be recycled, remediated or disposed. This is usually done manually by surveying over the outside of the piping only, with limited effectiveness and risk of worker exposure. The pipe must be accessible to workers, and embedded pipes in concrete or in the ground would have to be excavated at high cost and risk of exposure to workers. The advantage of the Pipe Explorer is its ability to perform in-situ characterization of pipe internals

  19. Cryogenic and Gas System Piping Pressure Tests (A Collection of PT Permits)

    International Nuclear Information System (INIS)

    Rucinski, Russell A.

    2002-01-01

    This engineering note is a collection of pipe pressure testing documents for various sections of piping for the D-Zero cryogenic and gas systems. High pressure piping must conform with FESHM chapter 5031.1. Piping lines with ratings greater than 150 psig have a pressure test done before the line is put into service. These tests require the use of pressure testing permits. It is my intent that all pressure piping over which my group has responsibility conforms to the chapter. This includes the liquid argon and liquid helium and liquid nitrogen cryogenic systems. It also includes the high pressure air system, and the high pressure gas piping of the WAMUS and MDT gas systems. This is not an all inclusive compilation of test documentation. Some piping tests have their own engineering note. Other piping section test permits are included in separate safety review documents. So if it isn't here, that doesn't mean that it wasn't tested. D-Zero has a back up air supply system to add reliability to air compressor systems. The system includes high pressure piping which requires a review per FESHM 5031.1. The core system consists of a pressurized tube trailer, supply piping into the building and a pressure reducing regulator tied into the air compressor system discharge piping. Air flows from the trailer if the air compressor discharge pressure drops below the regulator setting. The tube trailer is periodically pumped back up to approximately 2000 psig. A high pressure compressor housed in one of the exterior buildings is used for that purpose. The system was previously documented, tested and reviewed for Run I, except for the recent addition of piping to and from the high pressure compressor. The following documents are provided for review of the system: (1) Instrument air flow schematic, drg. 3740.000-ME-273995 rev. H; (2) Component list for air system; (3) Pressure testing permit for high pressure piping; (4) Documentation from Run I contained in D-Zero Engineering note

  20. Applications of equivalent linearization approaches to nonlinear piping systems

    International Nuclear Information System (INIS)

    Park, Y.; Hofmayer, C.; Chokshi, N.

    1997-01-01

    The piping systems in nuclear power plants, even with conventional snubber supports, are highly complex nonlinear structures under severe earthquake loadings mainly due to various mechanical gaps in support structures. Some type of nonlinear analysis is necessary to accurately predict the piping responses under earthquake loadings. The application of equivalent linearization approaches (ELA) to seismic analyses of nonlinear piping systems is presented. Two types of ELA's are studied; i.e., one based on the response spectrum method and the other based on the linear random vibration theory. The test results of main steam and feedwater piping systems supported by snubbers and energy absorbers are used to evaluate the numerical accuracy and limitations

  1. Dynamic response of piping system subject to flow acoustic excitation

    International Nuclear Information System (INIS)

    Wang, T.; Sun, Y.S.

    1988-01-01

    Through the use of a theoretically derived and test data-calibrated forcing function, the dynamic response of a piping system subject to flow-acoustic induced vibration is analyzed. It is shown that the piping behavior can be predicted when consideration is given to both the wall flexural vibration and the piping system vibration. Piping responded as a system to the transversal excitation due to the swirling motion of the fluid flow, as well as flexurally to the high-frequency acoustic excitations. The transverse piping system response was calculated using a lumped mass piping model. The piping model has more stringent requirements than its counterpart for waterhammer and seismic modeling due to the shorter spiral wavelength and higher frequency of the forcing function. Proper modeling ensured that both the moment stress caused by system excitation and the local stress induced by the support reaction load were properly accounted for. Flexural vibration not only poses a threat to nipples and branch connections, but also contributes substantially to the resultant total stress experienced by the pipe. The forcing function approach has the advantage that the critical locations on the piping system can be identified by means of analysis, facilitating surveillance and inspection, as well as fatigue evaluation

  2. Structural and stress analysis of nuclear piping systems

    International Nuclear Information System (INIS)

    Hata, Hiromichi

    1982-01-01

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

  3. Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

    International Nuclear Information System (INIS)

    Lydell, Bengt; Olsson, Anders

    2008-01-01

    This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning meetings that were

  4. Reliability Data for Piping Components in Nordic Nuclear Power Plants 'R-Book'. Project Phase 1. Rev 1

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, Bengt (Scandpower Risk Management Inc., Houston, TX (US)); Olsson, Anders (Relcon Scandpower AB, Stockholm (SE))

    2008-01-15

    This report constitutes a planning document for a new RandD project to develop a piping component reliability parameter handbook for use in probabilistic safety assessment (PSA) and related activities. The Swedish acronym for this handbook is 'R-Book.' The objective of the project is to utilize the OECD Nuclear Energy Agency 'OECD Pipe Failure Data Exchange Project' (OPDE) database to derive piping component failure rates and rupture probabilities for input to internal flooding probabilistic safety assessment, high-energy line break' (HELB) analysis, risk-informed in-service inspection (RI-ISI) program development, and other activities related to PSA. This new RandD project is funded by member organizations of the Nordic PSA Group (NPSAG) - Forsmark AB, OKG AB, Ringhals AB, and the Swedish Nuclear Power Inspectorate (SKI). The history behind the current effort to produce a handbook of piping reliability parameters goes back to 1994 when SKI funded a 5-year RandD project to explore the viability of establishing an international database on the service experience with piping system components in commercial nuclear power plants. An underlying objective behind this 5-year program was to investigate the different options and possibilities for deriving pipe failure rates and rupture probabilities directly from service experience data as an alternative to probabilistic fracture mechanics. The RandD project culminated in an international piping reliability seminar held in the fall of 1997 in Sigtuna (Sweden) and a pilot project to demonstrate an application of the pipe failure database to the estimation of loss-of-coolant-accident (LOCA) frequency (SKI Report 98:30). A particularly important outcome of the 5-year project was a decision by SKI to transfer the pipe failure database including the lessons learned to an international cooperative effort under the auspices of the OECD Nuclear Energy Agency. Following on information exchange and planning

  5. Investigation of structure in the modular light pipe component for LED automotive lamp

    Science.gov (United States)

    Chen, Hsi-Chao; Zhou, Yang; Huang, Chien-Sheng; Jhong, Wan-Ling; Cheng, Bo-Wei; Jhang, Jhe-Ming

    2014-09-01

    Light-Emitting Diodes (LEDs) have the advantages of small length, long lifetime, fast response time (μs), low voltage, good mechanical properties and environmental protection. Furthermore, LEDs could replace the halogen lamps to avoid the mercury pollution and economize the use of energy. Therefore, the LEDs could instead of the traditional lamp in the future and became an important light source. The proposal of this study was to investigate the effects of the structure and length of the reflector component for a LED automotive lamp. The novel LED automotive lamp was assembled by several different modularization columnar. The optimized design of the different structure and the length to the reflector was simulated by software TracePro. The design result must met the vehicle regulation of United Nations Economic Commission for Europe (UNECE) such as ECE-R19 etc. The structure of the light pipe could be designed by two steps structure. Then constitute the proper structure and choose different power LED to meet the luminous intensity of the vehicle regulation. The simulation result shows the proper structure and length has the best total luminous flux and a high luminous efficiency for the system. Also, the stray light could meet the vehicle regulation of ECE R19. Finally, the experimental result of the selected structure and length of the light pipe could match the simulation result above 80%.

  6. Application of PHADEC method for the decontamination of radioactive steam piping components

    International Nuclear Information System (INIS)

    Lo Frano, R.; Pilo, F.; Aquaro, D.

    2013-01-01

    The dismantling of nuclear plants is a complex activity that originates often a large quantity of radioactive contaminated residue. In this paper the attention was focused on the PHADEC (Phosphoric Acid Decontamination) plant adopted for the clearance of Caorso NPP (in Italy) metallic systems and components contaminated by Co 60 (produced by the neutron capture in the iron materials), like the main steam lines, moisture separator of the turbine buildings, etc.. The PHADEC plant consists in a chemical off line treatment: the crud, deposited along the steam piping during life plant as an example, is removed by means of acid attacks in ponds coupled to a high pressure water washing. Due to the fact that the removed contaminated layers, essentially, iron oxides of various chemical composition, depend on components geometry, type of contamination and time of treatment in the PHADEC plant, it becomes of meaningful importance to suggest a procedure capable to improve the control of the PHADEC process parameters. This study aimed thus at the prediction and optimization of the mentioned treatment time in order to improve the efficiency of the plant itself and to achieve, in turn, the minimization of produced wastes. To the purpose an experimental campaign was carried out by analysing several samples, i.e. taken along the main steam piping line. Smear tests as well as metallographic analyses were carried out in order to determine respectively the radioactivity distribution and the crud composition on the inner surface of the components. Moreover the radioactivity in the crud thickness was measured. These values allowed finally to correlate the residence time in the acid attack ponds to the level of the achieved decontamination. (authors)

  7. Acoustic leak detection in piping systems, 4

    International Nuclear Information System (INIS)

    Kitajima, Akira; Naohara, Nobuyuki; Aihara, Akihiko

    1983-01-01

    To monitor a high-pressure piping of nuclear power plants, a possibility of acoustic leak detection method has been experimentally studied in practical field tests and laboratory tests. Characteristics of background noise in field test and the results of experiment are summarized as follows: (1) The level of background noise in primary loop (PWR) was almost constant under actual plant operation. But it is possible that it rises at the condition of the pressure in primary loop. (2) Based on many experience of laboratory tests and practical field tests. The leak monitoring system for practical field was designed and developed. To improve the reliability, a judgment of leak on this system is used three factors of noise level, duration time of phenomena and frequency spectrum of noise signal emitted from the leak point. (author)

  8. Development of nonlinear dynamic analysis program for nuclear piping systems

    International Nuclear Information System (INIS)

    Kamichika, Ryoichi; Izawa, Masahiro; Yamadera, Masao

    1980-01-01

    In the design for nuclear power piping, pipe-whip protection shall be considered in order to keep the function of safety related system even when postulated piping rupture occurs. This guideline was shown in U.S. Regulatory Guide 1.46 for the first time and has been applied in Japanese nuclear power plants. In order to analyze the dynamic behavior followed by pipe rupture, nonlinear analysis is required for the piping system including restraints which play the role of an energy absorber. REAPPS (Rupture Effective Analysis of Piping Systems) has been developed for this purpose. This program can be applied to general piping systems having branches etc. Pre- and post- processors are prepared in this program in order to easily input the data for the piping engineer and show the results optically by use of a graphic display respectively. The piping designer can easily solve many problems in his daily work by use of this program. This paper describes about the theoretical background and functions of this program and shows some examples. (author)

  9. Static analysis of a piping system with elbows

    International Nuclear Information System (INIS)

    Bryan, B.J.

    1994-01-01

    Vibration tests of elbows to failure were performed in Japan in the early 1970s. The piping system included two elbows and an eccentric mass. Tests were run both pressurized and unpressurized. This report documents a static analysis of the piping system in which the elbows are subjected to out of plane bending. The effects of internal pressure and material plasticity are investigated

  10. Characterization of pipes, drain lines, and ducts using the pipe explorer system

    International Nuclear Information System (INIS)

    Cremer, C.D.; Kendrick, D.T.; Cramer, E.

    1997-01-01

    As DOE dismantles its nuclear processing facilities, site managers must employ the best means of disposing or remediating hundreds of miles of potentially contaminated piping and duct work. Their interiors are difficult to access, and in many cases even the exteriors are inaccessible. Without adequate characterization, it must be assumed that the piping is contaminated, and the disposal cost of buried drain lines can be on the order of $1,200/ft and is often unnecessary as residual contamination levels often are below free release criteria. This paper describes the program to develop a solution to the problem of characterizing radioactive contamination in pipes. The technical approach and results of using the Pipe Explorer trademark system are presented. The heart of the system is SEA's pressurized inverting membrane adapted to transport radiation detectors and other tools into pipes. It offers many benefits over other pipe inspection approaches. It has video and beta/gamma detection capabilities, and the need for alpha detection has been addressed through the development of the Alpha Explorer trademark. These systems have been used during various stages of decontamination and decommissioning of DOE sites, including the ANL CP-5 reactor D ampersand D. Future improvements and extensions of their capabilities are discussed

  11. 2-component heating systems

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, W

    1987-03-01

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

  12. Effect of piping systems on surge in centrifugal compressors

    International Nuclear Information System (INIS)

    Tamaki, Hideaki

    2008-01-01

    There is a possibility that the exchange of the piping system may change the surge characteristic of a compressor. The piping system of a plant is not always the same as that of a test site. Then it is important to evaluate the effect of piping systems on surge characteristics in centrifugal compressors. Several turbochargers combined with different piping systems were tested. The lumped parameter model which was simplified to be solved easily was applied for the prediction of surge point. Surge lines were calculated with the linearlized lumped parameter model. The difference between the test and calculated results was within 10 %. Trajectory of surge cycle was also examined by solving the lumped parameter model. Mild surge and deep surge were successfully predicted. This study confirmed that the lumped parameter model was a very useful tool to predict the effect of piping systems on surge characteristics in centrifugal compressors, even though that was a simple model

  13. Further considerations for damping in heavily insulated pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Lindquist, M.R.; Severud, L.K.

    1985-01-01

    Over the past several years a body of test data has been accumulated which demonstrates that damping in small diameter heavily insulated pipe systems is much larger than presently recommended by Regulatory Guide 1.61. This data is generally based on pipe systems using a stand-off insulation design with a heater annulus. Additional tests have how been completed on similar pipe systems using a strap-on insulation design without the heater annulus. Results indicate some reduction in damping over the stand-off designs. Test data has also been obtained on a larger sixteen-inch diameter heavily insulated pipe system. Results of these two additional test series are presented. Revised damping values for seismic design of heavily insulated pipe systems are then recommended

  14. Further considerations for damping in heavily insulated pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Lindquist, M.R.; Severud, L.K.

    1985-01-01

    Over the past several years a body of test data has been accumulated which demonstrates that damping in small diameter heavily insulated pipe systems is much larger than presently recommended by Regulatory Code 1.61. This data is generally based on pipe systems using a stand-off insulation design with a heater annulus. Additional tests have now been completed on similar pipe systems using a strap-on insulation design without the heater annulus. Results indicate some reduction in damping over the stand-off designs. Test data has also been obtained on a larger sixteen-inch diameter heavily insulated pipe system. Results of these two additional test series are presented. Revised damping values for seismic design for heavily insulated pipe systems are then recommended

  15. Flow induced vibrations of piping system (Vibration sources - Mechanical response of the pipes)

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.; Villard, B.

    1978-01-01

    In order to design the supports of piping system, an estimation of the vibration induced by the fluid conveyed through the pipes are generally needed. For that purpose it is necessary. To evaluate the power spectra of all the main sources generated by the flow. These sources are located at the singular points of the circuit (enlargements, bends, valves, etc. ...). To calculate the modal parameters of fluid containing pipes. This paper presents: a methodical study of the most current singularities. Inter-correlation spectra of local pressure fluctuation downstream from the singularity and correlation spectra of associated acoustical sources have been measured. A theory of noise generation by unsteady flow in internal acoustics has been developed. All these results are very useful for evaluating the source characteristics in most practical pipes. A comparison between the calculation and the results of an experimental test has shown a good agreement

  16. Pipe support for use in a nuclear system

    International Nuclear Information System (INIS)

    Pollono, L.P.; Mello, R.M.

    1976-01-01

    Description is given of a vertical pipe support system. It comprises a tubular pipe support structure having the same inside diameter and the same wall thickness as the pipe, the pipe support structure having a generally triangularly shaped extension formed integral with and extending circumferentially around its outward side, the bottom side of this extension generally forming a ledge; an annular load-bearing insulation formed adjacent to the extension; means for clamping the load-bearing insulation to extension; and means for providing constant vertical support to means for clamping [fr

  17. Laser-GMA Hybrid Pipe Welding System

    Science.gov (United States)

    2007-11-01

    Investigation of varying laser power. The welded pipe is shown, with close -ups of the rootside reinforcement and macro sections...68 Figure 44. Investigation of varying laser stand-off. The welded pipe is shown, along with close -ups of backside...conventional beveled joints. With appropriate joint configuration and preparation, deep keyhole penetration provided by the laser and additional filler

  18. A spatial decision support system for pipe-break susceptibility ...

    African Journals Online (AJOL)

    lying properties. Existing decision support systems available in the field of water distribution system maintenance mainly focus on leak detection and pipe rehabilitation/replacement strategies. These existing systems, however, do not address the ...

  19. Inspection, maintenance, and repair of large pumps and piping systems using advanced robotic tools

    International Nuclear Information System (INIS)

    Lewis, R.K.; Radigan, T.M.

    1998-01-01

    Operating and maintaining large pumps and piping systems can be an expensive proposition. Proper inspections and monitoring can reduce costs. This was difficult in the past, since detailed pump inspections could only be performed by disassembly and many portions of piping systems are buried or covered with insulation. Once these components were disassembled, a majority of the cost was already incurred. At that point, expensive part replacement usually took place whether it was needed or not. With the completion of the Pipe Walkertrademark/LIP System and the planned development of the Submersible Walkertrademark, this situation is due to change. The specifications for these inspection and maintenance robots will ensure that. Their ability to traverse both horizontal and vertical, forward and backward, make them unique tools. They will open the door for some innovative approaches to inspection and maintenance of large pumps and piping systems

  20. Bolted Flanged Connection for Critical Plant/Piping Systems

    International Nuclear Information System (INIS)

    Efremov, Anatoly

    2006-01-01

    A novel type of Bolted Flanged Connection with bolts and gasket manufactured on a basis of advanced Shape Memory Alloys is examined. Presented approach combined with inverse flexion flange design of plant/piping joint reveals a significant increase of internal pressure under conditions of a variety of operating temperatures relating to critical plant/piping systems. (author)

  1. FSI analysis of piping systems under seismic excitation

    International Nuclear Information System (INIS)

    Uras, R.A.; Ma, D.C.; Chang, Yao W.; Liu, Wing Kam

    1991-01-01

    A formulation which accounts for fluid-structure interaction of piping system under seismic excitation is presented. The governing equations of the fluid and the structure to model the pipe are stated. Using the finite element method the discretized equations are obtained. A transformation procedure for proper assembly of matrices is introduced. A solution algorithm is described. 9 refs., 2 figs

  2. Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

    International Nuclear Information System (INIS)

    Lowry, W.

    1994-01-01

    The objective for the development of the Pipe Explorer trademark radiological characterization system is to achieve a cost effective, low risk means of characterizing gamma radioactivity on the inside surface of pipes. The unique feature of this inspection system is the use of a pneumatically inflated impermeable membrane which transports the detector into the pipe as it inverts. The membrane's internal air pressure tows the detector and tether through the pipe. This mechanism isolates the detector and its cabling from the contaminated surface, yet allows measurement of radioactive emissions which can readily penetrate the thin plastic membrane material (such as gamma and high energy beta emissions). In Phase 1, an initial survey of DOE facilities was conducted to determine the physical and radiological characteristics of piping systems. The inverting membrane deployment system was designed and extensively tested in the laboratory. A range of membrane materials was tested to evaluate their ruggedness and deployment characteristics. Two different sizes of gamma scintillation detectors were procured and tested with calibrated sources. Radiation transport modeling evaluated the measurement system's sensitivity to detector position relative to the contaminated surface, the distribution of the contamination, background gamma levels, and gamma source energy levels. In the culmination of Phase 1, a field demonstration was conducted at the Idaho National Engineering Laboratory's Idaho Chemical Processing Plant. The project is currently in transition from Phase 1 to Phase 2, where more extensive demonstrations will occur at several sites. Results to date are discussed

  3. Assessment and management of ageing of major nuclear power plant components important to safety. Primary piping in PWRs

    International Nuclear Information System (INIS)

    2003-07-01

    guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. This report addresses the primary piping in PWRs including main coolant piping, surge and spray lines, Class 1 piping in attached systems, and small diameter piping that cannot be isolated from the primary coolant system. Maintaining the structural integrity of this piping throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

  4. Integrated model of port oil piping transportation system safety including operating environment threats

    OpenAIRE

    Kołowrocki, Krzysztof; Kuligowska, Ewa; Soszyńska-Budny, Joanna

    2017-01-01

    The paper presents an integrated general model of complex technical system, linking its multistate safety model and the model of its operation process including operating environment threats and considering variable at different operation states its safety structures and its components safety parameters. Under the assumption that the system has exponential safety function, the safety characteristics of the port oil piping transportation system are determined.

  5. Structural analysis program of plant piping system. Introduction of AutoPIPE V8i new feature. JSME PPC-class 2 piping code

    International Nuclear Information System (INIS)

    Motohashi, Kazuhiko

    2009-01-01

    After an integration with ADLPipe, AutoPIPE V8i (ver.9.1) became the structural analysis program of plant piping system featured with analysis capability for the ASME NB Class 1 and JSME PPC-Class 2 piping codes including ASME NC Class 2 and ASME ND Class 3. This article described analysis capability for the JSME PPC-Class 2 piping code as well as new general features such as static analysis up to 100 thermal, 10 seismic and 10 wind load cases including different loading scenarios and pipe segment edit function: join, split, reverse and re-order segments. (T. Tanaka)

  6. Laser-GMA Hybrid Pipe Welding System

    National Research Council Canada - National Science Library

    Reutzel, Edward W; Kern, Ludwig; Sullivan, Michael J; Tressler, Jay F; Avalos, Juan

    2007-01-01

    The combination of laser welding with conventional gas metal arc welding technology offers substantial increases in production rate of joining pipe through single-pass joining compared to multi-pass...

  7. Seismic margins and calibration of piping systems

    International Nuclear Information System (INIS)

    Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

    1985-01-01

    The Seismic Safety Margins Research Program (SSMRP) is a US Nuclear Regulatory Commission-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its objective is to develop a complete, fully coupled analysis procedure for estimating the risk of earthquake-induced radioactive release from a commercial nuclear power plant and to determine major contributors to the state-of-the-art seismic and systems analysis process and explicitly includes the uncertainties in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I of SSMRP, the overall seismic risk assessment methodology was developed and assembled. The application of this methodology to the seismic PRA (Probabilistic Risk Assessment) at the Zion Nuclear Power Plant has been documented. This report documents the method deriving response factors. The response factors, which relate design calculated responses to best estimate values, were used in the seismic response determination of piping systems for a simplified seismic probablistic risk assessment. 13 references, 31 figures, 25 tables

  8. Service Life Of Main Piping Component Due To Low Thermal Stresses.Fatigue

    International Nuclear Information System (INIS)

    Miroshnik, R.; Jeager, A.; Ben Haim, H.

    1998-01-01

    The paper deals with estimating the service life of the power station Main piping component and describing the repair process for extending of its service life. After a long period of service, several circular fatigue cracks have been discovered at the bottom of the Main piping component chamber. Finite element analyses of transient thermal stresses, caused by power station startup, are carried out in the paper. The calculation results show good agreement between the theoretical locations of the maximum stresses and the actual locations of the cracks. There is a good agreement between theoretical evaluation and actual service life, as well. The possibility of machining out the cracks in order to prevent their growing is examined here. The machining enables us to extend the power station component's life service

  9. Heat pipe as a cooling mechanism in an aeroponic system

    Energy Technology Data Exchange (ETDEWEB)

    Srihajong, N.; Terdtoon, P.; Kamonpet, P. [Department of Mechanical Engineering, Faculty of Engineering, Chiang Mai University, Chiang Mai 50200 (Thailand); Ruamrungsri, S. [Department of Horticulture, Faculty of Agriculture, Chiang Mai University, Chiang Mai 50200 (Thailand); Ohyama, T. [Department of Applied Biological Chemistry, Faculty of Agriculture, Niigata University (Japan)

    2006-02-01

    This paper presents an establishment of a mathematical model explaining the operation of an aeroponic system for agricultural products. The purpose is to study the rate of energy consumption in a conventional aeroponic system and the feasibility of employing a heat pipe as an energy saver in such a system. A heat pipe can be theoretically employed to remove heat from the liquid nutrient that flows through the growing chamber of an aeroponic system. When the evaporator of the heat pipe receives heat from the nutrient, the inside working fluid evaporates into vapor and flows to condense at the condenser section. The outlet temperature of the nutrient from the evaporator section is, therefore, decreased by the heat removal mechanism. The heat pipe can also be used to remove heat from the greenhouse by applying it on the greenhouse wall. By doing this, the nutrient temperature before entering into the nutrient tank decreases and the cooling load of evaporative cooling will subsequently be decreased. To justify the heat pipe application as an energy saver, numerical computations have been done on typical days in the month of April from which maximum heating load occurs and an appropriate heat pipe set was theoretically designed. It can be seen from the simulation that the heat pipe can reduce the electric energy consumption of an evaporative cooling and a refrigeration systems in a day by 17.19% and 10.34% respectively. (author)

  10. Alternate procedures for the seismic analysis of multiply supported piping systems

    International Nuclear Information System (INIS)

    Subudhi, M.; Bezler, P.

    1985-01-01

    The seismic design of secondary systems such as piping requires knowledge of the motions at various locations of the primary structures. When the structure or buildings are subjected to earthquake-like excitations at the ground level, the responses at different floor levels may be quite different from each other. This difference depends on the building and soil frequency characteristics, the characteristics of the input signals, the damping levels, and soil-structure interaction effects. When multiple independent excitations are considered in the analysis of piping systems, the responses can be considered to have two distinct components. One is due to the inertia of masses alone (dynamic component) and the other is due to the time varying differential motion of the support points (pseudo-static component). To address this problem, a sample of six piping systems, two of which were subjected to thirty-three earthquakes, were studied to develop a statistical assessment of different methods of predicting the dynamic, pseudo-static and combined response. Both uniform and independent support motion methods were considered. The results are obtained in tabular form. The mean and standard deviation for the two piping systems subjected to thirty-three earthquakes were obtained to allow an assessment of the adequacy and level of conservatism associated with each method. These results are also displayed in graphical form for selected, critical locations in the piping systems. The limitations of each method and recommendations are discussed

  11. Small Bore Piping Socket Weld Evaluation System

    International Nuclear Information System (INIS)

    Lee, Dong Min; Cho, Hong Seok; Choi, Sang Hoon; Cho, Ki Hyun; Lee, Jang Wook

    2009-01-01

    Kori unit 3 had stopped operation due to leakage at Steam Generator drain line socket weld on June 6, 2008. The Cause of socket weld damage was known as a fatigue crack. According to this case, all socket welds located in RCS pressure boundary are carrying out Radiographic Testing. But to inspect socket welds by RT has some problems. The result of EPRI study showed that RT has limitation to find flaws at socket welds.The orientation of flaws has big influence on RT inspection capability and there is not enough space at socket welds for RT, dose problems as well. Although the gap between coupling and pipe at socket welds must follow up code, surface inspection can't inspect the gap. If there is absence of the gap, socket welds are damaged during operation. The gap should be identified by RT but the distance of gap can't be measured. As this paper, the ultrasonic inspection system was introduced to figure out indication and gap in the socket welds

  12. Liquid hydrogen transfer pipes and level regulation systems

    International Nuclear Information System (INIS)

    Marquet, M.; Prugne, P.; Roubeau, P.

    1961-01-01

    Describes: 1) Transfer pipes - Plunging rods in liquid hydrogen Dewars; transfer pipes: knee-joint system for quick and accurate positioning of plunging Dewar rods; system's rods: combined valve and rod; valves are activated either by a bulb pressure or by a solenoid automatically or hand controlled. The latter allows intermittent filling. 2) Level regulating systems: Level bulbs: accurate to 1 or 4 m; maximum and minimum level bulbs: automatic control of the liquid hydrogen valve. (author) [fr

  13. Radiation detector system having heat pipe based cooling

    Science.gov (United States)

    Iwanczyk, Jan S.; Saveliev, Valeri D.; Barkan, Shaul

    2006-10-31

    A radiation detector system having a heat pipe based cooling. The radiation detector system includes a radiation detector thermally coupled to a thermo electric cooler (TEC). The TEC cools down the radiation detector, whereby heat is generated by the TEC. A heat removal device dissipates the heat generated by the TEC to surrounding environment. A heat pipe has a first end thermally coupled to the TEC to receive the heat generated by the TEC, and a second end thermally coupled to the heat removal device. The heat pipe transfers the heat generated by the TEC from the first end to the second end to be removed by the heat removal device.

  14. Analysis of a piping system under seismic load using incremental hinge technique

    International Nuclear Information System (INIS)

    Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Singh, R.K.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.; Ramesh Babu, R.

    2008-01-01

    ASME Boiler and Pressure Vessel Code treats piping system as a series of components but not as an overall structural system. Limit analyses and collapse tests at component level are used to establish stress allowables on seismic stresses. The code does not consider the load redistributions and structural redundancy existing in piping systems that prevent system collapse even when one or more individual components loaded beyond their collapse levels. This necessitates a simple analytical method for evaluation of inelastic seismic response at system level. The present paper presents a simplified analytical procedure for predicting inelastic response of a typical piping system subjected to seismic load. The analytical method known as incremental hinge technique is based on plastic system behavior in which the yielded components are replaced with hinge models when a critical hinge moment is reached. It also takes into account the inelastic response spectrum reduction factors and displacement ductility. The analytical method is used to obtain the inelastic response, location of hinge formation and level of base excitation needed for hinge formation. The predicted hinge locations and hinge ordering is compared with the results of a shake table test conducted on the piping system. (author)

  15. Study of a risk-based piping inspection guideline system.

    Science.gov (United States)

    Tien, Shiaw-Wen; Hwang, Wen-Tsung; Tsai, Chih-Hung

    2007-02-01

    A risk-based inspection system and a piping inspection guideline model were developed in this study. The research procedure consists of two parts--the building of a risk-based inspection model for piping and the construction of a risk-based piping inspection guideline model. Field visits at the plant were conducted to develop the risk-based inspection and strategic analysis system. A knowledge-based model had been built in accordance with international standards and local government regulations, and the rational unified process was applied for reducing the discrepancy in the development of the models. The models had been designed to analyze damage factors, damage models, and potential damage positions of piping in the petrochemical plants. The purpose of this study was to provide inspection-related personnel with the optimal planning tools for piping inspections, hence, to enable effective predictions of potential piping risks and to enhance the better degree of safety in plant operations that the petrochemical industries can be expected to achieve. A risk analysis was conducted on the piping system of a petrochemical plant. The outcome indicated that most of the risks resulted from a small number of pipelines.

  16. Comparison of elastic and inelastic seismic response of high temperature piping systems

    International Nuclear Information System (INIS)

    Thomas, F.M.; McCabe, S.L.; Liu, Y.

    1994-01-01

    A study of high temperature power piping systems is presented. The response of the piping systems is determined when subjected to seismic disturbances. Two piping systems are presented, a main steam line, and a cold reheat line. Each of the piping systems are modeled using the ANSYS computer program and two analyses are performed on each piping system. First, each piping system is subjected to a seismic disturbance and the pipe material is assumed to remain linear and elastic. Next the analysis is repeated for each piping system when the pipe material is modeled as having elastic-plastic behavior. The results of the linear elastic analysis and elastic-plastic analysis are compared for each of the two pipe models. The pipe stresses, strains, and displacements, are compared. These comparisons are made so that the effect of the material yielding can be determined and to access what error is made when a linear analysis is performed on a system that yields

  17. Non-metallic structural wrap systems for pipe

    International Nuclear Information System (INIS)

    Walker, R.H.; Wesley Rowley, C.

    2001-01-01

    The use of thermoplastics and reinforcing fiber has been a long-term application of non-metallic material for structural applications. With the advent of specialized epoxies and carbon reinforcing fiber, structural strength approaching and surpassing steel has been used in a wide variety of applications, including nuclear power plants. One of those applications is a NSWS for pipe and other structural members. The NSWS is system of integrating epoxies with reinforcing fiber in a wrapped geometrical configuration. This paper specifically addresses the repair of degraded pipe in heat removal systems used in nuclear power plants, which is typically caused by corrosion, erosion, or abrasion. Loss of structural material leads to leaks, which can be arrested by a NSWS for the pipe. The technical aspects of using thermoplastics to structurally improve degraded pipe in nuclear power plants has been addressed in the ASME B and PV Code Case N-589. Using the fundamentals described in that Code Case, this paper shows how this technology can be extended to pipe repair from the outside. This NSWS has already been used extensively in non-nuclear applications and in one nuclear application. The cost to apply this NSWS is typically substantially less than replacing the pipe and may be technically superior to replacing the pipe. (author)

  18. Application of the results of pipe stress analyses into fracture mechanics defect analyses for welds of nuclear piping components; Uebernahme der Ergebnisse von Rohrsystemanalysen (Spannungsanalysen) fuer bruchmechanische Fehlerbewertungen fuer Schweissnaehte an Rohrleitungsbauteilen in kerntechnischen Anlagen

    Energy Technology Data Exchange (ETDEWEB)

    Dittmar, S.; Neubrech, G.E.; Wernicke, R. [TUeV Nord SysTec GmbH und Co.KG (Germany); Rieck, D. [IGN Ingenieurgesellschaft Nord mbH und Co.KG (Germany)

    2008-07-01

    For the fracture mechanical assessment of postulated or detected crack-like defects in welds of piping systems it is necessary to know the stresses in the un-cracked component normal to the crack plane. Results of piping stress analyses may be used if these are evaluated for the locations of the welds in the piping system. Using stress enhancing factors (stress indices, stress factors) the needed stress components are calculated from the component specific sectional loads (forces and moments). For this procedure the tabulated stress enhancing factors, given in the standards (ASME Code, German KTA regulations) for determination and limitation of the effective stresses, are not always and immediately adequate for the calculation of the stress component normal to the crack plane. The contribution shows fundamental possibilities and validity limits for adoption of the results of piping system analyses for the fracture mechanical evaluation of axial and circumferential defects in welded joints, with special emphasis on typical piping system components (straight pipe, elbow, pipe fitting, T-joint). The lecture is supposed to contribute to the standardization of a code compliant and task-related use of the piping system analysis results for fracture mechanical failure assessment. [German] Fuer die bruchmechanische Bewertung von postulierten oder bei der wiederkehrenden zerstoerungsfreien Pruefung detektierten rissartigen Fehlern in Schweissnaehten von Rohrsystemen werden die Spannungen in der ungerissenen Bauteilwand senkrecht zur Rissebene benoetigt. Hierfuer koennen die Ergebnisse von Rohrsystemanalysen (Spannungsanalysen) genutzt werden, wenn sie fuer die Orte der Schweissnaehte im Rohrsystem ausgewertet werden. Mit Hilfe von Spannungserhoehungsfaktoren (Spannungsindizes, Spannungsbeiwerten) werden aus den komponentenweise berechneten Schnittlasten (Kraefte und Momente) die benoetigten Spannungskomponenten berechnet. Dabei sind jedoch die in den Regelwerken (ASME

  19. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis. (orig./GL)

  20. Electronic components and systems

    CERN Document Server

    Dennis, W H

    2013-01-01

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

  1. Aeroacoustics of pipe systems with closed branches

    NARCIS (Netherlands)

    Tonon, D.; Hirschberg, A.; Golliard, J.; Ziada, S.

    2011-01-01

    Flow induced pulsations in resonant pipe networks with closed branches are considered in this review paper. These pulsations, observed in many technical applications, have been identified as self-sustained aeroacoustic oscillations driven by the instability of the flow along the closed branches. The

  2. Fatigue analysis of HANARO primary cooling system piping

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo

    1998-05-01

    A main form of piping failure which occurring leak before break (LBB) is fatigue failure. The fatigue analysis of HANARO primary cooling system (PCS) piping was performed. The PCS piping had been designed in accordance with ASME Class 3 for service conditions. However fatigue analysis is not required in Class 3. In this study the quantitative fatigue analysis was carried out according to ASME Class 1. The highest stress points which have the largest possibility of ASME class 1. The highest stress points which have the largest possibility of the fatigue were determined from the piping stress analysis for each subsection piping. The fatigue analysis was performed for 3 highest stress points, i.e., branch connection, anchor point and butt welding joint. After calculating the peak stress intensity range the fatigue usage factors were evaluated considering operating cycles and S-N curve. The cumulative usage factors for 3 highest stress points were much less than 1. The results show that the possibility of fatigue failure for PCS piping subjected to thermal expansion and seismic loads is very small. The structural integrity of the HANARO PCS piping for fatigue failure was proved to apply the LBB. (author). 11 tabs., 6 figs

  3. Autogenous Metallic Pipe Leak Repair in Potable Water Systems.

    Science.gov (United States)

    Tang, Min; Triantafyllidou, Simoni; Edwards, Marc A

    2015-07-21

    Copper and iron pipes have a remarkable capability for autogenous repair (self-repair) of leaks in potable water systems. Field studies revealed exemplars that metallic pipe leaks caused by nails, rocks, and erosion corrosion autogenously repaired, as confirmed in the laboratory experiments. This work demonstrated that 100% (N = 26) of 150 μm leaks contacting representative bulk potable water in copper pipes sealed autogenously via formation of corrosion precipitates at 20-40 psi, pH 3.0-11.0, and with upward and downward leak orientations. Similar leaks in carbon steel pipes at 20 psi self-repaired at pH 5.5 and 8.5, but two leaks did not self-repair permanently at pH 11.0 suggesting that water chemistry may control the durability of materials that seal the leaks and therefore the permanence of repair. Larger 400 μm holes in copper pipes had much lower (0-33%) success of self-repair at pH 3.0-11.0, whereas all 400 μm holes in carbon steel pipes at 20 psi self-repaired at pH 4.0-11.0. Pressure tests indicated that some of the repairs created at 20-40 psi ambient pressure could withstand more than 100 psi without failure. Autogenous repair has implications for understanding patterns of pipe failures, extending the lifetime of decaying infrastructure, and developing new plumbing materials.

  4. Optimization of a pump-pipe system by dynamic programming

    DEFF Research Database (Denmark)

    Vidal, Rene Victor Valqui; Ferreira, Jose S.

    1984-01-01

    In this paper the problem of minimizing the total cost of a pump-pipe system in series is considered. The route of the pipeline and the number of pumping stations are known. The optimization will then consist in determining the control variables, diameter and thickness of the pipe and the size of...... of the pumps. A general mathematical model is formulated and Dynamic Programming is used to find an optimal solution....

  5. Analytical considerations in the code qualification of piping systems

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1995-01-01

    The paper addresses several analytical topics in the design and qualification of piping systems which have a direct bearing on the prediction of stresses in the pipe and hence on the application of the equations of NB, NC and ND-3600 of the ASME Boiler and Pressure Vessel Code. For each of the analytical topics, the paper summarizes the current code requirements, if any, and the industry practice

  6. Commercial high efficiency dehumidification systems using heat pipes

    Energy Technology Data Exchange (ETDEWEB)

    1993-09-01

    An improved heat pipe design using separately connected two-section one-way flow heat pipes with internal microgrooves instead of wicks is described. This design is now commercially available for use to increase the dehumidification capacity of air conditioning systems. The design also includes a method of introducing fresh air into buildings while recovering heat and controlling the humidity of the incoming air. Included are applications and case studies, load calculations and technical data, and installation, operation, and maintenance information.

  7. Some aspects of the dynamic analysis of piping systems

    International Nuclear Information System (INIS)

    Galeao, A.C.N.R.

    1981-04-01

    Some aspects of vibration and dynamic response of piping systems are presented. The following subjects were analysed: sources of dynamic excitation; steady-state response-periodic excitation; resonance; flow induced vibrations; transient response - seismic excitations; non-linear transient response - pipe - whip. For each of these topics, the mathematical models, the governing equations and the approximate methods of solution, showing some numerical results obtained from the literature. (Author) [pt

  8. BOA II: pipe-asbestos insulation removal system

    International Nuclear Information System (INIS)

    Schempf, H.; Mutschler; Boehmke, S.; Chemel, B.; Piepgras, C.

    1996-01-01

    BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to high labor costs and high level of radioactive contamination, making manual removal costly and inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee

  9. In-situ rehabilitation cleans, lines, and renews pipe systems

    International Nuclear Information System (INIS)

    Munden, B.A.

    1990-01-01

    This article discusses how, in the past five years, developments in coating and lining material technology have found their way into pipe line application and have yielded successful results. The thick film, high solids material often used to repair tanks, vessels and offshore structures has now been adapted for existing pipe lines. One of the most promising of these systems in successful service is an epoxy, high solids (95%) material originally developed for nuclear service as a lining for reactor containment vessels

  10. Characterisation of girth pipe weld for primary heat transport system of pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Singh, P.K.; Vaze, K.K.; Kushwaha, H.S.

    2002-01-01

    The weld and heat affected zone (HAZ) associated with the girth weld are most vulnerable regions of the piping system. The different regions of the weld joint such as the weld metal, HAZ and base metal lead to heterogeneous mechanical and metallurgical properties of the joints. Due to their different metallurgical and mechanical properties, the amounts of damage produced in these regions are different when the component is subjected to service condition. Thus, it is imperative to know the characteristics of these regions of a pipe weld in order to identify the weakest zone for safe designing of high energy piping components. In view of this necessity the present study has been planned to carry out complete characterisation of the weld joint of SA 333 Gr.6 steel pipe, in terms of its metallurgical, mechanical and fracture properties. The mechanical and fracture mechanics properties of the base metal, weld deposit and HAZ have been compared and correlated with reference to their microstructures. Weld joints of SA 333 Gr.6 steel pipe have been prepared by using GTAW root pass and SMAW filling of V-grove as per recommended welding procedure specifications (WPS) conforming to ASME Sec IX commonly used to fabricate nuclear piping system components. The emphasis of the study is to characterise base, weld and HAZ of the pipe weld in terms of chemical, metallurgical, mechanical and fracture mechanics properties. The fracture toughness behaviour of the welds and HAZ has been characterised by J-integral parameters. The fatigue crack growth rate has been characterised by Paris Law. Stretched zone width (SZW) has been measured under SEM to evaluate initiation fracture toughness. The estimated initiation fracture toughness based on SZW and blunting line given by EGF recommendation have been compared. The fracture mechanics properties of base, weld and HAZ has been determined and compared. The fracture mechanics properties of the weld and HAZ have been correlated to their

  11. Thermal-hydraulic analysis of the improved TOPAZ-II power system using a heat pipe radiator

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Wenwen; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

    2016-10-15

    Highlights: • The system thermal-hydraulic model of the improved space thermionic reactor is developed. • The temperature reactivity feedback effects of the moderator, UO2 fuel, electrodes and reflector are considered. • The alkali metal heat pipe radiator is modeled with the two dimensional heat pipe model. • The steady state and the start-up procedure of the system are analyzed. - Abstract: A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process.

  12. High performance passive solar heating system with heat pipe energy transfer and latent heat storage

    NARCIS (Netherlands)

    Dijk, van H.A.L.; Galen, van E; Hensen, J.L.M.; Wit, de M.H.

    1983-01-01

    Preliminar results are reported from a current project on the development of a high performance passive solar heating system. Two special components are introduced: a. A heat pipe as a thermal diode tube for the efficient transfer of collected solar heat from the absorber plate to behind an

  13. Piping data bank and erection system of Angra 2: structure, computational resources and systems

    International Nuclear Information System (INIS)

    Abud, P.R.; Court, E.G.; Rosette, A.C.

    1992-01-01

    The Piping Data Bank of Angra 2 called - Erection Management System - Was developed to manage the piping erection of the Nuclear Power Plant of Angra 2. Beyond the erection follow-up of piping and supports, it manages: the piping design, the material procurement, the flow of the fabrication documents, testing of welds and material stocks at the Warehouse. The works developed in the sense of defining the structure of the Data Bank, Computational Resources and System are here described. (author)

  14. Pipe rupture hardware minimization in pressurized water reactor system

    International Nuclear Information System (INIS)

    Mukherjee, S.K.; Szyslowski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.; Quinones, D.; Server, W.

    1987-01-01

    For much of the high energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station - Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in (152 mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination

  15. Field experience with a novel pipe protection and monitoring system for large offshore pipeline construction projects

    Energy Technology Data Exchange (ETDEWEB)

    Magerstaedt, Michael; Blitz, Gunther [ROSEN Swiss AG, Stans (Switzerland); Sabido, Carlos E. [ROSEN Technology and Research Center, Lingen (Germany)

    2012-07-01

    and, transmit this information in real time via one of the largest self-assembling wireless sensor networks ever built to a central control room thousands of km away. The first of 2 strings of 48 inches is in operation since 2011; the second one will come on stream this year. The Protection system was designed to be reused for the second pipe string. After the second use, the pipe caps were disassembled, the electronics components removed, and the high-performance polyurethane material (altogether thousands of tons) was recycled in a novel process, thus leaving the project with the smallest possible ecological footprint. The recycled material (> 95% recycling content) is used in applications where moving objects have to be protected from high abrasion and wear. Examples will be given in this presentation. The customer saved a very large amount of money compared to conventional storage methods that would require pipe cleaning and inspection before load-out. (author)

  16. Pipe line systems in nuclear power plant

    International Nuclear Information System (INIS)

    Sasada, Yasuhiro; Tanno, Kazuo; Shibato, Eizo.

    1979-01-01

    Purpose: To prevent stress corrosion cracks, in particular, for branched pipeways by conducting water quality control in the branched pipeways as well as in the main pipeways, and reducing the thermal stress in the branched pipeways. Constitution: A water quality monitoring device is provided to a drain pipe and a failed element detection pipe to monitor the quality of stagnated water continuously or periodically. If the impurity concentration or oxygen concentration exceeds a specified value in the stagnated water, a drain valve or a check valve is opened by a signal from the water quality monitoring device to replace the stagnated water with recycling water in the main pipeway. The temperature for the branched loop pipeway and the main pipeway are collectively kept to a same temperature to thereby reduce the thermal stress in the branched pipeway. (Kawakami, Y.)

  17. Validation of the dynamic structural integrity of a nuclear piping component using static inelastic modelling technique

    International Nuclear Information System (INIS)

    Leonard, J.W.

    1975-01-01

    This work is concerned with the evaluation of a quasi-static method as applied to a swing check valve designed to provide emergency shut-off capability subsequent to a postulated break in a steam line. The impact analysis of swinging disk upon the valve seat is an asymmetric problem in dynamic elastoplasticity with potentially large displacements and strains resulting from the impact. To perform a quasi-static analysis for this component the disk and seat region of the valve was isolated from the piping system by special boundary elements and an elastic-plastic finite element model was generated assuming axisymmetric solid ring elements. An equivalent static axisymmetric incremental load system was used to approximate the nonsymmetric initial velocity of impact. Subsequent to the nonlinear incremental finite element analysis by a standard computer software package (MARC-CDC program), a special post-processing program was employed to calculate the incremental sum of external work due to the defined load system. Equating this external work to the initial kinetic energy of impact, parametric curves for displacements, stresses, and strains were obtained as functions of various levels of kinetic energy imparted to the valve at closure. To verify the conservative nature of the assumptions made in the quasi-static model, a comparison was made with a time-dependent, nonlinear, axisymmetric, elastic-plastic finite difference simulation. Another standard computer software package (PISCES-2DL) was used for this dynamic simulation. For a check-point value of initial impact kinetic energy, correlation between the quasi-static finite element and dynamic finite difference analyses is presented. Validations of the assumptions made in the quasi-static analysis and of the results obtained are discussed in detail

  18. Flow induced vibrations in a PWR piping system

    International Nuclear Information System (INIS)

    Seligmann, D.C.; Guillou, J.P.

    1995-01-01

    In this paper, we present and industrial study of the dynamic behaviour of the piping system of a French 1300 M We nuclear power plant. High-amplitude vibrations had been noticed on a safeguard system during the periodical operation startup tests. These vibrations, due to acoustical pump sources, cause fatigue-damage and it is therefore necessary to propose an estimation of the service-life of the piping and to propose modification of piping system to reduce vibrations. First, we define a mechanical model readjusted according to gauged vibratory speeds and construct a vibro-acoustic coupled model and a pump-behaviour model as a source of excitation. Second, we simulate a modification of the supports. The influence of this modification is analysed by comparison of the root mean square values of vibratory speeds and the stresses between the initial system and the modified system. 3 refs., 7 figs

  19. Pipe stress analysis on HCCR-TBS ancillary systems in conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Eo Hwak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    Highlights: • Pipe stress is performed on Korean HCCR-TBS for the load combinations including seismic events. • The resultant stress meets the requirement of the design code & standard except one position where modification is needed. • The results gives useful information for the design evolution in the next desgin phase. - Abstract: Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be tested in ITER to demonstrate feasibility of the breeding blanket concept. The HCCR-TBS comprises Test Blanket Module (TBM) with associated shield, and ancillary systems located in various positions of ITER building. Currently, conceptual design for the HCCR-TBS is in progress. This paper presents pipe stress analysis results for the HCCR-TBS ancillary systems. The pipe stress analysis was performed in accordance with ASME B31.3 for major pipes of the Helium Cooling System (HCS) and the Coolant Purification System (CPS), which are operated in high pressure and temperature. The pipe stress for various load cases and load combinations were calculated. Operational pressure and temperature during plasma operation are applied as pressure load and thermal load, respectively. In addition seismic events were combined to investigate the code compliance for sustained load case and occasional load case. It was confirmed that the resultant stress meets the requirements of ASME B31.3 except one position in which it needs modification. These results give useful information for the next design phase, for example, nozzle loads for the component selection, the support design parameters, etc.

  20. Pipe stress analysis on HCCR-TBS ancillary systems in conceptual design

    International Nuclear Information System (INIS)

    Ahn, Mu-Young; Cho, Seungyon; Lee, Eo Hwak; Park, Yi-Hyun; Lee, Youngmin

    2016-01-01

    Highlights: • Pipe stress is performed on Korean HCCR-TBS for the load combinations including seismic events. • The resultant stress meets the requirement of the design code & standard except one position where modification is needed. • The results gives useful information for the design evolution in the next desgin phase. - Abstract: Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be tested in ITER to demonstrate feasibility of the breeding blanket concept. The HCCR-TBS comprises Test Blanket Module (TBM) with associated shield, and ancillary systems located in various positions of ITER building. Currently, conceptual design for the HCCR-TBS is in progress. This paper presents pipe stress analysis results for the HCCR-TBS ancillary systems. The pipe stress analysis was performed in accordance with ASME B31.3 for major pipes of the Helium Cooling System (HCS) and the Coolant Purification System (CPS), which are operated in high pressure and temperature. The pipe stress for various load cases and load combinations were calculated. Operational pressure and temperature during plasma operation are applied as pressure load and thermal load, respectively. In addition seismic events were combined to investigate the code compliance for sustained load case and occasional load case. It was confirmed that the resultant stress meets the requirements of ASME B31.3 except one position in which it needs modification. These results give useful information for the next design phase, for example, nozzle loads for the component selection, the support design parameters, etc.

  1. Mechanized ultrasonic inspection of austenitic pipe systems; Mechanisierte Ultraschallpruefung von austenitischen Rohrleitungen

    Energy Technology Data Exchange (ETDEWEB)

    Dressler, K.; Luecking, J.; Medenbach, S. [ABB ZAQ GmbH, Essen (Germany)

    1999-08-01

    The contribution explains the system of standard testing methods elaborated by ABB ZAQ GmbH for inspection of austenitic plant components. The inspection tasks explained in greater detail are basic materials testing (straight pipes, bends, and pipe specials), and inspection of welds and dissimilar welds. The techniques discussed in detail are those for detection and sizing of defects. (orig./CB) [Deutsch] Das Ziel dieses Beitrages ist die Vorstellung der von der ABB ZAQ GmbH eingesetzten Standardprueftechniken fuer die Pruefung austenitischer Anlagenkomponenten. Im einzelnen wird die Grundwerkstoffpruefung (Rohre, Boegen, Formstuecke), die Schweissnahtpruefung und die Mischnahtpruefung angesprochen. Es werden dabei die Techniken fuer `Detection` und `Sizing` differenziert betrachtet und erlaeutert. (orig.)

  2. Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

    International Nuclear Information System (INIS)

    Yoder, Graydon L. Jr.; Harvey, Karen; Ferrada, Juan J.

    2011-01-01

    A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

  3. Seismic fragility test of a 6-inch diameter pipe system

    International Nuclear Information System (INIS)

    Chen, W.P.; Onesto, A.T.; DeVita, V.

    1987-02-01

    This report contains the test results and assessments of seismic fragility tests performed on a 6-inch diameter piping system. The test was funded by the US Nuclear Regulatory Commission (NRC) and conducted by ETEC. The objective of the test was to investigate the ability of a representative nuclear piping system to withstand high level dynamic seismic and other loadings. Levels of loadings achieved during seismic testing were 20 to 30 times larger than normal elastic design evaluations to ASME Level D limits would permit. Based on failure data obtained during seismic and other dynamic testing, it was concluded that nuclear piping systems are inherently able to withstand much larger dynamic seismic loadings than permitted by current design practice criteria or predicted by the probabilistic risk assessment (PRA) methods and several proposed nonlinear methods of failure analysis

  4. Multi-mode vibration control of piping system

    International Nuclear Information System (INIS)

    Minowa, Takeshi; Seto, Kazuto; Iiyama, Fumiya; Sodeyama, Hiroshi

    1999-01-01

    In this paper, dual dynamic absorbers are applied to the piping system in order to control the multiple vibration modes. ANSYS, which is one of the software based on FEM(finite element method), is used for the design of dual dynamic absorbers as well as for the determination of their optimum installing positions. The dual dynamic absorbers designed optimally for controlling the first three vibration modes perform just like a houde damper in higher frequency and have an effect on controlling higher modes. To use this advantage, three dual dynamic absorbers are installed in positions where they influence higher modes, and not only the first three modes of the piping system but also the extensive modes are controlled. Practical experimental study has also been carried out and it is shown that a dual dynamic absorber is suitable for controlling the vibration of the piping system. (author)

  5. IEA-R1 renewed primary coolant piping system stress analysis

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

    2015-01-01

    A partial replacement of the IEA-R1 piping system was conducted in 2014. The aim of this work is to perform the stress analysis of the renewed primary piping system of the IEA-R1, taking into account the as built conditions and the pipe modifications. The nuclear research reactor IEA-R1 is a pool type reactor designed by Babcox-Willcox, which is operated by IPEN since 1957. The primary coolant system is responsible for removing the residual heat of the Reactor core. As a part of the life management, a regular inspection detected some degradation in the primary piping system. In consequence, part of the piping system was replaced. The partial renewing of the primary piping system did not imply in major piping layout modifications. However, the stress condition of the piping systems had to be reanalyzed. The structural stress analysis of the primary piping systems is now presented and the final results are discussed. (author)

  6. Integrated model of port oil piping transportation system safety including operating environment threats

    Directory of Open Access Journals (Sweden)

    Kołowrocki Krzysztof

    2017-06-01

    Full Text Available The paper presents an integrated general model of complex technical system, linking its multistate safety model and the model of its operation process including operating environment threats and considering variable at different operation states its safety structures and its components safety parameters. Under the assumption that the system has exponential safety function, the safety characteristics of the port oil piping transportation system are determined.

  7. Evaluation of temporary non-code repairs in safety class 3 piping systems

    International Nuclear Information System (INIS)

    Godha, P.C.; Kupinski, M.; Azevedo, N.F.

    1996-01-01

    Temporary non-ASME Code repairs in safety class 3 pipe and piping components are permissible during plant operation in accordance with Nuclear Regulatory Commission Generic Letter 90-05. However, regulatory acceptance of such repairs requires the licensee to undertake several timely actions. Consistent with the requirements of GL 90-05, this paper presents an overview of the detailed evaluation and relief request process. The technical criteria encompasses both ductile and brittle piping materials. It also lists appropriate evaluation methods that a utility engineer can select to perform a structural integrity assessment for design basis loading conditions to support the use of temporary non-Code repair for degraded piping components. Most use of temporary non-code repairs at a nuclear generating station is in the service water system which is an essential safety related system providing the ultimate heat sink for various plant systems. Depending on the plant siting, the service water system may use fresh water or salt water as the cooling medium. Various degradation mechanisms including general corrosion, erosion/corrosion, pitting, microbiological corrosion, galvanic corrosion, under-deposit corrosion or a combination thereof continually challenge the pressure boundary structural integrity. A good source for description of corrosion degradation in cooling water systems is provided in a cited reference

  8. Seismic analysis of piping systems subjected to multiple support excitations

    International Nuclear Information System (INIS)

    Sundararajan, C.; Vaish, A.K.; Slagis, G.C.

    1981-01-01

    The paper presents the results of a comparative study between the multiple response spectrum method and the time-history method for the seismic analysis of nuclear piping systems subjected to different excitation at different supports or support groups. First, the necessary equations for the above analysis procedures are derived. Then, three actual nuclear piping systems subjected to single and multiple excitations are analyzed by the different methods, and extensive comparisons of the results (stresses) are made. Based on the results, it is concluded that the multiple response spectrum analysis gives acceptable results as compared to the ''exact'', but much more costly, time-history analysis. 6 refs

  9. Flexible mobile robot system for smart optical pipe inspection

    Science.gov (United States)

    Kampfer, Wolfram; Bartzke, Ralf; Ziehl, Wolfgang

    1998-03-01

    Damages of pipes can be inspected and graded by TV technology available on the market. Remotely controlled vehicles carry a TV-camera through pipes. Thus, depending on the experience and the capability of the operator, diagnosis failures can not be avoided. The classification of damages requires the knowledge of the exact geometrical dimensions of the damages such as width and depth of cracks, fractures and defect connections. Within the framework of a joint R&D project a sensor based pipe inspection system named RODIAS has been developed with two partners from industry and research institute. It consists of a remotely controlled mobile robot which carries intelligent sensors for on-line sewerage inspection purpose. The sensor is based on a 3D-optical sensor and a laser distance sensor. The laser distance sensor is integrated in the optical system of the camera and can measure the distance between camera and object. The angle of view can be determined from the position of the pan and tilt unit. With coordinate transformations it is possible to calculate the spatial coordinates for every point of the video image. So the geometry of an object can be described exactly. The company Optimess has developed TriScan32, a special software for pipe condition classification. The user can start complex measurements of profiles, pipe displacements or crack widths simply by pressing a push-button. The measuring results are stored together with other data like verbal damage descriptions and digitized images in a data base.

  10. Frequency domain analysis of piping systems under short duration loading

    International Nuclear Information System (INIS)

    Sachs, K.; Sand, H.; Lockau, J.

    1981-01-01

    In piping analysis two procedures are used almost exclusively: the modal superposition method for relatively long input time histories (e.g., earthquake) and direct integration of the equations of motion for short input time histories. A third possibility, frequency domain analysis, has only rarely been applied to piping systems to date. This paper suggests the use of frequency domain analysis for specific piping problems for which only direct integration could be used in the past. Direct integration and frequency domain analysis are compared, and it is shown that the frequency domain method is less costly if more than four or five load cases are considered. In addition, this method offers technical advantages, such as more accurate representation of modal damping and greater insight into the structural behavior of the system. (orig.)

  11. Failure and factors of safety in piping system design

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1993-01-01

    An important body of test and performance data on the behavior of piping systems has led to an ongoing reassessment of the code stress allowables and their safety margin. The codes stress allowables, and their factors of safety, are developed from limits on the incipient yield (for ductile materials), or incipient rupture (for brittle materials), of a test specimen loaded in simple tension. In this paper, we examine the failure theories introduced in the B31 and ASME III codes for piping and their inherent approximations compared to textbook failure theories. We summarize the evolution of factors of safety in ASME and B31 and point out that, for piping systems, it is appropriate to reconsider the concept and definition of factors of safety

  12. Periodic inspection for safety of CANDU heat transport piping systems

    International Nuclear Information System (INIS)

    Ellyin, F.

    1979-10-01

    Periodic inspection of heat transport and emergency core cooling piping systems is intended to maintain an adequate level of safety throughout the life of the plant, and to protect plant personnel and the public from the consequences of a failure and release of fission products. This report outlines a rational approach to the periodic inspection based on a fully probabilistic model. It demonstrates the methodology based on theoretical treatment and experimental data whereby the strength of a pressurized pipe or vessel containing a defect could be evaluated. It also shows how the extension of the defect at various lifetimes could be predicted. These relationships are prerequisite for the probabilistic formulation and analysis for the periodic inspection of piping systems

  13. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Anglaret, G.; Lasne, M.

    1983-08-01

    Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

  14. BOA: Pipe-asbestos insulation removal robot system

    International Nuclear Information System (INIS)

    Schempf, H.; Bares, J.; Mutschler, E.

    1995-01-01

    This paper describes the BOA system, a mobile pipe-external crawler used to remotely strip and bag (possibly contaminated) asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations across the DOE weapons complex. The mechanical removal of ACLIM is very cost-effective due to the relatively low productivity and high cost involved in human removal scenarios. BOA, a mechanical system capable of removing most forms of lagging (paper, plaster, aluminum sheet, clamps, screws and chicken-wire), and insulation (paper, tar, asbestos fiber, mag-block) uses a circular cutter and compression paddles to cut and strip the insulation off the pipe through compression, while a HEPA-filter and encapsulant system maintain a certifiable vacuum and moisture content inside the system and on the pipe, respectively. The crawler system has been built and is currently undergoing testing. Key design parameters and performance parameters are developed and used in performance testing. Since the current system is a testbed, we also discuss future enhancements and outline two deployment scenarios (robotic and manual) for the final system to be designed and completed by the end of FY '95. An on-site demonstration is currently planned for Fernald in Ohio and Oak Ridge in Tennessee

  15. An experimental study on damping characteristics of mechanical snubber for nuclear power plant piping systems

    International Nuclear Information System (INIS)

    Chiba, T.; Kobayashi, H.; Kitamura, K.; Ando, K.; Koyanagi, R.

    1983-01-01

    The objectives of this study are 1) to clarify the damping characteristics and the dynamic stiffness of mechanical snubber, 2) to take the damping characteristics of mechanical snubber into the damping evaluation method obtained in SDREP. Therefore, following vibration tests were conducted. 1) Component test: As a first step, mechanical snubbers were excited with sinusoidal wave, and damping ratio and dynamic stiffness were measured at several loading levels. 2) Piping model test: Second, a 8'' diameter x 16 m length 3-dimensional piping model simulating the supporting conditions of actual piping systems was tested. Damping ratio and made shapes of piping model with mechanical snubbers were measured at several supporting conditions and response levels. From the results of these tests, the damping characteristics and the dynamic stiffness of mechanical snubber can be summarized as follows: 1) The damping effect of mechanical snubber is as strong as that of oil snubber. 2) Mechanical snubber contributes effectively to the damping of piping system, and it is indicated that the damping characteristics of mechanical snubber is applicable to the damping evaluation method obtained in SDREP. (orig./HP)

  16. Test of Seal System for Flexible Pipe End Fitting

    DEFF Research Database (Denmark)

    Banke, Lars; Jensen, Thomas Gregers

    1999-01-01

    The purpose of the end fitting seal system is to ensure leak proof termination of flexible pipes. The seal system of an NKT end fitting normally consists of a number of ring joint gaskets mounted in a steel sleeve on the outside of the polymeric inner liner of the pipe. The seal system is activated...... by compression of the gaskets, thus using the geometry to establish a seal towards the inner liner of the pipe and the steel sleeve of the end fitting. This paper describes how the seal system of an end fitting can be tested using an autoclave. By regulating temperature and pressure, the seal system can...... be tested up to 130oC and 51.7 MPa. Pressure, temperature and the mechanical behaviours of the pipe are measured for use in further research. The set-up is used to test the efficiency of the seal system as function of parameters such as cross sectional shapes of the gaskets, tolerances between gaskets...

  17. Piping information centralized management system for nuclear plant, PIMAS

    International Nuclear Information System (INIS)

    Matsumoto, Masaru

    1977-01-01

    Piping works frequently cause many troubles in the progress of construction works, because piping is the final procedure in design and construction and is forced to suffer the problems in earlier stages. The enormous amount of data on quality control and management leads to the employment of many unskilled designers of low technical ability, and it causes confusion in installation and inspection works. In order to improve the situation, the ''piping information management system for nuclear plants (PIMAS)'' has been introduced attempting labor-saving and speed-up. Its main purposes are the mechanization of drafting works, the centralization of piping informations, labor-saving and speed-up in preparing production control data and material management. The features of the system are as follows: anyone can use the same informations whenever he requires them because the informations handled in design works are contained in a large computer; the system can be operated on-line, and the terminals are provided in the sections which require informations; and the sub-systems are completed for preparing a variety of drawings and data. Through the system, material control has become possible by using the material data in each plant, stock material data and the information on the revision of drawings in the design department. Efficiency improvement and information centralization in the manufacturing department have also been achieved because the computer has prepared many kinds of slips based on unified drawings and accurate informations. (Wakatsuki, Y.)

  18. Failure rate of piping in hydrogen sulphide systems

    International Nuclear Information System (INIS)

    Hare, M.G.

    1993-08-01

    The objective of this study is to provide information about piping failures in hydrogen sulphide service that could be used to establish failures rates for piping in 'sour service'. Information obtained from the open literature, various petrochemical industries and the Bruce Heavy Water Plant (BHWP) was used to quantify the failure analysis data. On the basis of this background information, conclusions from the study and recommendations for measures that could reduce the frequency of failures for piping systems at heavy water plants are presented. In general, BHWP staff should continue carrying out their present integrity and leak detection programmes. The failure rate used in the safety studies for the BHWP appears to be based on the rupture statistics for pipelines carrying sweet natural gas. The failure rate should be based on the rupture rate for sour gas lines, adjusted for the unique conditions at Bruce

  19. Development and testing of restraints for nuclear piping systems

    International Nuclear Information System (INIS)

    Kelly, J.M.; Skinner, M.S.

    1980-06-01

    As an alternative to current practice of pipe restraint within nuclear power plants it has been proposed to adopt restraints capable of dissipating energy in the piping system. The specific mode of energy dissipation focused upon in these studies is the plastic yielding of steels utilizing relative movement between the pipe and the base of the restraint, a general mechanism which has been proven as reliable in several allied studies. This report discusses the testing of examples of two energy-absorbing devices, the results of this testing and the conclusions drawn. This study concentrated on the specific relevant performance characteristics of hysteretic behavior and degradation with use. The testing consisted of repetitive continuous loadings well into the plastic ranges of the devices in a sinusoidal or random displacement controlled mode

  20. Analysis of piping system response to seismic excitations

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    This paper describes a numerical algorithm for analyzing piping system response to seismic excitations. The numerical model of the piping considers hoop, flexural, axial, and torsional modes of deformation. Hoop modes generated from internal hydrodynamic loading are superimposed on the bending and twisting modes by two extra degrees of freedom. A time-history analysis technique using the implicit temporal integration scheme is addressed. The time integrator uses a predictor-corrector successive iterative scheme which satisfies the equation of motion. Both geometrical and material nonlinearities are considered. Multiple support excitations, fluid effect, piping insulation, and material dampings can be included in the analysis. Two problems are presented to illustrate the method. The results are discussed in detail

  1. BOA: Asbestos Pipe-Insulation Abatement Robot System

    International Nuclear Information System (INIS)

    Schempf, H.

    1996-01-01

    The BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to the high labor costs and high level of radioactive contamination, making manual removal extremely costly and highly inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee

  2. BOA: Pipe-asbestos insulation removal robot system

    Energy Technology Data Exchange (ETDEWEB)

    Schempf, H.; Bares, J.; Schnorr, W. [Carnegie Mellon Univ., Pittsburgh, PA (United States)

    1995-10-01

    The BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to the high labor costs and high level of radioactive contamination, making manual removal extremely costly and highly inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee.

  3. BOA: Pipe-asbestos insulation removal robot system

    International Nuclear Information System (INIS)

    Schempf, H.; Bares, J.; Schnorr, W.

    1995-01-01

    The BOA system is a mobile pipe-external robotic crawler used to remotely strip and bag asbestos-containing lagging and insulation materials (ACLIM) from various diameter pipes in (primarily) industrial installations. Steam and process lines within the DOE weapons complex warrant the use of a remote device due to the high labor costs and high level of radioactive contamination, making manual removal extremely costly and highly inefficient. Currently targeted facilities for demonstration and remediation are Fernald in Ohio and Oak Ridge in Tennessee

  4. Comparison of ICEPEL predictions with single elbow flexible piping system experiment

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.W.

    1978-01-01

    The ICEPEL Code for coupled hydrodynamic-structural response analysis of piping systems is used to analyze an experiment on the response of flexible piping systems to internal pressure pulses. The piping system consisted of two flexible Nickel-200 pipes connected in series through a 90 0 thick-walled stainless steel elbow. A tailored pressure pulse generated by a calibrated pulse gun is stabilized in a long thick-walled stainless steel pipe leading to the flexible piping system which ended with a heavy blind flange. The analytical results of pressure and circumferential strain histories are discussed and compared against the experimental data obtained by Stanford Research Institute

  5. Leak detection system for a high temperature fluid pipe

    International Nuclear Information System (INIS)

    Puyal, C.; Meuwisse, C.

    1989-01-01

    The leak detection system is made by a cable with at least two isolated electrical conductors, close to the wall of the pipe. The material of the cable is chosen so as to change its electrical characteristics if a leak causes heating of the cable. A detector at one end of the cable can measure the modifications of the electrical characteristics [fr

  6. ASBESTOS PIPE-INSULATION REMOVAL ROBOT SYSTEM; FINAL

    International Nuclear Information System (INIS)

    Unknown

    2000-01-01

    This final topical report details the development, experimentation and field-testing activities for a robotic asbestos pipe-insulation removal robot system developed for use within the DOE's weapon complex as part of their ER and WM program, as well as in industrial abatement. The engineering development, regulatory compliance, cost-benefit and field-trial experiences gathered through this program are summarized

  7. Role of theoretical dynamics in vibration diagnostics of pipe systems

    International Nuclear Information System (INIS)

    Rejent, B.

    1992-01-01

    The importance of vibration diagnostics of pipe systems and the relevance of theoretical dynamics are shown using examples. The problems are discussed of vibration diagnostics of the primary circuit of a nuclear power plant with viscous seismic dampers installed. (M.D.) 7 figs., 5 refs

  8. Development of new damping devices for piping

    International Nuclear Information System (INIS)

    Kobayashi, Hiroe

    1991-01-01

    An increase of the damping ratio is known to be very effective for the seismic design of a piping system. Increasing the damping ratio and reducing the seismic response of the piping system, the following three types of damping devices for piping systems are introduced: (1) visco-elastic damper, (2) elasto-plastic damper and (3) compact dynamic damper. The dynamic characteristics of these damping devices were investigated by the component test and the applicability of them to the piping system was confirmed by the vibration test using a three dimensional piping model. These damping devices are more effective than mechanical snubbers to reduce the vibration of the piping system. (author)

  9. Flow Tones in a Pipeline-Cavity System: Effect of Pipe Asymmetry

    Energy Technology Data Exchange (ETDEWEB)

    D. Erdem; D. rockwell; P. Oshkai; M. Pollack

    2002-05-29

    Flow tones in a pipeline-cavity system are characterized in terms of unsteady pressure within the cavity and along the pipe. The reference case corresponds to equal lengths of pipe connected to the inlet and outlet ends of the cavity. Varying degrees of asymmetry of this pipe arrangement are investigated. The asymmetry is achieved by an extension of variable length, which is added to the pipe at the cavity outlet. An extension length as small as a few percent of the acoustic wavelength of the resonant mode can yield a substantial reduction in the pressure amplitude of the flow tone. This amplitude decrease occurs in a similar fashion within both the cavity and the pipe resonator, which indicates that it is a global phenomenon. Furthermore, the decrease of pressure amplitude is closely correlated with a decrease of the Q (quality)-factor of the predominant spectral component of pressure. At a sufficiently large value of extension length, however, the overall form of the pressure spectrum recovers to the form that exists at zero length of the extension. Further insight is provided by variation of the inflow velocity at selected values of extension length. Irrespective of its value, both the magnitude and frequency of the peak pressure exhibit a sequence of resonant-like states. Moreover, the maximum attainable magnitude of the peak pressure decreases with increasing extension length.

  10. Flow Tones in a Pipeline-Cavity System: Effect of Pipe Asymmetry

    Energy Technology Data Exchange (ETDEWEB)

    D. Erdem; D. Rockwell; P.L. Oshkai; M. Pollack

    2001-02-28

    Flow tones in a pipeline-cavity system are characterized in terms of unsteady pressure within the cavity and along the pipe. The reference case corresponds to equal lengths of pipe connected to the inlet and outlet ends of the cavity. Varying degrees of asymmetry of this pipe arrangement are investigated. The asymmetry is achieved by an extension of variable length, which is added to the pipe at the cavity outlet. An extension length as small as a few percent of the acoustic wavelength of the resonant mode can yield a substantial reduction in the pressure amplitude of the flow tone. This amplitude decrease occurs in a similar fashion within both the cavity and the pipe resonator, which indicates that it is a global phenomenon. Furthermore, the decrease of pressure amplitude is closely correlated with a decrease of the Q (quality)-factor of the predominant spectral component of pressure. At a sufficiently large value of extension length, however, the overall form of the pressure spectrum recovers to the form that exists at zero length of the extension. Further insight is provided by variation of the inflow velocity at selected values of extension length. Irrespective of its value, both the magnitude and frequency of the peak pressure exhibit a sequence of resonant-like states. moreover, the maximum attainable magnitude of the peak pressure decreases with increasing extension length.

  11. Test and evaluation about damping characteristics of hanger supports for nuclear power plant piping systems (Seismic Damping Ratio Evaluation Program)

    International Nuclear Information System (INIS)

    Shibata, H.; Ito, A.; Tanaka, K.; Niino, T.; Gotoh, N.

    1981-01-01

    Generally, damping phenomena of structures and equipments is caused by very complex energy dissipation. Especially, as piping systems are composed of many components, it is very difficult to evaluate damping characteristics of its system theoretically. On the other hand, the damping value for aseismic design of nuclear power plants is very important design factor to decide seismic response loads of structures, equipments and piping systems. The very extensive studies titled SDREP (Seismic Damping Ratio Evaluation Program) were performed to establish proper damping values for seismic design of piping as a joint work among a university, electric companies and plant makers. In SDREP, various systematic vibration tests were conducted to investigate factors which may contribute to damping characteristics of piping systems and to supplement the data of the pre-operating tests. This study is related to the component damping characteristics tests of that program. The object of this study is to clarify damping characteristics and mechanism of hanger supports used in piping systems, and to establish the evaluation technique of dispersing energy at hanger support points and its effect to the total damping ability of piping system. (orig./WL)

  12. Automatic seismic support design of piping system by an object oriented expert system

    International Nuclear Information System (INIS)

    Nakatogawa, T.; Takayama, Y.; Hayashi, Y.; Fukuda, T.; Yamamoto, Y.; Haruna, T.

    1990-01-01

    The seismic support design of piping systems of nuclear power plants requires many experienced engineers and plenty of man-hours, because the seismic design conditions are very severe, the bulk volume of the piping systems is hyge and the design procedures are very complicated. Therefore we have developed a piping seismic design expert system, which utilizes the piping design data base of a 3 dimensional CAD system and automatically determines the piping support locations and support styles. The data base of this system contains the maximum allowable seismic support span lengths for straight piping and the span length reduction factors for bends, branches, concentrated masses in the piping, and so forth. The system automatically produces the support design according to the design knowledge extracted and collected from expert design engineers, and using design information such as piping specifications which give diameters and thickness and piping geometric configurations. The automatic seismic support design provided by this expert system achieves in the reduction of design man-hours, improvement of design quality, verification of design result, optimization of support locations and prevention of input duplication. In the development of this system, we had to derive the design logic from expert design engineers and this could not be simply expressed descriptively. Also we had to make programs for different kinds of design knowledge. For these reasons we adopted the object oriented programming paradigm (Smalltalk-80) which is suitable for combining programs and carrying out the design work

  13. Dimensional control of buttwelding pipe fitting for nuclear power plant Class 1 piping systems

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Moore, S.E.; Robinson, J.N.

    1976-11-01

    Dimensional controls of wrought steel buttwelding fittings are examined from the standpoint of design adequacy. A fairly large number of fittings were purchased from different manufacturers. The dimensions of each fitting were measured and correlated along with additional information obtained from the manufacturers in an effort to establish ''standard'' shapes. This information and a critical examination of the present ANSI standards is used to develop a ''Supplementary Standard.'' The Supplementary Standard is intended to provide improved dimensional control and more complete design information for fittings used in Class 1 nuclear power plant piping systems

  14. Practical application of equivalent linearization approaches to nonlinear piping systems

    International Nuclear Information System (INIS)

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

    1995-01-01

    The use of mechanical energy absorbers as an alternative to conventional hydraulic and mechanical snubbers for piping supports has attracted a wide interest among researchers and practitioners in the nuclear industry. The basic design concept of energy absorbers (EA) is to dissipate the vibration energy of piping systems through nonlinear hysteretic actions of EA exclamation point s under design seismic loads. Therefore, some type of nonlinear analysis needs to be performed in the seismic design of piping systems with EA supports. The equivalent linearization approach (ELA) can be a practical analysis tool for this purpose, particularly when the response approach (RSA) is also incorporated in the analysis formulations. In this paper, the following ELA/RSA methods are presented and compared to each other regarding their practice and numerical accuracy: Response approach using the square root of sum of squares (SRSS) approximation (denoted RS in this paper). Classical ELA based on modal combinations and linear random vibration theory (denoted CELA in this paper). Stochastic ELA based on direct solution of response covariance matrix (denoted SELA in this paper). New algorithms to convert response spectra to the equivalent power spectral density (PSD) functions are presented for both the above CELA and SELA methods. The numerical accuracy of the three EL are studied through a parametric error analysis. Finally, the practicality of the presented analysis is demonstrated in two application examples for piping systems with EA supports

  15. Terahertz inline wall thickness monitoring system for plastic pipe extrusion

    Energy Technology Data Exchange (ETDEWEB)

    Hauck, J., E-mail: j.hauck@skz.de, E-mail: d.stich@skz.de, E-mail: p.heidemeyer@skz.de, E-mail: m.bastian@skz.de, E-mail: t.hochrein@skz.de; Stich, D., E-mail: j.hauck@skz.de, E-mail: d.stich@skz.de, E-mail: p.heidemeyer@skz.de, E-mail: m.bastian@skz.de, E-mail: t.hochrein@skz.de; Heidemeyer, P., E-mail: j.hauck@skz.de, E-mail: d.stich@skz.de, E-mail: p.heidemeyer@skz.de, E-mail: m.bastian@skz.de, E-mail: t.hochrein@skz.de; Bastian, M., E-mail: j.hauck@skz.de, E-mail: d.stich@skz.de, E-mail: p.heidemeyer@skz.de, E-mail: m.bastian@skz.de, E-mail: t.hochrein@skz.de; Hochrein, T., E-mail: j.hauck@skz.de, E-mail: d.stich@skz.de, E-mail: p.heidemeyer@skz.de, E-mail: m.bastian@skz.de, E-mail: t.hochrein@skz.de [SKZ - German Plastics Center, Wuerzburg (Germany)

    2014-05-15

    Conventional and commercially available inline wall thickness monitoring systems for pipe extrusion are usually based on ultrasonic or x-ray technology. Disadvantages of ultrasonic systems are the usual need of water as a coupling media and the high damping in thick walled or foamed pipes. For x-ray systems special safety requirements have to be taken into account because of the ionizing radiation. The terahertz (THz) technology offers a novel approach to solve these problems. THz waves have many properties which are suitable for the non-destructive testing of plastics. The absorption of electrical isolators is typically very low and the radiation is non-ionizing in comparison to x-rays. Through the electromagnetic origin of the THz waves they can be used for contact free measurements. Foams show a much lower absorption in contrast to acoustic waves. The developed system uses THz pulses which are generated by stimulating photoconductive switches with femtosecond laser pulses. The time of flight of THz pulses can be determined with a resolution in the magnitude of several ten femtoseconds. Hence the thickness of an object like plastic pipes can be determined with a high accuracy by measuring the time delay between two reflections on materials interfaces e.g. at the pipe's inner and outer surface, similar to the ultrasonic technique. Knowing the refractive index of the sample the absolute layer thickness from the transit time difference can be calculated easily. This method in principle also allows the measurement of multilayer systems and the characterization of foamed pipes.

  16. IEA-R1 primary and secondary coolant piping systems coupled stress analysis

    International Nuclear Information System (INIS)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A.; Mattar Neto, Miguel

    2013-01-01

    The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The operation to 5 MW power limit was only possible after the conduction of life management and modernization programs in the last two decades. In these programs the components of the coolant systems, which are responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping systems is presented and the final results are discussed. (author)

  17. Research on pipe welding information management system basedon RFID

    Directory of Open Access Journals (Sweden)

    Liu Xun

    2016-01-01

    Full Text Available This paper introduces the construction background, construction target and construction principle of the pipe welding management system based on RFID. Then, describes the specific requirements of the system. The basic principle and key technology of the system are introduced. The structure of the system (including the system design, the selections of handheld devices and high frequency passive RFID tags is described .Then the system management software designs (including software structure, the main functions of the management center system and the main functions of the handheld detection system are described in detail. Finally, the management system is implemented, and it is deployed to several Gas Co, which has chieved good results.

  18. A countermeasure for external stress corrosion cracking in piping components by means of residual stress improvement on the outer surface

    International Nuclear Information System (INIS)

    Tanaka, Yasuhiro; Umemoto, Tadahiro

    1988-01-01

    Many techniques have been proposed as countermeasures for the External Stress Corrosion Cracking (ESCC) on austenitic stainless steel piping caused by sea salt particles. However, not one seems perfect. The method proposed here is an expansion of IHSI (Induction Heating Stress Improvement) which has been successfully implemented in many nuclear power plants as a remedy for Intergranular Stress Corrossion Cracking. The proposed method named EIHSI (External IHSI) can make the residual stress compressive on the outer surface of the piping components. In order to confirm the effectiveness of EIHSI, one series of tests were conducted on a weld joint between the pipe flange and the straight pipe. The measured residual stresses and also the results of the cracking test revealed that EIHSI is a superior method to suppress the ESCC. The outline of EIHSI and the verification tests are presented in this paper. (author)

  19. Impact of the amount of working fluid in loop heat pipe to remove waste heat from electronic component

    Directory of Open Access Journals (Sweden)

    Smitka Martin

    2014-03-01

    Full Text Available One of the options on how to remove waste heat from electronic components is using loop heat pipe. The loop heat pipe (LHP is a two-phase device with high effective thermal conductivity that utilizes change phase to transport heat. It was invented in Russia in the early 1980’s. The main parts of LHP are an evaporator, a condenser, a compensation chamber and a vapor and liquid lines. Only the evaporator and part of the compensation chamber are equipped with a wick structure. Inside loop heat pipe is working fluid. As a working fluid can be used distilled water, acetone, ammonia, methanol etc. Amount of filling is important for the operation and performance of LHP. This work deals with the design of loop heat pipe and impact of filling ratio of working fluid to remove waste heat from insulated gate bipolar transistor (IGBT.

  20. Two-Pipe Chilled Beam System for Both Cooling and Heating of Office Buildings

    DEFF Research Database (Denmark)

    Afshari, Alireza; Gordnorouzi, Rouzbeh; Hultmark, Göran

    2013-01-01

    Simulations were performed to compare a conventional 4-pipe chilled beam system and a 2-pipe chilled beam system. The objective was to establish requirements, possibilities and limitations for a well-functioning 2-pipe chilled beam system for both cooling and heating of office buildings. The buil...

  1. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Existing boilers, pressure vessels or piping systems. 50... ENGINEERING GENERAL PROVISIONS Application § 50.05-5 Existing boilers, pressure vessels or piping systems. (a) Whenever doubt exists as to the safety of an existing boiler, pressure vessel, or piping system, the marine...

  2. Heat-Pipe-Associated Localized Thermoelectric Power Generation System

    Science.gov (United States)

    Kim, Pan-Jo; Rhi, Seok-Ho; Lee, Kye-Bock; Hwang, Hyun-Chang; Lee, Ji-Su; Jang, Ju-Chan; Lee, Wook-Hyun; Lee, Ki-Woo

    2014-06-01

    The present study focused on how to improve the maximum power output of a thermoelectric generator (TEG) system and move heat to any suitable space using a TEG associated with a loop thermosyphon (loop-type heat pipe). An experimental study was carried out to investigate the power output, the temperature difference of the thermoelectric module (TEM), and the heat transfer performance associated with the characteristic of the researched heat pipe. Currently, internal combustion engines lose more than 35% of their fuel energy as recyclable heat in the exhaust gas, but it is not easy to recycle waste heat using TEGs because of the limited space in vehicles. There are various advantages to use of TEGs over other power sources, such as the absence of moving parts, a long lifetime, and a compact system configuration. The present study presents a novel TEG concept to transfer heat from the heat source to the sink. This technology can transfer waste heat to any location. This simple and novel design for a TEG can be applied to future hybrid cars. The present TEG system with a heat pipe can transfer heat and generate power of around 1.8 V with T TEM = 58°C. The heat transfer performance of a loop-type heat pipe with various working fluids was investigated, with water at high heat flux (90 W) and 0.05% TiO2 nanofluid at low heat flux (30 W to 70 W) showing the best performance in terms of power generation. The heat pipe can transfer the heat to any location where the TEM is installed.

  3. Evaluation of stress corrosion crack growth in BWR piping systems

    International Nuclear Information System (INIS)

    Kassir, M.; Sharma, S.; Reich, M.; Chang, M.T.

    1985-05-01

    This report presents the results of a study conducted to evaluate the effects of stress intensity factor and environment on the growth behavior of intergranular stress corrosion cracks in type 304 stainless steel piping systems. Most of the detected cracks are known to be circumferential in shape, and initially started at the inside surface in the heat affected zone near girth welds. These cracks grow both radially in-depth and circumferentially in length and, in extreme cases, may cause leakage in the installation. The propagation of the crack is essentially due to the influence of the following simultaneous factors: (1) the action of applied and residual stress; (2) sensitization of the base metal in the heat affected zone adjacent to girth weld; and (3) the continuous exposure of the material to an aggressive environment of high temperature water containing dissolved oxygen and some levels of impurities. Each of these factors and their effects on the piping systems is discussed in detail in the report. The report also evaluates the time required for hypothetical cracks in BWR pipes to propagate to their critical size. The pertinent times are computed and displayed graphically. Finally, parametric study is performed in order to assess the relative influence and sensitivity of the various input parameters (residual stress, crack growth law, diameter of pipe, initial size of defect, etc.) which have bearing on the growth behavior of the intergranular stress corrosion cracks in type 304 stainless steel. Cracks in large-diameter as well as in small-diameter pipes are considered and analyzed. 27 refs., 25 figs., 10 tabs

  4. Engineering design aspects of the heat-pipe power system

    International Nuclear Information System (INIS)

    Capell, B.M.; Houts, M.G.; Poston, D.I.; Berte, M.

    1997-10-01

    The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations

  5. Engineering design aspects of the heat-pipe power system

    Science.gov (United States)

    Capell, B. M.; Houts, M. G.; Poston, D. I.; Berte, M.

    1997-01-01

    The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.

  6. Alternative methods for the seismic analysis of piping systems

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This document is a review of 12 methods and criteria for the seismic analysis of piping systems. Each of the twelve chapters in this document cover the important technical aspects of a given method. The technical aspects presented are those the Subcommittee on Dynamic Stress Criteria believe important to the application of the method, and should not be considered as a positive or negative endorsement for any of the methods. There are many variables in an analysis of a piping system that can influence the selection of the analysis method and criteria to be applied. These variable include system configuration, technical issues, precedent, licensing considerations, and regulatory acceptance. They must all be considered in selecting the appropriate seismic analysis method and criteria. This is relevant for nuclear power plants

  7. Identification of significant problems related to light water reactor piping systems

    International Nuclear Information System (INIS)

    1980-07-01

    Work on the project was divided into three tasks. In Task 1, past surveys of LWR piping system problems and recent Licensee Event Report summaries are studied to identify the significant problems of LWR piping systems and the primary causes of these problems. Pipe cracking is identified as the most recurring problem and is mainly due to the vibration of pipes due to operating pump-pipe resonance, fluid-flow fluctuations, and vibration of pipe supports. Research relevant to the identified piping system problems is evaluated. Task 2 studies identify typical LWR piping systems and the current loads and load combinations used in the design of these systems. Definitions of loads are reviewed. In Task 3, a comparative study is carried out on the use of nonlinear analysis methods in the design of LWR piping systems. The study concludes that the current linear-elastic methods of analysis may not predict accurately the behavior of piping systems under seismic loads and may, under certain circumstances, result in nonconservative designs. Gaps at piping supports are found to have a significant effect on the response of the piping systems

  8. 46 CFR 28.255 - Bilge pumps, bilge piping, and dewatering systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Bilge pumps, bilge piping, and dewatering systems. 28... the Aleutian Trade § 28.255 Bilge pumps, bilge piping, and dewatering systems. (a) Each vessel must be equipped with a bilge pump and bilge piping capable of draining any watertight compartment, other than...

  9. Critical element development of standard components for pipe welding/cutting by CO{sub 2} laser

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1994-11-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor(ITER), an internal access is inevitable for welding/cutting of cooling pipes of in-vessel components, because of spatial constraint due to a narrow port opening space. An internal-access pipe welding/cutting equipment is being developed in JAERI. Internal access is to approach through inside a pipe to a welding/cutting position, to use 10kW CO{sub 2} laser beam, and to be applicable to both welding and cutting with using a same processing head. A welding/cutting processing head with 10kW CO{sub 2} laser beam has been fabricated and the basic feasibility has been successfully demonstrated for studies of the internal-access pipe welding/cutting concept using 100-A stainless steel pipe with a thickness of 6.3mm. In this study, the optimum focal point of laser beam, laser power and traveling speed of the head have been investigated together with an adjusting mechanism of a relative distance between the head and the pipe wall. In addition, the radiation resistance of critical elements such as optical lens has been investigated. (author).

  10. Situation of secondary system piping wearing in overseas nuclear power plants

    International Nuclear Information System (INIS)

    Chiba, Goro

    2005-01-01

    In consideration of secondary system piping rupture accident at Mihama Nuclear Power Station Unit 3 of Kansai Electric Power Company in August 2004, the management system of secondary pipe wall thickness of Japan and foreign countries were investigated. Moreover, the tendency of the secondary piping thinning events on overseas which the Institute of Nuclear Safety System, Inc. (INSS) obtained was analyzed in order to verify the validity of the Japanese management system. Consequently, it was shown that in the U.S., the fault phenomenon of secondary system piping was reported continuously, and there were also many cases of both degradation and penetration of pipe wall. (author)

  11. Development of thermal fatigue evaluation methods of piping systems

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Itoh, Takamoto; Okazaki, Masakazu; Okuda, Yukihiko; Kamaya, Masayuki; Nakamura, Akira; Nakamura, Hitoshi; Machida, Hideo; Matsumoto, Masaaki

    2014-01-01

    Nuclear piping has various kinds of thermal fatigue failure modes. Main causes of thermal loads are structural responses to fluid temperature changes during plant operation. These phenomena have complex mechanisms and many patterns, so that their problems still occur in spite of well-known issues. The guideline of the JSME (Japan Society of Mechanical Engineering) for estimation of thermal fatigue failures in piping system is employed as Japanese regulation. To improve this guideline, generation mechanisms of thermal load and fatigue failure have been investigated and summarized into the knowledgebase. And numerical simulation methods to replace experimental based methods were studied. Furthermore, probabilistic failure analysis approach with main influence parameters was investigated to be applied for the plant system safety. Thus, based on the knowledge, estimation methods revised from the JSME guideline were proposed. (author)

  12. Smart Pipe System for a Shipyard 4.0.

    Science.gov (United States)

    Fraga-Lamas, Paula; Noceda-Davila, Diego; Fernández-Caramés, Tiago M; Díaz-Bouza, Manuel A; Vilar-Montesinos, Miguel

    2016-12-20

    As a result of the progressive implantation of the Industry 4.0 paradigm, many industries are experimenting a revolution that shipyards cannot ignore. Therefore, the application of the principles of Industry 4.0 to shipyards are leading to the creation of Shipyards 4.0. Due to this, Navantia, one of the 10 largest shipbuilders in the world, is updating its whole inner workings to keep up with the near-future challenges that a Shipyard 4.0 will have to face. Such challenges can be divided into three groups: the vertical integration of production systems, the horizontal integration of a new generation of value creation networks, and the re-engineering of the entire production chain, making changes that affect the entire life cycle of each piece of a ship. Pipes, which exist in a huge number and varied typology on a ship, are one of the key pieces, and its monitoring constitutes a prospective cyber-physical system. Their improved identification, traceability, and indoor location, from production and through their life, can enhance shipyard productivity and safety. In order to perform such tasks, this article first conducts a thorough analysis of the shipyard environment. From this analysis, the essential hardware and software technical requirements are determined. Next, the concept of smart pipe is presented and defined as an object able to transmit signals periodically that allows for providing enhanced services in a shipyard. In order to build a smart pipe system, different technologies are selected and evaluated, concluding that passive and active RFID (Radio Frequency Identification) are currently the most appropriate technologies to create it. Furthermore, some promising indoor positioning results obtained in a pipe workshop are presented, showing that multi-antenna algorithms and Kalman filtering can help to stabilize Received Signal Strength (RSS) and improve the overall accuracy of the system.

  13. Smart Pipe System for a Shipyard 4.0

    Directory of Open Access Journals (Sweden)

    Paula Fraga-Lamas

    2016-12-01

    Full Text Available As a result of the progressive implantation of the Industry 4.0 paradigm, many industries are experimenting a revolution that shipyards cannot ignore. Therefore, the application of the principles of Industry 4.0 to shipyards are leading to the creation of Shipyards 4.0. Due to this, Navantia, one of the 10 largest shipbuilders in the world, is updating its whole inner workings to keep up with the near-future challenges that a Shipyard 4.0 will have to face. Such challenges can be divided into three groups: the vertical integration of production systems, the horizontal integration of a new generation of value creation networks, and the re-engineering of the entire production chain, making changes that affect the entire life cycle of each piece of a ship. Pipes, which exist in a huge number and varied typology on a ship, are one of the key pieces, and its monitoring constitutes a prospective cyber-physical system. Their improved identification, traceability, and indoor location, from production and through their life, can enhance shipyard productivity and safety. In order to perform such tasks, this article first conducts a thorough analysis of the shipyard environment. From this analysis, the essential hardware and software technical requirements are determined. Next, the concept of smart pipe is presented and defined as an object able to transmit signals periodically that allows for providing enhanced services in a shipyard. In order to build a smart pipe system, different technologies are selected and evaluated, concluding that passive and active RFID (Radio Frequency Identification are currently the most appropriate technologies to create it. Furthermore, some promising indoor positioning results obtained in a pipe workshop are presented, showing that multi-antenna algorithms and Kalman filtering can help to stabilize Received Signal Strength (RSS and improve the overall accuracy of the system.

  14. BOA: Asbestos pipe insulation removal robot system. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Schempf, H.; Bares, J.E.

    1995-02-01

    The project described in this report targets the development of a mechanized system for safe, cost-efficient and automated abatement of asbestos containing materials used as pipe insulation. Based on several key design criteria and site visits, a proof-of-concept prototype robot system, dubbed BOA, was designed and built, which automatically strips the lagging and insulation from the pipes, and encapsulates them under complete vacuum operation. The system can operate on straight runs of piping in horizontal or vertical orientations. Currently we are limited to four-inch diameter piping without obstacles as well as a somewhat laborious emplacement and removal procedure -- restrictions to be alleviated through continued development. BOA removed asbestos at a rate of 4-5 ft./h compared to 3 ft./h for manual removal of asbestos with a 3-person crew. The containment and vacuum system on BOA was able to achieve the regulatory requirement for airborne fiber emissions of 0.01 fibers/ccm/ 8-hr. shift. This program consists of two phases. The first phase was completed and a demonstration was given to a review panel, consisting of DOE headquarters and site representatives as well as commercial abatement industry representatives. Based on the technical and programmatic recommendations drafted, presented and discussed during the review meeting, a new plan for the Phase II effort of this project was developed. Phase 11 will consist of a 26-month effort, with an up-front 4-month site-, market-, cost/benefit and regulatory study before the next BOA robot (14 months) is built, and then deployed and demonstrated (3 months) at a DOE site (such as Fernald or Oak Ridge) by the beginning of FY`97.

  15. Leak-before-break behaviour of nuclear piping systems

    International Nuclear Information System (INIS)

    Bartholome, G.; Wellein, R.

    1992-01-01

    The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs

  16. Inelastic analysis of piping systems. A beam-type method for creep and plasticity

    International Nuclear Information System (INIS)

    Roche, R.L.; Hoffmann, A.; Millard, A.

    1979-01-01

    Since many years, piping systems are designed and calculated under elasticity assumptions, using a beam-type method. Thus, the analysis of large systems may be performed at a relatively low cost, using a finite element program. However such a method can not account for inelastic phenomena like plastic deformations or creep. The application of refined three-dimensional shell type method is possible for local components such as curved sections but leads to prohibitive costs for complete piping systems. Therefore simplified methods have been developed, based on a 'global plasticity or creep model'. Following the conventional elastic approach, the pipe element is characterized by variables associated with the center line in the following way: generalized stresses are obtained by integration of local stresses giving way to hoop and tension stresses and to bending and torsional moments; the conjugated strains are identified with uniform hoop and longitudinal strains and variations in neutral axis curvatuves. For plasticity problems, the yield surface is defined by a diagonal quadratic function in terms of the generalized stresses and work hardening parameters. By addition of the Hill's principle and a hardening rule, the formulation is similar to the one commonly used in finite element method. Geometric non linearity due to important deformations of the cross section (often termed 'ovalization') may be treated simultaneously with material non linearity. For this purpose the displacement normal to the pipe surface is represented by trigonometric series expansion, the coefficients of which are determined by minimizing the strain energy over the cross section. The method presented is believed to be a simple economical and accurate tool, for dimensioning computations of large piping systems

  17. PERANCANGAN SISTEM PERPIPAAN KM. NUSANTARA (PIPING SYSTEM

    Directory of Open Access Journals (Sweden)

    Aulia Windyandari

    2013-10-01

    Full Text Available Sistem perpipaan merupakan sistem komplek yang didesain seefektif dan  seefisien mungkin untuk memenuhi kebutuhan dalam kapal ,crew ,muatan dan menjaga keamanan kapal baik saat berlayar ataupun berlabuh. Secara umum sistem pipa dapat diartikan sebagai  bagian utama suatu sistem yang menghubungkan titik dimana fluida di simpan ke titik pengeluaran semua pipa baik untuk memindahkan tenaga atau pemompaan harus dipertimbangkan secara teliti karena keamanan dari sebuah kapal akan tergantung pada susunan perpipaaan seperti halnya pada perlengkapan kapal lainnya Paper ini akan menguraikan tahap-tahap yang harus dilakukan serta pertimbangan-pertimbangan matematis yang diambil  oleh seorang ship engineer  dalam merancang suatu system perpipaan pada kapal KM. Nusantara. Hasil akhir dari paper ini adalah sebuah desain system perpipaan pada pada sebuah kapal,yaitu KM Nusantara, dengan mempertimbangkan system perpipaan yang paling efektif dalam pengoperasiannya.

  18. Inelastic analysis methods for piping systems

    International Nuclear Information System (INIS)

    Boyle, J.T.; Spence, J.

    1980-01-01

    The analysis of pipework systems which operate in an environment where local inelastic strains are evident is one of the most demanding problems facing the stress analyst in the nuclear field. The spatial complexity of even the most modest system makes a detailed analysis using finite element techniques beyond the scope of current computer technology. For this reason the emphasis has been on simplified methods. It is the aim of this paper to provide a reasonably complete, state-of-the-art review of inelastic pipework analysis methods and to attempt to highlight areas where reliable information is lacking and further work is needed. (orig.)

  19. Reconciliation of equipment flexibility effects on piping system dynamic response

    International Nuclear Information System (INIS)

    Geraets, L.H.

    1987-01-01

    Piping systems are connected to equipment; if the equipment cannot be considered as ''rigid'' relative to excitation frequencies, nozzle response spectra techniques, or equipment modeling techniques are used. If the equipment is considered rigid, a fixed anchor is assumed. However, occasionally after (seismic) dynamic analysis has been completed, tests or detailed equipment dynamic analyses demonstrate that the assumption of ''infinite stiff'' is questionable. This paper reviews several classes of equipment (pumps, vessels, reservoirs, heat exchangers), and the associated (piping stresses, support loads, equipment nozzle allowables). Significant divergences between design and ''as built'' results are shown (for heat exchangers in particular). The paper discusses the reconciliation process performed for a belgian PWR plant through the use of less conservative seismic damping data (Code Case N-411)

  20. Detection of underground water distribution piping system and leakages using ground penetrating radar (GPR)

    Science.gov (United States)

    Amran, Tengku Sarah Tengku; Ismail, Mohamad Pauzi; Ahmad, Mohamad Ridzuan; Amin, Mohamad Syafiq Mohd; Sani, Suhairy; Masenwat, Noor Azreen; Ismail, Mohd Azmi; Hamid, Shu-Hazri Abdul

    2017-01-01

    A water pipe is any pipe or tubes designed to transport and deliver water or treated drinking with appropriate quality, quantity and pressure to consumers. The varieties include large diameter main pipes, which supply entire towns, smaller branch lines that supply a street or group of buildings or small diameter pipes located within individual buildings. This distribution system (underground) is used to describe collectively the facilities used to supply water from its source to the point of usage. Therefore, a leaking in the underground water distribution piping system increases the likelihood of safe water leaving the source or treatment facility becoming contaminated before reaching the consumer. Most importantly, leaking can result in wastage of water which is precious natural resources. Furthermore, they create substantial damage to the transportation system and structure within urban and suburban environments. This paper presents a study on the possibility of using ground penetrating radar (GPR) with frequency of 1GHz to detect pipes and leakages in underground water distribution piping system. Series of laboratory experiment was designed to investigate the capability and efficiency of GPR in detecting underground pipes (metal and PVC) and water leakages. The data was divided into two parts: 1. detecting/locating underground water pipe, 2. detecting leakage of underground water pipe. Despite its simplicity, the attained data is proved to generate a satisfactory result indicating GPR is capable and efficient, in which it is able to detect the underground pipe and presence of leak of the underground pipe.

  1. Pipe-CUI-profiler: a portable nucleonic system for detecting corrosion under insulation (CUI) of steel pipes

    International Nuclear Information System (INIS)

    Jaafar Abdullah; Rasif Mohd Zain; Roslan Yahya

    2003-01-01

    Corrosion under insulation (CUI) on the external wall of steel pipes is a common problem in many types of industrial plants. This is mainly due to the presence of moisture or water in the insulation materials. A portable nucleonic system that can be used to detect CUI without the need to remove the insulation materials, has been developed. The system is based on dual-beam gamma-ray absorption technique. It is designed to inspect pipes of internal diameter 50, 65, 80, 90, 100 and 150 mm. Pipeline of these sizes with aluminium or thin steel sheathing, containing fibre-glass or calcium silicate insulation to thicknesses of 25, 40 and 50 mm can be inspected. The system has proven to be a safe, fast and effective method of inspecting insulated pipes. This paper describes the new nucleonic system that has been developed. This paper describes the basic principle of the system and outlines its performance. (Author)

  2. Analytical study for frequency effects on the EPRI/USNRC piping component tests. Part 1: Theoretical basis and model development

    International Nuclear Information System (INIS)

    Adams, T.M.; Branch, E.B.; Tagart, S.W. Jr.

    1994-01-01

    As part of the engineering effort for the Advanced Light Water Reactor the Advanced Reactor Corporation formed a Piping Technical Core Group to develop a set of improved ASME Boiler and Pressure Vessel Code, Section III design rules and approaches for ALWR plant piping and support design. The technical basis for the proposed changes to the ASME Boiler and Pressure Vessel Code developed by Technical Core Group for the design of piping relies heavily on the failure margins determined from the EPRI/USNRC piping component test program. The majority of the component tests forming the basis for the reported margins against failure were run with input frequency to natural frequency ratios (Ω/ω) in the range of 0.74 to 0.87. One concern investigated by the Technical Core Group was the effect which could exist on measured margins if the tests had been run at higher or lower frequency ratios than those in the limited frequency ratio range tested. Specifically, the concern investigated was that the proposed Technical Core Group Piping Stress Criteria will allow piping to be designed in the low frequency range (Ω/ω ≥ 2.0) for which there is little test data from the EPRI/USNRC test program. The purpose of this analytical study was to: (1) evaluate the potential for margin variation as a function of the frequency ratio (R ω = Ω/ω, where Ω is the forcing frequency and ω is the natural component frequency), (2) recommend a margin reduction factor (MRF) that could be applied to margins determined from the EPRI/USNRC test program to adjust those margins for potential margin variation with frequency ratio. Presented in this paper is the analytical approach and methodology, which are inelastic analysis, which was the basis of the study. Also, discussed is the development of the analytical model, the procedure used to benchmark the model to actual test results, and the various parameter studies conducted

  3. Margins for an in-plant piping system under dynamic loading

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. 4 refs., 6 tabs

  4. Review: heat pipe heat exchangers at IROST

    OpenAIRE

    E. Azad

    2012-01-01

    The use of the heat pipe as a component in a heat recovery device has gained worldwide acceptance. Heat pipes are passive, highly reliable and offer high heat transfer rates. This study summarizes the investigation of different types of heat pipe heat recovery systems (HPHRSs). The studies are classified on the basis of the type of the HPHRS. This research is based on 30 years of experience on heat pipe and heat recovery systems that are presented in this study. Copyright , Oxford University ...

  5. Variation of structural damping with response amplitude in piping systems

    International Nuclear Information System (INIS)

    Ware, A.G.

    1986-01-01

    From tests conducted over the last several years, it has become apparent that structural damping is not a single number applicable to all piping systems, but is highly dependent on piping system parameters such as supports, response amplitude, and insulation. As a result, there is considerable scatter in the available data. Furthermore, the relationships between the parameters and damping are often highly complex, interrelated, and difficult to predict. From tests of piping supported by various typical methods, two basic types of energy dissipation in the supports can be observed. The first is friction such as between spring hangers and their housings or in the internal mechanisms of constant force hangers. The second is impacting such as occurs in snubbers, rigid struts, and rod hangers. Overall, these effects lead to a wide variety of possibilities that can occur at low vibration levels and can change with only a slight perturbation of vibration amplitude. This can account for much of the scatter in the data at low strain levels. Thus damping is almost impossible to predict at low amplitudes, and extrapolation of this type data to higher amplitudes is cautioned. However, once strain levels rise above 100 to 200 micro in/in, the damping trend becomes easier to characterize. From the 100 to 200 micro in/in to 800 to 1000 micro in/in range the damping is fairly constant and is induced primarily by the supports. At the upper end of this range a threshold is reached in which damping increases with increasing strain amplitude. Data in the high strain (plastic range) is sparse since the test usually renders the pipe unsuitable for further use. 15 refs

  6. Tensile and fracture properties of primary heat transport system piping material

    International Nuclear Information System (INIS)

    Singh, P.K.; Chattopadhyay, J.; Kushwaha, H.S.

    1997-07-01

    The fracture mechanics calculations in leak-before-break analysis of nuclear piping system require material tensile data and fracture resistance properties in the form of J-R curve. There are large variations in fracture parameters due to variation in chemical composition and process used in making the steel components. Keeping this in view, a comprehensive program has been planned to generate the material data base for primary heat transport system piping using the specimens machined from actual pipes used in service. The material under study are SA333 Gr.6 (base as well as weld) and SA350 LF2 (base). Since the operating temperatures of 500 MWe Indian PHWR PHT system piping range from 260 degC to 304 degC the test temperature chosen are 28 degC, 200 degC, 250 degC and 300 degC. Tensile and compact tension specimens have been fabricated from actual pipe according to ASTM standard. Fracture toughness of base metal has been observed to be higher compared to weld metal in SA333 Gr.6 material for the temperature under consideration. Fracture toughness has been observed to be higher for LC orientation (notch in circumferential direction) compared to CL orientation (notch is in longitudinal direction) for the temperature range under study. Fracture toughness value decreases with increase in temperature for the materials under study. Finally, chemical analysis has been carried out to investigate the reason for high toughness of the material. It has been concluded that low percentage of carbon and nitrogen, low inclusion rating and fine grain size has enhanced the fracture toughness value

  7. BOA: Asbestos pipe-insulation removal robot system. Phase I. Topical report, November 1993--December 1994

    International Nuclear Information System (INIS)

    Schempf, H.; Bares, J.E.

    1995-01-01

    Based on several key design criteria and site visits, we developed a Robot design and built a system which automatically strips the lagging and insulation from the pipes, and encapsulates them under complete vacuum operation. The system can operate on straight runs of piping in horizontal or vertical orientations. Currently we are limited to four-inch diameter piping without obstacles as well as a somewhat laborious emplacement and removal procedure. Experimental results indicated that the current robotic abatement process is sound yet needs to be further expanded and modified. One of the main discoveries was that a longitudinal cut to fully allow the paddles to dig in and compress the insulation off the pipe is essential. Furthermore, a different cutting method might be explored to alleviate the need for a deeper cut and to enable a combination of certain functions such as compression and cutting. Unfortunately due to a damaged mechanism caused by extensive testing, we were unable to perform vertical piping abatement experiments, but foresee no trouble in implementing them in the next proposed Phase. Other encouraging results have BOA removing asbestos at a rate of 4-5 ft./h compared to 3 ft./h for manual removal of asbestos with a 3-person crew. However, we feel confident that we can double the asbestos removal rate by improving cutting speed, and increasing the length of the BOA robot. The containment and vacuum system on BOA is able to achieve the regulatory requirement for airborne fiber emissions of 0.01 fibers/ccm/8-hr. shift. Currently, BOA weighs about 117 pounds which is more than a human is permitted to lift overhead under OSHA requirements (i.e., 25 pounds). We are considering designing the robot into two components (i.e., locomotor section and cutter/removal section) to aid human installation as well as incorporating composite materials. A more detailed list of all the technical modifications is given in this topical report

  8. BOA: Asbestos pipe-insulation removal robot system. Phase I. Topical report, November 1993--December 1994

    Energy Technology Data Exchange (ETDEWEB)

    Schempf, H.; Bares, J.E.

    1995-01-01

    Based on several key design criteria and site visits, we developed a Robot design and built a system which automatically strips the lagging and insulation from the pipes, and encapsulates them under complete vacuum operation. The system can operate on straight runs of piping in horizontal or vertical orientations. Currently we are limited to four-inch diameter piping without obstacles as well as a somewhat laborious emplacement and removal procedure. Experimental results indicated that the current robotic abatement process is sound yet needs to be further expanded and modified. One of the main discoveries was that a longitudinal cut to fully allow the paddles to dig in and compress the insulation off the pipe is essential. Furthermore, a different cutting method might be explored to alleviate the need for a deeper cut and to enable a combination of certain functions such as compression and cutting. Unfortunately due to a damaged mechanism caused by extensive testing, we were unable to perform vertical piping abatement experiments, but foresee no trouble in implementing them in the next proposed Phase. Other encouraging results have BOA removing asbestos at a rate of 4-5 ft./h compared to 3 ft./h for manual removal of asbestos with a 3-person crew. However, we feel confident that we can double the asbestos removal rate by improving cutting speed, and increasing the length of the BOA robot. The containment and vacuum system on BOA is able to achieve the regulatory requirement for airborne fiber emissions of 0.01 fibers/ccm/8-hr. shift. Currently, BOA weighs about 117 pounds which is more than a human is permitted to lift overhead under OSHA requirements (i.e., 25 pounds). We are considering designing the robot into two components (i.e., locomotor section and cutter/removal section) to aid human installation as well as incorporating composite materials. A more detailed list of all the technical modifications is given in this topical report.

  9. Pipe support

    International Nuclear Information System (INIS)

    Pollono, L.P.

    1979-01-01

    A pipe support for high temperature, thin-walled piping runs such as those used in nuclear systems is described. A section of the pipe to be suppported is encircled by a tubular inner member comprised of two walls with an annular space therebetween. Compacted load-bearing thermal insulation is encapsulated within the annular space, and the inner member is clamped to the pipe by a constant clamping force split-ring clamp. The clamp may be connected to pipe hangers which provide desired support for the pipe

  10. SNR coolant system components

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Integrated CAE system for nuclear power plants. Development of piping design check system

    International Nuclear Information System (INIS)

    Narikawa, Noboru; Sato, Teruaki

    1994-01-01

    Toshiba Corporation has developed and operated the integrated CAE system for nuclear power plants, the core of which is the engineering data base to manage accurately and efficiently enormous amount of data on machinery, equipment and piping. As the first step of putting knowledge base system to practical use, piping design check system has been developed. By automatically checking up piping design, this system aims at the prevention of overlooking mistakes, efficient design works and the overall quality improvement of design. This system is based on the thought that it supports designers, and final decision is made by designers. This system is composed of the integrated data base, a two-dimensional CAD system and three-dimensional CAD system. The piping design check system is one of the application systems of the integrated CAE system. Object-oriented programming is the base of the piping design check system, and design knowledge and CAD data are necessary. As to the method of realizing the check system, the flow of piping design, the checkup functions, the checkup of interference and attribute base, and the integration of the system are explained. (K.I)

  12. Program to justify life extension of older nuclear piping systems

    International Nuclear Information System (INIS)

    Burr, T.K.; Dwight, J.E. Jr.; Morton, D.K.

    1991-01-01

    The Idaho National Engineering Laboratory (INEL) has a history of more than 40 years devoted to the operation of nuclear reactors designed for research and experiments. The Advanced Test Reactor (ATR) is one such operating reactor whose mission requires continued operation for an additional 25 years or more. Since the ATR is approaching its design life of twenty years, life extension evaluations have been initiated. Of particular importance are the associated high temperature, high pressure loop piping system supporting in--reactor experiments. Failure of this piping could challenge core safety margins. Since regulatory rules for nuclear power plant life extension are only in the formulation stage, the current technical guidance on this subject provided by the Department of Energy (DOE) or the commercial nuclear industry is incomplete. In the interim, order to assure continued safe operation of this piping beyond its initial design life, a program has been developed to provide the necessary technical justification for life extension. This paper describes a program that establishes Section 11 of the ASME Boiler and Pressure Vessel Code as the governing criteria document, retains B31.1 as the Code of record for Section 11 activities, specifies additional inservice inspection requirements more strict than Section 11, and relies heavily on flaw detection and fracture mechanics evaluations. 18 refs., 2 figs

  13. Implementing An Image Understanding System Architecture Using Pipe

    Science.gov (United States)

    Luck, Randall L.

    1988-03-01

    This paper will describe PIPE and how it can be used to implement an image understanding system. Image understanding is the process of developing a description of an image in order to make decisions about its contents. The tasks of image understanding are generally split into low level vision and high level vision. Low level vision is performed by PIPE -a high performance parallel processor with an architecture specifically designed for processing video images at up to 60 fields per second. High level vision is performed by one of several types of serial or parallel computers - depending on the application. An additional processor called ISMAP performs the conversion from iconic image space to symbolic feature space. ISMAP plugs into one of PIPE's slots and is memory mapped into the high level processor. Thus it forms the high speed link between the low and high level vision processors. The mechanisms for bottom-up, data driven processing and top-down, model driven processing are discussed.

  14. Reliable pipeline repair system for very large pipe size

    Energy Technology Data Exchange (ETDEWEB)

    Charalambides, John N.; Sousa, Alexandre Barreto de [Oceaneering International, Inc., Houston, TX (United States)

    2004-07-01

    The oil and gas industry worldwide has been mainly depending on the long-term reliability of rigid pipelines to ensure the transportation of hydrocarbons, crude oil, gas, fuel, etc. Many other methods are also utilized onshore and offshore (e.g. flexible lines, FPSO's, etc.), but when it comes to the underwater transportation of very high volumes of oil and gas, the industry commonly uses large size rigid pipelines (i.e. steel pipes). Oil and gas operators learned to depend on the long-lasting integrity of these very large pipelines and many times they forget or disregard that even steel pipelines degrade over time and more often that that, they are also susceptible to various forms of damage (minor or major, environmental or external, etc.). Over the recent years the industry had recognized the need of implementing an 'emergency repair plan' to account for such unforeseen events and the oil and gas operators have become 'smarter' by being 'pro-active' in order to ensure 'flow assurance'. When we consider very large diameter steel pipelines such as 42' and 48' nominal pipe size (NPS), the industry worldwide does not provide 'ready-made', 'off-the-shelf' repair hardware that can be easily shipped to the offshore location and effect a major repair within acceptable time frames and avoid substantial profit losses due to 'down-time' in production. The typical time required to establish a solid repair system for large pipe diameters could be as long as six or more months (depending on the availability of raw materials). This paper will present in detail the Emergency Pipeline Repair Systems (EPRS) that Oceaneering successfully designed, manufactured, tested and provided to two major oil and gas operators, located in two different continents (Gulf of Mexico, U.S.A. and Arabian Gulf, U.A.E.), for two different very large pipe sizes (42'' and 48'' Nominal Pipe Sizes

  15. The stress analysis evaluation and pipe support layout for pressurizer discharge system

    International Nuclear Information System (INIS)

    Mao Qing; Wang Wei; Zhang Yixiong

    2000-01-01

    The author presents the stress analysis and evaluation of pipe layout and support adjustment process for Qinshan phase II pressurizer discharge system. Using PDL-SYSPIPE INTERFACE software, the characteristic parameters of the system are gained from 3-D CAD engineering design software PDL and outputted as the input date file format of special pipe stress analysis program SYSPIPE. Based on that, SYSPIPE program fast stress analysis function is applied in adjusting pipe layout , support layout and support types. According to RCC-M standard, the pipe stress analysis and evaluation under deadweight, internal pressure, thermal expansion, seismic, pipe rupture and discharge loads are fulfilled

  16. Component reliability for electronic systems

    CERN Document Server

    Bajenescu, Titu-Marius I

    2010-01-01

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

  17. Pipe/duct system design for tornado missile impact loads

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Wang, S.; Johnson, W., E-mail: whjohnso@bechtel.com

    2014-04-01

    For nuclear power plant life extension projects, it may be convenient and in some instances necessary to locate safety-related steel ducts and pipes outside of the main structures, exposing them to extreme environmental loads such as tornado missile impact. Examples of this application include emergency firewater lines and Control Room vent ducts. A typical exposed commodity run could be comprised of a rectangular or circular cross-section with horizontal and vertical segments supported at variable spans off of roof and wall panels, respectively. Efficient and economical design of such a tornado-impacted duct or pipe system, consisting of the commodity and its supports, must exploit all of the system's capability to absorb the impact energy by deforming plastically to the fullest extent allowable. Energy can be absorbed locally in the vicinity of impact on the commodity, globally through rotation at flexural plastic hinges, and through yielding of the supports. In this paper a simplified NDOF lumped parameter nonlinear analysis methodology is presented and applied to the coupled commodity/support system subjected to tornado impulse loading. The analysis methodology is confirmed using a detailed ANSYS nonlinear finite element model. Optimization of the initial trial design is achieved by progressively decreasing the support resistances, while monitoring the response ductilities throughout the system. Evaluation methodologies are provided for the four types of plastic deformation responses which occur in the system: local response in the immediate vicinity of impact, flexural and membrane response of the sidewall out to one or two times the commodity depth beyond the point of impact, global response of the commodity as a beam spanning between supports, and the shear and flexural response of support. The inelastic responses are evaluated against AISC N690 acceptance criteria (ANSI, 2006), supplemented as appropriate by triaxiality considerations for inelastic

  18. Component Reification in Systems Modelling

    DEFF Research Database (Denmark)

    Bendisposto, Jens; Hallerstede, Stefan

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

  19. Pressure-dependent fragilities for piping components: Pilot study on Davis-Besse Nuclear Power Station

    International Nuclear Information System (INIS)

    Wesley, D.A.; Nakaki, D.K.; Hadidi-Tamjed, H.; Kipp, T.R.

    1990-10-01

    The capacities of four, low-pressure fluid systems to withstand pressures and temperatures above the design levels were established for the Davis-Besse Nuclear Power Station. The results will be used in evaluating the probability of plant damage from Interfacing System Loss of Coolant Accidents (ISLOCA) as part of the probabilistic risk assessment of the Davis-Besse nuclear power station undertaken by EG ampersand G Idaho, Inc. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The probabilities of failure, as a function of internal pressure, are evaluated as well as the variabilities associated with them. Leak rates or leak areas are estimated for the controlling modes of failure. The pressure capacities for the pipes and vessels are evaluated using limit-state analyses for the various failure modes considered. The capacities are dependent on several factors, including the material properties, modeling assumptions, and the postulated failure criteria. The failure modes for gasketed-flange connections, valves, and pumps do not lend themselves to evaluation by conventional structural mechanics techniques and evaluation must rely primarily on the results from ongoing gasket research test programs and available vendor information and test data. 21 refs., 7 figs., 52 tabs

  20. Comparison of multiple support excitation solution techniques for piping systems

    International Nuclear Information System (INIS)

    Sterkel, H.P.; Leimbach, K.R.

    1980-01-01

    Design and analysis of nuclear power plant piping systems exposed to a variety of dynamic loads often require multiple support excitation analysis by modal or direct time integration methods. Both methods have recently been implemented in the computer program KWUROHR for static and dynamic analysis of piping systems, following the previous implementation of the multiple support excitation response spectrum method (see papers K 6/15 and K 6/15a of the SMiRT-4 Conference). The results of multiple support excitation response spectrum analyses can be examined by carrying out the equivalent time history analyses which do not distort the time phase relationship between the excitations at different support points. A frequent point of discussion is multiple versus single support excitation. A single support excitation analysis is computationally straightforward and tends to be on the conservative side, as the numerical results show. A multiple support excitation analysis, however, does not incur much more additional computer cost than the expenditure for an initial static solution involving three times the number, L, of excitation levels, i.e. 3L static load cases. The results are more realistic than those from a single support excitation analysis. A number of typical nuclear plant piping systems have been analyzed using single and multiple support excitation algorithms for: (1) the response spectrum method, (2) the modal time history method via the Wilson, Newmark and Goldberg integration operators and (3) the direct time history method via the Wilson integration operator. Characteristic results are presented to compare the computational quality of all three methods. (orig.)

  1. An overview of erosion corrosion models and reliability assessment for corrosion defects in piping system

    International Nuclear Information System (INIS)

    Srividya, A.; Suresh, H.N.; Verma, A.K.; Gopika, V.; Santosh

    2006-01-01

    Piping systems are part of passive structural elements in power plants. The analysis of the piping systems and their quantification in terms of failure probability is of utmost importance. The piping systems may fail due to various degradation mechanisms like thermal fatigue, erosion-corrosion, stress corrosion cracking and vibration fatigue. On examination of previous results, erosion corrosion was more prevalent and wall thinning is a time dependent phenomenon. The paper is intended to consolidate the work done by various investigators on erosion corrosion in estimating the erosion corrosion rate and reliability predictions. A comparison of various erosion corrosion models is made. The reliability predictions based on remaining strength of corroded pipelines by wall thinning is also attempted. Variables in the limit state functions are modelled using normal distributions and Reliability assessment is carried out using some of the existing failure pressure models. A steady state corrosion rate is assumed to estimate the corrosion defect and First Order Reliability Method (FORM) is used to find the probability of failure associated with corrosion defects over time using the software for Component Reliability evaluation (COMREL). (author)

  2. Component and system simulation models for High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

  3. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  4. Experimental basis for parameters contributing to energy dissipation in piping systems

    International Nuclear Information System (INIS)

    Ibanez, P.; Ware, A.G.

    1985-01-01

    The paper reviews several pipe testing programs to suggest the phenomena causing energy dissipation in piping systems. Such phenomena include material damping, plasticity, collision in gaps and between pipes, water dynamics, insulation straining, coupling slippage, restraints (snubbers, struts, etc.), and pipe/structure interaction. These observations are supported by a large experimental data base. Data are available from in-situ and laboratory tests (pipe diameters up to about 20 inches, response levels from milli-g's to responses causing yielding, and from excitation wave forms including sinusoid, snapback, random, and seismic). A variety of pipe configurations have been tested, including simple, bare, straight sections and complex lines with bends, snubbers, struts, and insulation. Tests have been performed with and without water and at zero to operating pressure. Both light water reactor and LMFBR piping have been tested

  5. Statistical models for the analysis of water distribution system pipe break data

    International Nuclear Information System (INIS)

    Yamijala, Shridhar; Guikema, Seth D.; Brumbelow, Kelly

    2009-01-01

    The deterioration of pipes leading to pipe breaks and leaks in urban water distribution systems is of concern to water utilities throughout the world. Pipe breaks and leaks may result in reduction in the water-carrying capacity of the pipes and contamination of water in the distribution systems. Water utilities incur large expenses in the replacement and rehabilitation of water mains, making it critical to evaluate the current and future condition of the system for maintenance decision-making. This paper compares different statistical regression models proposed in the literature for estimating the reliability of pipes in a water distribution system on the basis of short time histories. The goals of these models are to estimate the likelihood of pipe breaks in the future and determine the parameters that most affect the likelihood of pipe breaks. The data set used for the analysis comes from a major US city, and these data include approximately 85,000 pipe segments with nearly 2500 breaks from 2000 through 2005. The results show that the set of statistical models previously proposed for this problem do not provide good estimates with the test data set. However, logistic generalized linear models do provide good estimates of pipe reliability and can be useful for water utilities in planning pipe inspection and maintenance

  6. Study of a two-pipe chilled beam system for both cooling and heating of office buildings

    Energy Technology Data Exchange (ETDEWEB)

    Norouzi, R. [Univ. of Boraes, Boraes (Sweden); Hultmark, G. [Lindab Comfort A/S, Farum (Denmark); Afshari, A. (ed.); Bergsoee, N.C. [Aalborg Univ.. Statens Byggeforskningsinstitut (SBi), Copenhagen (Denmark)

    2013-05-15

    The main aim of this master thesis was to investigate possibilities and limitations of a new system in active chilled beam application for office buildings. Lindab Comfort A/S pioneered the presented system. The new system use two-pipe system, instead of the conventional active chilled beam four-pipe system for heating and cooling purposes. The Two-Pipe System which is studied in this project use high temperature cooling and low temperature heating with water temperatures of 20 deg. C to 23 deg. C, available for free most of the year. The system can thus take advantage of renewable energy. It was anticipated that a Two-Pipe System application enables transfer of energy from warm spaces to cold spaces while return flows, from cooling and heating beams, are mixed. BSim software was chosen as a simulation tool to model a fictional office building and calculate heating and cooling loads of the building. Moreover, the effect of using outdoor air as a cooling energy source (free cooling) is investigated through five possible scenarios in both the four pipe system and the Two-Pipe System. The calculations served two purposes. Firstly, the effect of energy transfer in the Two-Pipe System were calculated and compared with the four pipe system. Secondly, free cooling effect was calculated in the Two-Pipe System and compared with the four pipe system. The simulation results showed that the energy transfer, as an inherent characteristic in the Two-Pipe System, is able to reduce up to 3 % of annual energy use compared to the four pipe system. Furthermore, different free cooling applications in the Two-Pipe System and the four pipe system respectively showed that the Two-Pipe System requires 7-15 % less total energy than the four pipe system in one year. In addition, the Two-Pipe System can save 18-57 % of annual cooling energy when compared to the four pipe system. (Author)

  7. Acoustic system for pipe rupture monitoring and leak detection

    International Nuclear Information System (INIS)

    Herzog, W.; Jonas, H.

    1982-06-01

    As a safety aspect pipe rupture and leakage effects are of particular interest in nuclear power plants where severe consequences for the reactor may result. Counter measures against postulated pipe breaks and leakages in nuclear power plants are necessary whenever the main safety goals: safe shut-down, safe afterheat removal and retention of radioactivity, are endangered. The requirements to be met by a leak detection system depend on the time available for counter actions. If this time is short so that automatic actions are necessary the German safety criteria for nuclear power plants (Criterion 6.1) require two physically diverse signals to be monitored. One fairly obvious possibility of leak detection is to monitor process parameters (pressure, flow). As a diverse signal physical parameters outside the process may be employed: pressure transients temperature, humidity are principally suitable. In practical application, however, it is difficult to predict these parameters by way of calculation in order to establish the required set-point of the monitoring system. Experimental determination is possible only in special cases. A study of several ways of diverse leak detection methods leads to the very promising acoustic method. We investigated experimentally the feasibility of monitoring the sound created by a leakage. Air borne sound as well as body borne sound was analyzed

  8. Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

    International Nuclear Information System (INIS)

    Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

    2016-01-01

    Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

  9. Experimental investigation of high cycle thermal fatigue in a T-junction piping system

    Energy Technology Data Exchange (ETDEWEB)

    Selvam, P. Karthick; Kulenovic, Rudi; Laurien, Eckart [Stuttgart Univ. (Germany). Inst. of Nuclear Technology and Energy Systems (IKE)

    2015-10-15

    High cycle thermal fatigue damage of structure in the vicinity of T-junction piping systems in nuclear power plants is of importance. Mixing of coolant streams at significant temperature differences causes thermal fluctuations near piping wall leading to gradual thermal degradation. Flow mixing in a T-junction is performed. The determined factors result in bending stresses being imposed on the piping system ('Banana effect').

  10. Design considerations for CRBRP heat transport system piping operating at elevated temperatures

    International Nuclear Information System (INIS)

    Pollono, L.P.; Mello, R.M.

    1979-01-01

    The heat transport system sodium piping for the Clinch River Breeder Reactor Plant (CRBRP) within the reactor containment building must withstand high temperatures for long periods of time. Each phase of the mechanical design process of the piping system is influenced by elevated temperature considerations which include material thermal creep effects, ratchetting caused by rapid temperature transients and stress relaxation, and material degradation effects. The structural design philosophy taken to design the CRBRP piping operating in a high temperature environment is described. The resulting design of the heat transport system piping is presented along with a discussion of special features that resulted from the elevated temperature considerations

  11. Qualification methodologies for mechanical component, I and C, piping using test lab

    International Nuclear Information System (INIS)

    Ichikawa, Toshio

    2001-01-01

    There are many methods of verifying the intensity of a structure, a function, a vibration characteristics, etc. The seismic test which verifies the function during the earthquake of a components simple substance (seismic test which checks durability according to components types). How to verify the analysis technique by the scale model and to check the intensity of plant operating conditions by the scale model results. The model of the same size as the actual plant is created and there is a method of verifying intensity and the function directly. A seismic test is restrained by the frequency of an evaluation objective, and the capability of actuator equipment, and is carried out. Moreover, otherwise, restrictions are the size of a table, actuation power, environment, etc. Here, further examples are introduced, such as evaluation by the examination that combined analysis, experimental test use and analysis, and the experimental test, and the method of proving only by test, and have the seismic check method by test learned in this lecture. Typical examples are explained. Based on the seismic test result carried out with experimental research equipment, how to verify that the required function to components, such as a structure of reactor internals, is maintained at the time of an earthquake is explained. In this case, differences of the simulation environment of the model in. a test, earthquake conditions simulated by shaker table of test conditions and actual plant conditions are important for the evaluation method determination. In nuclear equipment, there is what is required to achieve the static function to hold pressure boundary to the high temperature inside apparatus piping - high-pressure flow, and dynamic functions, such as insertion of a valve, a pump, and a control rod. Moreover, in order to maintain and carry out the safe stop of the safe operation, there is I and C for controlling - supervising these components. In order for this functional maintenance

  12. Vibration analysis for IHTS piping system of LMR conveying hot liquid sodium

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, Hyeong Yeon; Lee, Jae Han

    2001-01-01

    In this paper, the vibration characteristics of IHTS(Intermediate Heat Transfer System) piping system of LMR(Liquid Metal Reactor) conveying hot liquid sodium are investigated to eliminate the pipe supports for economic reasons. To do this, a 3-dimensional straight pipe element and a curved pipe element conveying fluid are formulated using the dynamic stiffness method of the wave approach and coded to be applied to any complex piping system. Using this method, the dynamic characteristics including the natural frequency, the frequency response functions, and the dynamic instability due to the pipe internal flow velocity are analyzed. As one of the design parameters, the vibration energy flow is also analyzed to investigate the disturbance transmission paths for the resonant excitation and the non-resonant excitations

  13. Computer-Aided Design System Development of Fixed Water Distribution of Pipe Irrigation System

    OpenAIRE

    Zhou , Mingyao; Wang , Susheng; Zhang , Zhen; Chen , Lidong

    2010-01-01

    International audience; It is necessary to research a cheap and simple fixed water distribution device according to the current situation of the technology of low-pressure pipe irrigation. This article proposed a fixed water distribution device with round table based on the analysis of the hydraulic characteristics of low-pressure pipe irrigation systems. The simulation of FLUENT and GAMBIT software conducted that the flow of this structure was steady with a low head loss comparing to other t...

  14. Seismic analysis response factors and design margins of piping systems

    International Nuclear Information System (INIS)

    Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

    1985-01-01

    The objective of the simplified methods project of the Seismic Safety Margins Research Program is to develop a simplified seismic risk methodology for general use. The goal is to reduce seismic PRA costs to roughly 60 man-months over a 6 to 8 month period, without compromising the quality of the product. To achieve the goal, it is necessary to simplify the calculational procedure of the seismic response. The response factor approach serves this purpose. The response factor relates the median level response to the design data. Through a literature survey, we identified the various seismic analysis methods adopted in the U.S. nuclear industry for the piping system. A series of seismic response calculations was performed. The response factors and their variabilities for each method of analysis were computed. A sensitivity study of the effect of piping damping, in-structure response spectra envelop method, and analysis method was conducted. In addition, design margins, which relate the best-estimate response to the design data, are also presented

  15. Degradation of safety injection system and containment spray piping and tank fracture toughness analysis

    International Nuclear Information System (INIS)

    Douglas, A.; Doubel, P.; Wicker, C.

    2011-01-01

    Extensive stress corrosion cracking (SCC), induced by the marine environment and the presence of high residual stresses arising from the respective manufacturing processes has been encountered in the safety injection system piping (RIS), containment spray system piping (EAS) and reactor and spent fuel storage tank (PTR), or refuelling water storage tank (RWST) of the Koeberg plant. Type 304L steels from the RIS system and replacement components for the RIS and RWST systems have been subject to mechanical and fracture toughness testing. The following conclusions have been drawn. -) The piping sections of both the original and replacement components exhibit residual cold work. The level of cold work imparted to the piping and elbow have been estimated to be 2, 2 to 3, 9% and 5, 7 to 7, 3% respectively. -) Re-annealing produces different responses in type 304L as a function of prior cold work level. Re-annealing of material cold worked to low levels i.e. 3.5% maintain the cold worked level of UTS but can exhibit 0, 2% PS. levels below that of the mill annealed condition. There is the potential for the ASTM A312 minimum 0, 2% level to be breached. At higher levels of cold work i.e. 7% re-annealing results in extensive grain growth, a significant reduction in 0, 2% PS from the mill annealed condition and the recovery of the UTS to the mill annealed level. -) Cold work at the levels obtained significantly reduces the SOL initiation toughness Ji. The reduction in toughness can be greater than 50%. The resistance to ductile crack propagation, dJ/da, remains unchanged at least up to 5 % cold work. -) The defect assessment for the RIS/EAS systems have used highly conservative values of initiation toughness such that no crack initiation would occur under the loading conditions considered and in a non-hostile environment. -) Under the marine environment to which the RIS/EAS components are still subjected, the limiting criterion for operation of the RIS/EAS system remains a

  16. Pipe support program at Pickering

    International Nuclear Information System (INIS)

    Sahazizian, L.A.; Jazic, Z.

    1997-01-01

    This paper describes the pipe support program at Pickering. The program addresses the highest priority in operating nuclear generating stations, safety. We present the need: safety, the process: managed and strategic, and the result: assurance of critical piping integrity. In the past, surveillance programs periodically inspected some systems, equipment, and individual components. This comprehensive program is based on a managed process that assesses risk to identify critical piping systems and supports and to develop a strategy for surveillance and maintenance. The strategy addresses all critical piping supports. Successful implementation of the program has provided assurance of critical piping and support integrity and has contributed to decreasing probability of pipe failure, reducing risk to worker and public safety, improving configuration management, and reducing probability of production losses. (author)

  17. Balance-of-plant options for the Heat-Pipe Power System

    International Nuclear Information System (INIS)

    Berte, M.; Capell, B.

    1997-09-01

    The Heat-Pipe Power System (HPS) is a near-term, low-cost space fission power system with the potential for utilizing various option for balance-of-plant options. The following options have been studied: a low-power thermoelectric design (14-kWe output), a small Brayton cycle system (60--75 kWe), and a large Brayton cycle system (250 kWe). These systems were analyzed on a preliminary basis, including mass, volume, and structure calculations. These analyses have shown that the HPS system can provide power outputs from 10--250 kWe with specific powers of ∼ 14 W/kg for a 14-kWe model to ∼ 100 W/kg for a 250-kWe model. The system designs considered in this study utilize a common component base to permit easy expansion and development

  18. Thermal fatigue evaluation of piping system Tee-connections

    International Nuclear Information System (INIS)

    Metzner, K.J.; Braillard, O.; Faidy, C.; Marcelles, I.; Solin, J.; Stumpfrock, L.

    2004-01-01

    Thermal fatigue is one significant long-term degradation mechanism nuclear power plants (NPP), in particular, as operating plants become older and life time extension activities have been initiated. In general, the common thermal fatigue issues are understood and controlled by plant instrumentation systems. However, incidents in some plants indicate that certain piping system Tees are susceptible to turbulent temperature mixing effects that cannot be adequately monitored by common thermocouple instrumentation. The THERFAT project has been initiated to advance the accuracy and reliability of thermal fatigue load determination in engineering tools and research oriented approaches to outline a science based practical methodology for managing thermal fatigue risks in Tee-connections susceptible to high cyclic thermal fatigue. (orig.)

  19. Application of LBB to high energy piping systems in operating PWR

    Energy Technology Data Exchange (ETDEWEB)

    Swamy, S.A.; Bhowmick, D.C. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01

    The amendment to General Design Criterion 4 allows exclusion, from the design basis, of dynamic effects associated with high energy pipe rupture by application of leak-before-break (LBB) technology. This new approach has resulted in substantial financial savings to utilities when applied to the Pressurized Water Reactor (PWR) primary loop piping and auxiliary piping systems made of stainless steel material. To date majority of applications pertain to piping systems in operating plants. Various steps of evaluation associated with the LBB application to an operating plant are described in this paper.

  20. Relative conservatisms of combination methods used in response spectrum analyses of nuclear piping systems

    International Nuclear Information System (INIS)

    Gupta, S.; Kustu, O.; Jhaveri, D.P.; Blume, J.A.

    1983-01-01

    The paper presents the conclusions of a comprehensive study that investigated the relative conservatisms represented by various combination techniques. Two approaches were taken for the study, producing mutually consistent results. In the first, 20 representative nuclear piping systems were systematically analyzed using the response spectrum method. The total response was obtained using nine different combination methods. One procedure, using the SRSS method for combining spatial components of response and the 10% method for combining the responses of different modes (which is currently acceptable to the U.S. NRC), was the standard for comparison. Responses computed by the other methods were normalized to this standard method. These response ratios were then used to develop cumulative frequency-distribution curves, which were used to establish the relative conservatism of the methods in a probabilistic sense. In the second approach, 30 single-degree-of-freedom (SDOF) systems that represent different modes of hypothetical piping systems and have natural frequencies varying from 1 Hz to 30 Hz, were analyzed for 276 sets of three-component recorded ground motion. A set of hypothetical systems assuming a variety of modes and frequency ranges was developed. The responses of these systems were computed from the responses of the SDOF systems by combining the spatial response components by algebraic summation and the individual mode responses by the Navy method, or combining both spatial and modal response components using the SRSS method. Probability density functions and cumulative distribution functions were developed for the ratio of the responses obtained by both methods. (orig./HP)

  1. Pipe damping

    International Nuclear Information System (INIS)

    Ware, A.G.

    1985-01-01

    Studies are being conducted at the Idaho National Engineering Laboratory to determine whether an increase in the damping values used in seismic structural analyses of nuclear piping systems is justified. Increasing the allowable damping would allow fewer piping supports which could lead to safer, more reliable, and less costly piping systems. Test data from availble literature were examined to determine the important parameters contributing to piping system damping, and each was investigated in separate-effects tests. From the combined results a world pipe damping data bank was established and multiple regression analyses performed to assess the relative contributions of the various parameters. The program is being extended to determine damping applicable to higher frequency (33 to 100 Hz) fluid-induced loadings. The goals of the program are to establish a methodology for predicting piping system damping and to recommend revised guidelines for the damping values to be included in analyses

  2. Gravity-assist heat pipes for thermal control systems

    International Nuclear Information System (INIS)

    Deverall, J.E.; Keddy, E.S.; Kemme, J.E.; Phillips, J.R.

    1975-06-01

    Sodium heat pipes, operating in the gravity-assist mode, have been incorporated into irradiation capsules to provide a means for establishing and controlling a desired specimen temperature. Investigations were made of new wick structures for potassium heat pipes to operate at lower temperatures and higher heat transfer rates, and a helical trough wick structure was developed with an improved heat transfer capability in the temperature range of interest. Test results of these heat pipes led to the study of a new heat pipe limit which had not previously been considered. (12 references) (U.S.)

  3. Performance predictions and measurements for space-power-system heat pipes

    International Nuclear Information System (INIS)

    Prenger, F.C. Jr.

    1981-01-01

    High temperature liquid metal heat pipes designed for space power systems have been analyzed and tested. Three wick designs are discussed and a design rationale for the heat pipe is provided. Test results on a molybdenum, annular wick heat pipe are presented. Performance limitations due to boiling and capillary limits are presented. There is evidence that the vapor flow in the adiabatic section is turbulent and that the transition Reynolds number is 4000

  4. Main steam system piping response under safety/relief valve opening events

    International Nuclear Information System (INIS)

    Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

    1980-01-01

    The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

  5. Proceedings of transient thermal-hydraulics and coupled vessel and piping system responses 1991

    International Nuclear Information System (INIS)

    Wang, G.Y.; Shin, Y.W.; Moody, F.J.

    1991-01-01

    This book reports on transient thermal-hydraulics and coupled vessel and piping system responses. Topics covered include: nuclear power plant containment designs; analysis of control rods; gate closure of hydraulic turbines; and shock wave solutions for steam water mixtures in piping systems

  6. Dynamic response of piping system on rack structure with gaps and frictions

    International Nuclear Information System (INIS)

    Kobayashi, Hiroe; Yoshida, Misutoyo; Ochi, Yoshio

    1989-01-01

    In the seismic design of a piping system on a rack structure, the interaction between the piping system and the rack structure must be evaluated under the condition that the rack structure is not stiff and heavy enough compared with the piping system. Moreover, there are local nonlinearities due to the gap and friction between the piping system and the rack structure. This paper presents the influence of the interaction and the local nonlinearities upon the seismic response by numerical study and a vibration test using a shaking table. In the numerical study, the piping system and the rack structure were represented by the three degrees of freedom mass-spring model taking a vibration mode of the piping system into account. The nonlinearities due to gap and friction were defined as a function of motion and treated as the pseudo force vector (additional applied force) in an equation of motion. From the results of the numerical study and the vibration test, it was clarified that seismic response of both the rack structure and the piping system is reduced by gap and friction. Moreover, the piping system and rack structure can be represented by the three degrees of freedom mass spring model. And the local nonlinearities can be treated by the pseudo force in an equation of motion. (orig.)

  7. Development of thermal fatigue evaluation methods of piping systems

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Itoh, Takamoto; Okazaki, Masakazu; Okuda, Yukihiko; Kamaya, Masayuki; Nakamura, Akira; Nakamura, Hitoshi; Machida, Hideo; Matsumoto, Masaaki

    2013-01-01

    Nuclear piping has various kinds of thermal fatigue failure modes. Main causes of thermal loads are structural responses to fluid temperature changes during plant operation. These phenomena have complex mechanisms and so many patterns, that their problems still occur even though well-known issues. To prevent thermal fatigue due to above thermal loads, the JSME guideline is adopted. Both thermal load and fatigue failure mechanism have been investigated and summarized into the knowledgebase. Based on above knowledge, improved methods for the JSME guideline and Numerical simulation methods for thermal fatigue evaluation were studied. Furthermore, probabilistic failure analysis approach with main influence parameters were investigated to be applied for the plant system safety. (author)

  8. Piping benchmark problems for the ABB/CE System 80+ Standardized Plant

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1994-07-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the ABB/Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the System 80+ standard design. It will be required that the combined license licensees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System 80+ piping benchmark is a time history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactor is an advanced PWR type

  9. Fracture evaluation of a crack in the service water piping system to an emergency diesel generator

    International Nuclear Information System (INIS)

    Rudland, D.; Scott, P.; Rahman, S.; Wilkowski, G.

    1995-01-01

    A pipe fracture experiment was conducted on a section of 6-inch nominal diameter pipe which was degraded by microbiologically induced corrosion (MIC) at a circumferential girth weld. The pipe was a section of one of the service water piping system to one of the emergency diesel generators at the Haddam Neck (Connecticut Yankee) plant. The experimental results will help validate future ASME Section XI pipe flaw evaluation criteria for other than Class 1 piping. A critical aspect of this experiment was an assessment of the degree of conservatism embodied in the ASME definition of flaw size. The ASME flaw size definition assumes a rectangular shaped, constant depth flaw with a depth equal to its maximum depth for its entire length. Since most service flaws are very irregular in shape, this definition can be very conservative. Alternative equivalent flaw size definitions for irregular shaped flaws are explored in this paper. (author). 7 refs., 2 figs., 4 tabs

  10. Heat-pipe development for the SPAR space-power system

    International Nuclear Information System (INIS)

    Ranken, W.A.

    1981-01-01

    The SPAR space power system design is based on a high temperature fast spectrum nuclear reactor that furnishes heat to a thermoelectric conversion system to generate an electrical power output of 100 kW/sub (e)/. An important feature of this design is the use of alkali metal heat pipes to provide redundant, reliable, and low-loss heat transfer at high temperature. Three sets of heat pipes are used in the system. These include sodium/molybdenum heat pipes to transfer heat from the reactor core to the conversion system, potassium/niobium heat pipes to couple the conversion system to the radiator in a redundant manner, and potassium/titanium heat pipes to distribute rejected heat throughout the radiator surface. The designs of these units are discussed and fabrication methods and testing results are described. 12 figures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-22

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

  12. Computation of the effect of pipe plasticity on pressure-pulse propagation in a fluid system

    International Nuclear Information System (INIS)

    Youngdahl, C.K.; Kot, C.A.

    1975-04-01

    A simple computational model is developed for incorporating the effect of elastic-plastic deformation of piping on pressure-transient propagation in a fluid system. A computer program (PLWV) is described that incorporates this structural interaction model into a one-dimensional method-of-characteristics procedure for fluid-hammer analysis. Computed results are shown to be in good agreement with available experimental data. The most significant effect of plastic deformation is to limit the peak pressure of a pulse leaving a pipe to approximately the yield pressure of the pipe, if the pipe is sufficiently long. 7 references. (U.S.)

  13. Choice of insulation standard for pipe networks in 4th generation district heating systems

    DEFF Research Database (Denmark)

    Lund, Rasmus Søgaard; Mohammadi, Soma

    2016-01-01

    and smart gas grids. Improving DH pipes by improving the insulation standard results in decreasing the heat and temperature losses from the pipe networks. When reducing heat losses from DH pipes, there is a trade-off between the increasing cost of pipe insulation and the associated savings in the heat...... supply system. This study presents a methodology to describe this balance for a specific case and its application for the case of Denmark. The methodology presented consists of a techno-economic analysis in two steps. In the first step, a DH grid model is used to assess the reduction in grid losses...

  14. The conservatism of the net-section stress procedure for predicting the failure of cracked piping systems: The effect of crack location on the degree of conservatism

    International Nuclear Information System (INIS)

    Smith, E.

    1993-01-01

    Interest in the integrity of cracked piping systems fabricated from ductile materials has been motivated, in large part, by the technological problem of intergranular stress corrosion cracking of Type 304 stainless steel piping in boiling water nuclear reactor piping systems. The failure of cracked steel piping is often predicted by assuming that failure conforms to a net-section stress criterion using as input an appropriate value for the critical net-section stress together with a knowledge of the anticipated loadings. The stresses at the cracked section are usually calculated via a purely elastic analysis based on the piping being uncracked. However because the piping is built-in at the ends into a larger component, and since the onset of crack extension requires some plastic deformation, use of the net-section stress approach can give overly conservative failure predictions. An earlier paper has quantified the extent of this conservatism, and has shown how it depends on the material ductility and the elastic flexibility of a piping system. Using the results of analyses for simple model systems, the present paper shows that, for the same cracked section geometry, the degree of conservatism is markedly influenced by the location of the cracked section within the system

  15. Efficient heat recovery: Integrated circuit systems and heat pipes; Gezielte Waermerueckgewinnung: KV-Systeme und Waermerohr

    Energy Technology Data Exchange (ETDEWEB)

    Kaup, C. [Howatherm, Bruecken (Germany)

    1995-09-18

    Integrated circuit systems and heat pipes are both known to be low-efficiency systems, but this shortcoming can be eliminated by constructive measures. (orig.) [Deutsch] Die beiden Verfahren - Kreislaufverbundsystem und das Waermerohr - sind als WRG-Systeme mit geringen Wirkungsgraden bekannt. Doch dieser Nachteil kann durch spezielle Konstruktionsmassnahmen eliminiert werden. (orig.)

  16. The modularization construction of piping system installation in AP1000 plant

    International Nuclear Information System (INIS)

    Lu Song; Wang Yuan; Wei Junming

    2012-01-01

    Modularization construction is the main technique used in AP1000 plants, the piping Modularization installation will impact directly to the module construction as the important part of the Modularization construction. After the piping system has took the modularization design in AP1000 plants, some installation works of piping system has moved from the site to fabrication shop. With improving the construction quality and minimizing the time frame of project, the critical paths can be optimized. This paper has analyzed the risk and challenge that met during the modularization construction period of piping systems though introducing the characteristic of modularization construction for AP1000 piping systems, and get construction experiences from the First AP1000 plants in the world, then it will be the firmly basics for the wide application of modularization construction in the future. (authors)

  17. Research on the ITOC based scheduling system for ship piping production

    Science.gov (United States)

    Li, Rui; Liu, Yu-Jun; Hamada, Kunihiro

    2010-12-01

    Manufacturing of ship piping systems is one of the major production activities in shipbuilding. The schedule of pipe production has an important impact on the master schedule of shipbuilding. In this research, the ITOC concept was introduced to solve the scheduling problems of a piping factory, and an intelligent scheduling system was developed. The system, in which a product model, an operation model, a factory model, and a knowledge database of piping production were integrated, automated the planning process and production scheduling. Details of the above points were discussed. Moreover, an application of the system in a piping factory, which achieved a higher level of performance as measured by tardiness, lead time, and inventory, was demonstrated.

  18. Component wall thinning and a corrosion-erosion monitoring system

    International Nuclear Information System (INIS)

    Bogard, T.; Batt, T.; Roarty, D.

    1989-01-01

    Since a 1986 incident involving failure of a piping elbow due to erosion-corrosion, the electric utility industry has been actively developing technology for implementing long term programs to address corrosion-erosion. This paper describes a typical corrosion-erosion monitoring program, the types of non-destructive examinations (NDE) performed on components, and the extensive NDE data obtained when the program is applied to components in a power plant. To facilitate evaluation of the NDE data on components, an automated NDE data manipulation and data display system is advisable and perhaps necessary due to the large amounts of NDE data typically obtained during a program. Such a comprehensive corrosion-erosion monitoring system (CEMS) needs to be integral with methods for selection of inspection locations and perform NDE data analysis to help in replace, repair, or run decisions. The structure for one CEMS is described which uses IBM PC compatible hardware and a set of software addressing most data evaluation and decision making needs. CEMS features include automated input/output for typical NDE devices, database structuring, graphics outputs including color 2-D or 3-D contour plots of components, trending and predictive evaluations for future inspection planning, EC severity determination, integration of piping isometrics and component properties, and desktop publishing capabilities

  19. Basic concepts about application of dual vibration absorbers to seismic design of nuclear piping systems

    International Nuclear Information System (INIS)

    Hara, F.; Seto, K.

    1987-01-01

    The design value of damping for nuclear piping systems is a vital parameter in ensuring safety in nuclear plants during large earthquakes. Many experiments and on-site tests have been undertaken in nuclear-industry developed countries to determine rational design values. However damping value in nuclear piping systems is so strongly influenced by many piping parameters that it shows a tremendous dispersion in its experimental values. A new trend has recently appeared in designing nuclear pipings, where they attempt to use a device to absorb vibration energy induced by seismic excitation. A typical device is an energy absorbing device, made of a special material having a high capacity of plasticity, which is installed between the piping and the support. This paper deals with the basic study of application of dual vibration absorbers to nuclear piping systems to accomplish high damping value and reduce consequently seismic response at resonance frequencies of a piping system, showing their effectiveness from not only numerical calculation but also experimental evaluation of the vibration responses in a 3D model piping system equipped with dual two vibration absorbers

  20. Study on seismic design margin based upon inelastic shaking test of the piping and support system

    International Nuclear Information System (INIS)

    Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

    2009-01-01

    In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

  1. Fabrication of mechanical components and piping design for Brazilian nuclear reactors

    International Nuclear Information System (INIS)

    Deppe, L.O.

    1987-01-01

    The supply of Brazilian equipment and piping design for Angra 2 (and Angra 3 in some cases) have reached an advanced status in spite of the continuous outside difficulties which affect these nuclear power plants. The achieved quality is similar to the quality achieved in foreign countries and the nationalization program foreseen in 1975 is being largely surpassed. In this paper the actual situation is presented as well as the future perspectives. (Author) [pt

  2. Seismic proving test of ultimate piping strength (current status of preliminary tests)

    International Nuclear Information System (INIS)

    Suzuki, K.; Namita, Y.; Abe, H.; Ichihashi, I.; Suzuki, K.; Ishiwata, M.; Fujiwaka, T.; Yokota, H.

    2001-01-01

    In 1998 Fiscal Year, the 6 year program of piping tests was initiated with the following objectives: i) to clarify the elasto-plastic response and ultimate strength of nuclear piping, ii) to ascertain the seismic safety margin of the current seismic design code for piping, and iii) to assess new allowable stress rules. In order to resolve extensive technical issues before proceeding on to the seismic proving test of a large-scale piping system, a series of preliminary tests of materials, piping components and simplified piping systems is intended. In this paper, the current status of the material tests and the piping component tests is reported. (author)

  3. Development of an integrated heat pipe-thermal storage system for a solar receiver

    Science.gov (United States)

    Keddy, E. S.; Sena, J. T.; Merrigan, M. A.; Heidenreich, G.; Johnson, S.

    1987-07-01

    The Organic Rankine Cycle (ORC) Solar Dynamic Power System (SDPS) is one of the candidates for Space Station prime power application. In the low Earth orbit of the Space Station approximately 34 minutes of the 94-minute orbital period is spent in eclipse with no solar energy input to the power system. For this period the SDPS will use thermal energy storage (TES) material to provide a constant power output. An integrated heat-pipe thermal storage receiver system is being developed as part of the ORC-SDPS solar receiver. This system incorporates potassium heat pipe elements to absorb and transfer the solar energy within the receiver cavity. The heat pipes contain the TES canisters within the potassium vapor space with the toluene heater tube used as the condenser region of the heat pipe. During the insolation period of the Earth orbit, solar energy is delivered to the heat pipe in the ORC-SDPS receiver cavity. The heat pipe transforms the non-uniform solar flux incident in the heat pipe surface within the receiver cavity to an essentially uniform flux at the potassium vapor condensation interface in the heat pipe. During solar insolation, part of the thermal energy is delivered to the heater tube and the balance is stored in the TES units. During the eclipse period of the orbit, the balance stored in the TES units is transferred by the potassium vapor to the toluene heater tube.

  4. Development of an integrated heat pipe-thermal storage system for a solar receiver

    Science.gov (United States)

    Keddy, E. S.; Sena, J. T.; Merrigan, M. A.; Heidenreich, G.; Johnson, S.

    1987-01-01

    The Organic Rankine Cycle (ORC) Solar Dynamic Power System (SDPS) is one of the candidates for Space Station prime power application. In the low Earth orbit of the Space Station approximately 34 minutes of the 94-minute orbital period is spent in eclipse with no solar energy input to the power system. For this period the SDPS will use thermal energy storage (TES) material to provide a constant power output. An integrated heat-pipe thermal storage receiver system is being developed as part of the ORC-SDPS solar receiver. This system incorporates potassium heat pipe elements to absorb and transfer the solar energy within the receiver cavity. The heat pipes contain the TES canisters within the potassium vapor space with the toluene heater tube used as the condenser region of the heat pipe. During the insolation period of the Earth orbit, solar energy is delivered to the heat pipe in the ORC-SDPS receiver cavity. The heat pipe transforms the non-uniform solar flux incident in the heat pipe surface within the receiver cavity to an essentially uniform flux at the potassium vapor condensation interface in the heat pipe. During solar insolation, part of the thermal energy is delivered to the heater tube and the balance is stored in the TES units. During the eclipse period of the orbit, the balance stored in the TES units is transferred by the potassium vapor to the toluene heater tube.

  5. Development of bore tools for pipe welding and cutting

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji

    1998-01-01

    In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

  6. Development of bore tools for pipe welding and cutting

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

  7. Leaks in the internal water supply piping systems

    Directory of Open Access Journals (Sweden)

    Orlov Evgeniy Vladimirovich

    2015-03-01

    Full Text Available Great water losses in the internal plumbing of a building lead to the waste of money for a fence, purification and supply of water volumes in excess. This does not support the concept of water conservation and resource saving lying today in the basis of any building’s construction having plumbing. Leakage means unplanned of water losses systems in domestic water supply systems (hot or cold as a result of impaired integrity, complicating the operation of a system and leading to high costs of repair and equipment restoration. A large number of leaks occur in old buildings, where the regulatory service life of pipelines has come to an end, and the scheduled repair for some reason has not been conducted. Steel pipelines are used in the systems without any protection from corrosion and they get out of order. Leakages in new houses are also not uncommon. They usually occur as a result of low-quality adjustment of the system by workers. It also important to note the absence of certain skills of plumbers, who don’t conduct the inspections of in-house systems in time. Sometimes also the residents themselves forget to keep their pipeline systems and water fittings in their apartment in good condition. Plumbers are not systematically invited for preventive examinations to detect possible leaks in the domestic plumbing. The amount of unproductive losses increases while simultaneous use of valve tenants, and at the increase of the number of residents in the building. Water leaks in the system depend on the amount of water system piping damages, and damages of other elements, for example, water valves, connections, etc. The pressure in the leak area also plays an important role.

  8. Statistics of Shared Components in Complex Component Systems

    Science.gov (United States)

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

    2018-04-01

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

  9. Statistics of Shared Components in Complex Component Systems

    Directory of Open Access Journals (Sweden)

    Andrea Mazzolini

    2018-04-01

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

  10. Coherent systems with multistate components

    International Nuclear Information System (INIS)

    Caldarola, L.

    1980-01-01

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

  11. Technical note on drainage systems:design of pipes and detention facilities for rainwater

    OpenAIRE

    Bentzen, Thomas Ruby

    2014-01-01

    This technical note will present simple but widely used methods for the design of drainage systems. The note will primarily deal with surface water (rainwater) which on a satisfactorily way should be transport into the drainage system. Traditional two types of sewer systems exist: A combined system, where rainwater and sewage is transported in the same pipe, and a separate system where the two types of water are transported in individual pipe. This note will only focus on the separate rain/st...

  12. Experimental investigation on an integrated thermal management system with heat pipe heat exchanger for electric vehicle

    OpenAIRE

    Zou, Huiming; Wang, Wei; Zhang, Guiying; Qin, Fei; Tian, Changqing; Yan, Yuying

    2016-01-01

    An integrated thermal management system combining a heat pipe battery cooling/preheating system with the heat pump air conditioning system is presented to fulfill the comprehensive energy utilization for electric vehicles. A test bench with battery heat pipe heat exchanger and heat pump air conditioning for a regular five-chair electric car is set up to research the performance of this integrated system under different working conditions. The investigation results show that as the system is d...

  13. Heat pipe heat exchangers in heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

    Stulc, P; Vasiliev, L L; Kiseljev, V G; Matvejev, Ju N

    1985-01-01

    The results of combined research and development activities of the National Research Institute for Machine Design, Prague, C.S.S.R. and the Institute for Heat and Mass Transfer, Minsk, U.S.S.R. concerning intensification heat pipes used in heat pipe heat exchangers are presented. This sort of research has been occasioned by increased interest in heat power economy trying to utilise waste heat produced by various technological processes. The developed heat pipes are deployed in construction of air-air, gas-air or gas-gas heat recovery exchangers in the field of air-engineering and air-conditioning. (author).

  14. Novel developments in linear modal description of piping system dynamic behavior

    International Nuclear Information System (INIS)

    Revesz, Z.

    1989-01-01

    Novel developments in dynamic analysis of piping systems are described. The ASME BPV Codes, 1986 describes methods that are considered as adequate to analyze piping systems under dynamic loading, and also states that the method described in the codes are not the only acceptable ones. With straightforward application of the principles and methods laid down in the code novel numerical techniques can be developed. These techniques allow to obtain correct, conservative estimates of the piping system response and to reduce the computed stresses the same time. Beyond that, the particular algorithm which is presented is also suitable to analyze systems which include non-linear (viscous) damping elements

  15. Pipe damping

    International Nuclear Information System (INIS)

    Ware, A.G.; Arendts, J.G.

    1984-01-01

    A program has been developed to assess the available piping damping data, to generate additional data and conduct seperate effects tests, and to establish a plan for reporting and storing future test results into a data bank. This effort is providing some of the basis for developing higher allowable damping values for piping seismic analyses, which will potentially permit removal of a considerable number of piping supports, particularly snubbers. This in turn will lead to more flexible piping systems which will be less susceptible to thermal cracking, will be easier to maintain and inspect, as well as less costly

  16. Vibration analysis of the piping system using the modal analysis method, 1

    International Nuclear Information System (INIS)

    Fujikawa, Takeshi; Kurohashi, Michiya; Inoue, Yoshio

    1975-01-01

    Modal analysis method was developed for the vibration analysis of piping system in nuclear or chemical plants, with finite element theory, and verified by sinusoidal vibration method. The natural vibration equation for pipings was derived with stiffness, attenuation and mass matrices, and eigenvalues are obtained with usual method, then the forced vibration equation for pipings was derived with the same manner, and the special solutions are given by modal method from the eigenvalues of the natural vibration equation. Three simple piping models (one, two and three dimensional) were made, and the natural vibration frequency was measured with forced input from an electrical dynamic shaker and a sound speaker. The experimental values of natural vibration frequency showed good agreement with the results by the analytical method. Therefore the theoretical approach for piping system vibration was proved to be valid. (Iwase, T.)

  17. Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

    International Nuclear Information System (INIS)

    1975-10-01

    The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

  18. Component Control System for a Vehicle

    Science.gov (United States)

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

    2016-01-01

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

  19. Pipe connector

    International Nuclear Information System (INIS)

    Sullivan, T.E.; Pardini, J.A.

    1978-01-01

    A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated

  20. Modelization of cooling system components

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  1. Sodium Variable Conductance Heat Pipe for Radioisotope Stirling Systems

    Science.gov (United States)

    Tarau, Calin; Anderson, William G.; Walker, Kara

    2009-01-01

    In a Stirling radioisotope system, heat must continually be removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. Normally, the Stirling convertor provides this cooling. If the converter stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS, and also ending the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) has been designed to allow multiple stops and restarts of the Stirling convertor in an Advanced Stirling Radioisotope Generator (ASRG). When the Stirling convertor is turned off, the VCHP will activate when the temperatures rises 30 C above the setpoint temperature. A prototype VCHP with sodium as the working fluid was fabricated and tested in both gravity aided and against gravity conditions for a nominal heater head temperature of 790 C. The results show very good agreement with the predictions and validate the model. The gas front was located at the exit of the reservoir when heater head temperature was 790 C while cooling was ON, simulating an operating Advanced Stirling Converter (ASC). When cooling stopped, the temperature increased by 30 C, allowing the gas front to move past the radiator, which transferred the heat to the case. After resuming the cooling flow, the front returned at the initial location turning OFF the VCHP. The against gravity working conditions showed a colder reservoir and faster transients.

  2. PSA-2, Stress Analysis, Thermal Expansion and Loads in Multi Anchor Piping System

    Energy Technology Data Exchange (ETDEWEB)

    Nickols, A N [Codes Coordinator, Atomics International, P. O. Box 309, Canoga Park, California 91304 (United States)

    1975-03-01

    1 - Description of problem or function: PSA2 computes the reactions and stresses caused by thermal expansion and loads in a multi-anchor piping system which may contain loops and may be partially restrained at any point in any direction. 2 - Method of solution: The linear equations for the statically indeterminate pipe system are set up by a generalization of Brock's matrix method. By a systematic use of linear transforms, the matrix of the system of linear equations can be obtained by incidence algebra in the form of a symmetric banded matrix. 2 - Restrictions on the complexity of the problem - Maximum of: 36 sections. 3 - Unusual features of the program - PSA2 takes into account: (a) elasticity of the attachment of the pipe to the foundation, (b) restraints on pipe displacements by anchors and intermediate partial constraints of linear type, (c) given constant forces and moments acting upon the pipe system, (d) thermal expansion, (e) any geometrical structure of the pipe system, (f) several cases of stressing per pipe system, and (g) both metric and English units.

  3. PSA-2, Stress Analysis, Thermal Expansion and Loads in Multi Anchor Piping System

    International Nuclear Information System (INIS)

    Nickols, A.N.

    1975-01-01

    1 - Description of problem or function: PSA2 computes the reactions and stresses caused by thermal expansion and loads in a multi-anchor piping system which may contain loops and may be partially restrained at any point in any direction. 2 - Method of solution: The linear equations for the statically indeterminate pipe system are set up by a generalization of Brock's matrix method. By a systematic use of linear transforms, the matrix of the system of linear equations can be obtained by incidence algebra in the form of a symmetric banded matrix. 2 - Restrictions on the complexity of the problem - Maximum of: 36 sections. 3 - Unusual features of the program - PSA2 takes into account: (a) elasticity of the attachment of the pipe to the foundation, (b) restraints on pipe displacements by anchors and intermediate partial constraints of linear type, (c) given constant forces and moments acting upon the pipe system, (d) thermal expansion, (e) any geometrical structure of the pipe system, (f) several cases of stressing per pipe system, and (g) both metric and English units

  4. Practical method of dynamic analysis considering coupling effects between equipment and piping systems

    International Nuclear Information System (INIS)

    Koyanagi, Ryoichi

    1984-01-01

    Many piping systems are supported by flexible structures or attached to thin shell walls so it is very important to consider the dynamic coupling effects between these systems in dynamic analysis. This paper presents a practical method of dynamic analysis of an individual system considering the dynamic coupling effects of coupled equipment-piping systems. In this method, dynamic responses are calculated by using the modal information which is obtained from the other analysis for associative structure. Analytical results for the complete model and of this method for an individual system are presented in the piping-supporting structure system and a piping-shell system. From the comparison of these results, it shows that this method is accurate, useful and economically applicable to the dynamic analysis of large model. (author)

  5. Review and assessment of research relevant to design aspects of nuclear power plant piping systems. Final report

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Maxey, W.A.; Eiber, R.J.

    1977-06-01

    Significant research on piping systems is evaluated, and the correlation of that research with design practices is presented. The objective is to quantify the research/design practices in terms of the reliability of piping used in nuclear power plants

  6. An experimental study of damping characteristics with emphasis on insulation for nuclear power plant piping system (Seismic Damping Ratio Evaluation Program)

    International Nuclear Information System (INIS)

    Shibata, H.; Ito, M.; Hayashi, T.; Chiba, T.; Kobayashi, H.; Kitamura, K.; Ando, K.; Koyanagi, R.

    1981-01-01

    To clarify the damping characteristics and mechanism in nuclear power plant piping systems, the study group was established and conducted to study SDREP (Seismic Damping Ratio Evaluation Program). As the Phase II of this study, vibration tests were conducted to investigate factors which might contribute to damping characteristics of piping systems. These tests are composed of the next three model tests: 1) The component damping characteristics test of thermal insulator 2) The simplified piping model test 3) The scale model test. In these tests, we studied damping characteristics with emphasis on thermal insulator (mainly calcium silicate insulator). The acceleartion level of pipings is the same as that of the actual seismic response. The excitation was by sinusoidal sweep method using the shaking table and by free vibration method using snapback. (orig./RW)

  7. Generic safety evaluation report regarding integrity of BWR scram system piping

    International Nuclear Information System (INIS)

    1981-08-01

    Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service

  8. Development of Structural Health Monitoring System for pipes in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Eom, H. S.; Choi, Y. C.; Shin, S. H.; Youn, D. B.; Park, J. H.

    2010-01-01

    Structural health monitoring (SHM) has becoming an important issue in the maintenance of various structures such as large steel plates, vessels, and pipes in nuclear power plants. There are important factors to be considered in developing an SHM system. With consideration of these factors, we have developed a computerized multi-channel ultrasonic system that can handle array transducers and generate a high-power pulse for online SHM of the plates and pipes. The proposed system is compact but has all the necessary functions for SHM of important structure such as pipes and plates in a NPP

  9. Prediction on corrosion rate of pipe in nuclear power system based on optimized grey theory

    International Nuclear Information System (INIS)

    Chen Yonghong; Zhang Dafa; Chen Dengke; Jiang Wei

    2007-01-01

    For the prediction of corrosion rate of pipe in nuclear power system, the pre- diction error from the grey theory is greater, so a new method, optimized grey theory was presented in the paper. A comparison among predicted results from present and other methods was carried out, and it is seem that optimized grey theory is correct and effective for the prediction of corrosion rate of pipe in nuclear power system, and it provides a fundamental basis for the maintenance of pipe in nuclear power system. (authors)

  10. Component codification and identification systems

    International Nuclear Information System (INIS)

    Pannenbaecker, K.

    1977-01-01

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

  11. Mechanized ultrasonic examination of piping systems in nuclear power plants

    International Nuclear Information System (INIS)

    Edelmann, X.; Pfister, O.; Allidi, F.

    1988-01-01

    The success of mechanized ultrasonic examination applied on welds in piping systems in nuclear power plants is highly dependent on its careful preparation. From the development of an adequate examination technique to its implementation on site, many problems are to be solved. This is especially the case when dealing with austenitic welds or dissimilar metal welds. In addition to the specific needs for examination technique based on material properties and requirements for minimum flaw size detection, accessibility and radiation aspects have to be considered. A crew of skilled and highly trained examination personnel is required. Experience in various nuclear power plants, - BWR's and PWR's of different designs - has shown, that even difficult examination problems can be successfully solved, provided that there is a good preparation. The necessary step by step proceeding is illustrated by examples concerning mechanized examination. Preservice inspections and in-service inspections with specific requirements, due to the types of flaws to be found or the type of material concerned, are discussed

  12. Managing the Cost of Plant Piping System Leakage

    International Nuclear Information System (INIS)

    Jenco, John M.; Keck, Donna R.; Johnson, Gary L.

    2002-01-01

    Recent studies have shown that the average annual cost impact of external piping system leakage on commercial nuclear plant operations and maintenance can easily range into the millions of dollars for each reactor unit. Evidence suggests that this significant O and M cost reduction opportunity has largely been overlooked, due to the number of diverse line items and budget areas affected. Results released last year from an EPRI pilot study of more than a dozen reactor units at seven plant sites operated by multiple utilities found that the average annual cost impact was indeed around $1.6 million per year per unit. Subsequent field experience has also demonstrated that an effective fluid leak management program can substantially reduce these costs within the first three years of implementation. This paper presents the general cost impact research results from various studies, outlines key elements of an effective plant fluid leak management program, discusses important implementation issues, and presents results from case studies covering different utility approaches to developing and implementing an effective fluid leak management program. Actual cost data will be included where appropriate. (authors)

  13. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

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

  14. A regulatory perspective on appropriate seismic loading stress criteria for advanced light water reactor piping systems

    International Nuclear Information System (INIS)

    Terao, D.

    1995-01-01

    In the foregoing sections, the author has discussed the NRC staff's perspective on the evolving seismic design criteria for piping systems. He also addressed the need for developing seismic loading stress criteria and provided several recommendations and considerations for ensuring piping functional capability, pressure integrity, and structural integrity. Overall, the general consensus in the NRC staff is that in the past several years, many initiatives have been developed and implemented by the industry and the NRC staff to reduce the excessive conservatisms that might have existed in nuclear piping system design criteria. The regulations, regulatory guides, and Standard Review Plan have been (or are currently in the process of being) revised to reflect these initiatives in an effort to produce requirements and guidelines that will continue to result in a safe and practical design of piping systems. However, further proposals to reduce margins are continually being submitted to the ASME Boiler and Pressure Vessel Code and the NRC for review and approval. Improvements to the piping seismic design criteria are always encouraged, but there is a point at which the benefits might be outweighed by drawbacks. Because of this rapidly evolving situation the need exists for the industry and the NRC staff to develop a course of action to ensure that piping seismic design criteria for future ALWR plants will result in piping system designs that provide adequate safety margins and practical designs at a reasonable cost

  15. Heat pipes as perspective base elements of heat recovery in heat supply and ventilating systems

    Directory of Open Access Journals (Sweden)

    Matveev Andrey

    2017-01-01

    Full Text Available Thermotechnical characteristics of heat pipes are considered as high-efficient heat-transfer devices, which can provide energy-saving technologies for heat supply and ventilating systems and for different branches of industry. Thermotechnical and working (”performance capability” characteristics of heat pipes are investigated. By ”performance capability” of heat pipes and heat-transfer devices on heat pipes we mean the system state, where it can perform set functions and keep parameter values (thermal power, conductivity, thermal resistance, heat-transfer coefficient, temperature level and differential, etc. within the regulations of standardized specifications. The article presents theoretical and experimental methods of «gaslock» length determination on noncondensable gases during long-lasting tests of ammonia heat pipes made of aluminum shape АS – КRА 7.5 – R1 (alloy АD – 31. The paper gives results of research of thermotechnical characteristics of heat pipes in horizontal and vertical states (separate and as a set part while using different systems of thermal insulation. The obtained results of thermotechnical and resource tests show the advantages of ammonia heat pipes as basic elements for heat exchanger design in heating and ventilation systems.

  16. RIMAP demonstration project. Risk-based life management of piping system in power plant Heilbronn

    International Nuclear Information System (INIS)

    Bareiss, J.; Buck, P.; Matschecko, B.; Jovanovic, A.; Balos, D.; Perunicic, M.

    2004-01-01

    In the framework of EU project RIMAP [Risk Based Inspection and Maintenance Procedures for European Industry (2000)] a new European Guideline for optimized risk based maintenance and inspection planning of industrial plants (RBLM, Risk Based Life Management) is being developed. The RIMAP project consists of the three clustered projects: - development (RTD); - demonstration (DEMO): - thematic network (TN). Current work and future, planned work in RIMAP demonstration project on applications of the RIMAP methodology in power plants are presented briefly in the first part of the paper. Also presented in the paper are the results of a preliminary analysis of piping system in power plant Heilbronn using the concept of risk-based monitoring as part of overall concept of risk-based life management. Shortly the following issues are discussed in the paper: - identification of critical components; - application of a multilevel risk analysis (...from 'screening' to 'detailed analysis'); - determination of PoF (Probability of Failure); - determination of CoF (Consequence of Failure); - optimation of inspection and maintenance plan. From our experience with the application of the RIMAP methodology the following conclusions can be drawn: The use of risk-based methods in inspection and maintenance of piping systems in power plants gives transparency to the decision making process and gives an optimized maintenance policy based on current state of the components. The results of the work clearly show the power of the proposed method for concentration on critical items: out of 64 monitored components 5 were selected for intermediate analysis and only 1 for the detailed analysis (probabilistic high temperature fracture mechanics)

  17. Performance study of heat-pipe solar photovoltaic/thermal heat pump system

    International Nuclear Information System (INIS)

    Chen, Hongbing; Zhang, Lei; Jie, Pengfei; Xiong, Yaxuan; Xu, Peng; Zhai, Huixing

    2017-01-01

    Highlights: • The testing device of HPS PV/T heat pump system was established by a finished product of PV panel. • A detailed mathematical model of heat pump was established to investigate the performance of each component. • The dynamic and static method was combined to solve the mathematical model of HPS PV/T heat pump system. • The HPS PV/T heat pump system was optimized by the mathematical model. • The influence of six factors on the performance of HPS PV/T heat pump system was analyzed. - Abstract: A heat-pipe solar (HPS) photovoltaic/thermal (PV/T) heat pump system, combining HPS PV/T collector with heat pump, is proposed in this paper. The HPS PV/T collector integrates heat pipes with PV panel, which can simultaneously generate electricity and thermal energy. The extracted heat from HPS PV/T collector can be used by heat pump, and then the photoelectric conversion efficiency is substantially improved because of the low temperature of PV cells. A mathematical model of the system is established in this paper. The model consists of a dynamic distributed parameter model of the HPS PV/T collection system and a quasi-steady state distributed parameter model of the heat pump. The mathematical model is validated by testing data, and the dynamic performance of the HPS PV/T heat pump system is discussed based on the validated model. Using the mathematical model, a reasonable accuracy in predicting the system’s dynamic performance with a relative error within ±15.0% can be obtained. The capacity of heat pump and the number of HPS collectors are optimized to improve the system performance based on the mathematical model. Six working modes are proposed and discussed to investigate the effect of solar radiation, ambient temperature, supply water temperature in condenser, PV packing factor, heat pipe pitch and PV backboard absorptivity on system performance by the validated model. It is found that the increase of solar radiation, ambient temperature and PV

  18. An evaluation of an operating BWR piping system damping during earthquake by applying auto regressive analysis

    International Nuclear Information System (INIS)

    Kitada, Y.; Makiguchi, M.; Komori, A.; Ichiki, T.

    1985-01-01

    The records of three earthquakes which had induced significant earthquake response to the piping system were obtained with the earthquake observation system. In the present paper, first, the eigenvalue analysis results for the natural piping system based on the piping support (boundary) conditions are described and second, the frequency and the damping factor evaluation results for each vibrational mode are described. In the present study, the Auto Regressive (AR) analysis method is used in the evaluation of natural frequencies and damping factors. The AR analysis applied here has a capability of direct evaluation of natural frequencies and damping factors from earthquake records observed on a piping system without any information on the input motions to the system. (orig./HP)

  19. BOA: Asbestos pipe-insulation removal robot system, Phase 2. Topical report, January--June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Schempf, H.; Bares, J.E.

    1995-06-01

    This report explored the regulatory impact and cost-benefit of a robotic thermal asbestos pipe-insulation removal system over the current manual abatement work practice. The authors are currently in the second phase of a two-phase program to develop a robotic asbestos abatement system, comprised of a ground-based support system (including vacuum, fluid delivery, computing/electronics/power, and other subsystems) and several on-pipe removal units, each sized to handle pipes within a given diameter range. The intent of this study was to (i) aid in developing design and operational criteria for the overall system to maximize cost-efficiency, and (ii) to determine the commercial potential of a robotic pipe-insulation abatement system.

  20. Analysis of two-phase flow induced vibrations in perpendiculary supported U-type piping systems

    International Nuclear Information System (INIS)

    Hiramatsu, Tsutomu; Komura, Yoshiaki; Ito, Atsushi.

    1984-01-01

    The perpose of this analysis is to predict the vibration level of a pipe conveying a two-phase flowing fluid. Experiments were carried out with a perpendiculary supported U-type piping system, conveying an air-water two-phase flow in a steady state condition. Fluctuation signals are observed by a void signal sensor, and power spectral densities and probability density functions are obtained from the void signals. Theoretical studies using FEM and an estimation of the exciting forces from the PSD of void signals, provided a good predictional estimation of vibration responses of the piping system. (author)

  1. Feasibility study of inside automatic welding system of cooling pipe of divertors for FER

    International Nuclear Information System (INIS)

    Yoshizawa, S.; Adachi, J.; Morishita, H.; Kakudate, S.; Taguchi, H.; Tada, E.

    1995-01-01

    In order to replace divertors for FER, cooling pipes of divertors should be cut and welded since they are too long to be replaced with divertors via horizontal maintenance ports. An inside cutting and welding system is also required because of an accessibility to pipes. A combination of an inside disc-cutting machine and an inside TIG-welding machine has been proposed as a candidate of the systems. We have made tests to confirm possibility to weld pipes which were cut with the disc-cutting machine. Possibility of welding has been proven. The tests result is described in the paper. (orig.)

  2. Heat pipe cooling system for underground, radioactive waste storage tanks

    International Nuclear Information System (INIS)

    Cooper, K.C.; Prenger, F.C.

    1980-02-01

    An array of 37 heat pipes inserted through the central hole at the top of a radioactive waste storage tank will remove 100,000 Btu/h with a heat sink of 70 0 F atmospheric air. Heat transfer inside the tank to the heat pipe is by natural convection. Heat rejection to outside air utilizes a blower to force air past the heat pipe condenser. The heat pipe evaporator section is axially finned, and is constructed of stainless steel. The working fluid is ammonia. The finned pipes are individually shrouded and extend 35 ft down into the tank air space. The hot tank air enters the shroud at the top of the tank and flows downward as it is cooled, with the resulting increased density furnishing the pressure difference for circulation. The cooled air discharges at the center of the tank above the sludge surface, flows radially outward, and picks up heat from the radioactive sludge. At the tank wall the heated air rises and then flows inward to comple the cycle

  3. The magnet components database system

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. The magnet components database system

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  5. Hydraulic simulation of the systems of a nuclear power plant for charges calculation in piping

    International Nuclear Information System (INIS)

    Masriera, N.

    1990-01-01

    This work presents a general description of the methodology used by the ENACE S.A. Fluids Working Group for hydraulics simulation of a nuclear power plant system for the calculation charges in piping. (Author) [es

  6. Transient heat pipe investigations for space power systems

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Keddy, E.S.; Sena, J.T.

    1985-01-01

    A 4-meter long, high temperature, high power, molybdenum-lithium heat pipe has been fabricated and tested in transient and steady state operation at temperatures to 1500 K. Maximum power throughput during the tests was approximately 37 kW/cm 2 for the 1.4 cm diameter vapor space of the annular wick heat pipe. The evaporator flux density for the tests was 150.0 W/cm 2 over a length of 40 cm. Condenser length was approximately 3.0 m with radiant heat rejection from the condenser to a coaxial, water cooled radiation calorimeter. A variable radiation shield, controllable from the outside of the vacuum enclosure, was used to vary the load on the heat pipe during the tests. 1 ref., 9 figs

  7. Structural dynamics and fracture mechanics calculations of the behaviour of a DN 425 test piping system subjected to transient loading by water hammer

    International Nuclear Information System (INIS)

    Kussmaul, K.; Kobes, E.; Diem, H.; Schrammel, D.; Brosi, S.

    1994-01-01

    Within the scope of the German HDR safety programme, several tests were carried out to investigate transient pipe loading initiated by a simulated double-ended guillotine break event, and subsequent closure of a feedwater check valve (water hammer, blow-down). Numerical analyses by means of finite element programmes were performed in parallel to the experiments. Using water hammer tests of a DN 425 piping system with predamaged components, the procedure of such analyses will be demonstrated. The results are presented, beginning with structural dynamic calculations of the undamaged piping; followed by coupling of structural dynamics and fracture mechanics computations with simple flaw elements (line spring); and finishing with costly three-dimensional fracture mechanics analyses. A good description of the real piping behaviour can be made by the numerical methods, even in the case of high plastification processes. ((orig.))

  8. Coupling component systems towards systems of systems

    OpenAIRE

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

    2008-01-01

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

  9. Advanced concepts, analysis approaches and criteria for nuclear piping system design

    International Nuclear Information System (INIS)

    Tang, H.T.; Tagart, S.W. Jr.; Tang, Y.K.

    1992-01-01

    Recent research in piping system design and analysis has resulted in advancements on damping values, independent support motion (ISM), static coefficient method, simplified inelastic method and ASME code criteria changes. In the support area, passive type of supports such as energy-absorbing device and gap stopper have been developed. These advancements provide bases for improved and cost-effective design of future nuclear piping systems. (author)

  10. IPIRG-2 task 1 - pipe system experiments with circumferential cracks in straight-pipe locations. Final report, September 1991--November 1995

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.; Olson, R.; Marschall, C.; Rudland, D. [and others

    1997-02-01

    This report presents the results from Task 1 of the Second International Piping Integrity Research Group (IPIRG-2) program. The IPIRG-2 program is an international group program managed by the US Nuclear Regulatory Commission (US NRC) and funded by a consortium of organizations from 15 nations including: Bulgaria, Canada, Czech Republic, France, Hungary, Italy, Japan, Republic of Korea, Lithuania, Republic of China, Slovak Republic, Sweden, Switzerland, the United Kingdom, and the United States. The objective of the program was to build on the results of the IPIRG-1 and other related programs by extending the state-of-the-art in pipe fracture technology through the development of data needed to verify engineering methods for assessing the integrity of nuclear power plant piping systems that contain defects. The IPIRG-2 program included five main tasks: Task 1 - Pipe System Experiments with Flaws in Straight Pipe and Welds Task 2 - Fracture of Flawed Fittings Task 3 - Cyclic and Dynamic Load Effects on Fracture Toughness Task 4 - Resolution of Issues From IPIRG-1 and Related Programs Task 5 - Information Exchange Seminars and Workshops, and Program Management. The scope of this report is to present the results from the experiments and analyses associated with Task 1 (Pipe System Experiments with Flaws in Straight Pipe and Welds). The rationale and objectives of this task are discussed after a brief review of experimental data which existed after the IPIRG-1 program.

  11. IPIRG-2 task 1 - pipe system experiments with circumferential cracks in straight-pipe locations. Final report, September 1991--November 1995

    International Nuclear Information System (INIS)

    Scott, P.; Olson, R.; Marschall, C.; Rudland, D.

    1997-02-01

    This report presents the results from Task 1 of the Second International Piping Integrity Research Group (IPIRG-2) program. The IPIRG-2 program is an international group program managed by the US Nuclear Regulatory Commission (US NRC) and funded by a consortium of organizations from 15 nations including: Bulgaria, Canada, Czech Republic, France, Hungary, Italy, Japan, Republic of Korea, Lithuania, Republic of China, Slovak Republic, Sweden, Switzerland, the United Kingdom, and the United States. The objective of the program was to build on the results of the IPIRG-1 and other related programs by extending the state-of-the-art in pipe fracture technology through the development of data needed to verify engineering methods for assessing the integrity of nuclear power plant piping systems that contain defects. The IPIRG-2 program included five main tasks: Task 1 - Pipe System Experiments with Flaws in Straight Pipe and Welds Task 2 - Fracture of Flawed Fittings Task 3 - Cyclic and Dynamic Load Effects on Fracture Toughness Task 4 - Resolution of Issues From IPIRG-1 and Related Programs Task 5 - Information Exchange Seminars and Workshops, and Program Management. The scope of this report is to present the results from the experiments and analyses associated with Task 1 (Pipe System Experiments with Flaws in Straight Pipe and Welds). The rationale and objectives of this task are discussed after a brief review of experimental data which existed after the IPIRG-1 program

  12. A new desalination system using a combination of heat pipe, evacuated tube and parabolic trough collector

    International Nuclear Information System (INIS)

    Jafari Mosleh, H.; Jahangiri Mamouri, S.; Shafii, M.B.; Hakim Sima, A.

    2015-01-01

    Highlights: • A new desalination uses a combination of heat pipe and parabolic trough collector. • A twin-glass evacuated tube is used to decrease the thermal losses from heat pipe. • Adding oil into the space between heat pipe and tube collector enhances the yield. • The yield and efficiency reach up to 0.933 kg/(m 2 h) and 65.2%, respectively. - Abstract: The solar collectors have been commonly used in desalination systems. Recent investigations show that the use of a linear parabolic trough collector in solar stills can improve the efficiency of a desalination system. In this work, a combination of a heat pipe and a twin-glass evacuated tube collector is utilized with a parabolic trough collector. Results show that the rate of production and efficiency can reach to 0.27 kg/(m 2 h) and 22.1% when aluminum conducting foils are used in the space between the heat pipe and the twin-glass evacuated tube collector to transfer heat from the tube collector to the heat pipe. When oil is used as a medium for the transfer of heat, filling the space between heat pipe and twin-glass evacuated tube collector, the production and efficiency can increase to 0.933 kg/(m 2 h) and 65.2%, respectively

  13. 46 CFR 154.503 - Piping and piping system components: Protection from movement.

    Science.gov (United States)

    2010-10-01

    ... movement. 154.503 Section 154.503 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN...: Protection from movement. Where thermal movement and movements of the cargo tank and the hull structure may... must be protected from movement by: (a) Offsets; (b) Loops; (c) Bends; (d) Mechanical expansion joints...

  14. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Bain, R.A.; Testa, M.F.

    1994-01-01

    Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

  15. Failure behavior of a pipe system with a circumferentially orientated flaw - analytical and experimental investigations

    International Nuclear Information System (INIS)

    Mikkola, T.P.J.; Diem, H.; Blind, D.; Hunger, H.

    1989-01-01

    At the german HDR-test-facility a pipe failure experiment was performed at a fullsize feedwater piping system under operating conditions of T=240 0 C, p=10.6 MPa and with an elevated oxygen content in the pressure medium. The loading was internal pressure and a cyclic varying bending moment with an R-ratio of 0.5. The in form of a circumferentially orientated notch initially weakened piping system failed after a total number of 4773 loaded cycles with different frequencies in form of a small leak. The analyses of the fracture surface indicated the strongly growing influence of corrosion effects on the crack propagation rate with decreasing loading frequency. The cyclic crack growth and the leak-before-break behavior of the piping system could be explained on the basis of results of finite element calculations using ADINA-code. (orig.)

  16. BOA II: Asbestos Pipe-Insulation Removal Robot System. Innovative Technology Summary Report

    International Nuclear Information System (INIS)

    None

    2001-01-01

    The objective of this task is to develop and demonstrate a mechanical, asbestos-removal system that can be remotely operated without a containment area. The technology, known as BOA, consists of a pipe-crawler removal head and a boom vehicle system with dual robots. BOA's removal head can be remotely placed on the outside of the pipe and can crawl along the pipe, removing lagging and insulation. The lagging and insulation is cut using a hybrid endmill water-jet cutter and then diced into 2-inch cube sections of ACM. These ACM sections are then removed from the pipe using a set of blasting fan- spray nozzles, vacuumed off through a vacuum hose, and bagged. Careful attention to vacuum and entrapment air flow ensures that the system can operate without a containment area while meeting local and federal standards for fiber count

  17. Development of methodologies for coupled water-hammer analysis of piping systems and supports

    International Nuclear Information System (INIS)

    Kamil, H.; Gantayat, A.; Attia, A.; Goulding, H.

    1983-01-01

    The paper presents the results of an investigation on the development of methodologies for coupled water-hammer analyses. The study was conducted because the present analytical methods for calculation of loads on piping systems and supports resulting from water-hammer phenomena are overly conservative. This is mainly because the methods do not usually include interaction between the fluid and the piping and thus predict high loads on piping systems and supports. The objective of the investigation presented in this paper was to develop methodologies for coupled water-hammer analyses, including fluid-structure interaction effects, to be able to obtain realistic loads on piping systems and supports, resulting in production of more economical designs. (orig./RW)

  18. The development of a practical pipe auto-routing system in a shipbuilding CAD environment using network optimization

    Directory of Open Access Journals (Sweden)

    Shin-Hyung Kim

    2013-09-01

    Full Text Available An automatic pipe routing system is proposed and implemented. Generally, the pipe routing design as a part of the shipbuilding process requires a considerable number of man hours due to the complexity which comes from physical and operational constraints and the crucial influence on outfitting construction productivity. Therefore, the automation of pipe routing design operations and processes has always been one of the most important goals for improvements in shipbuilding design. The proposed system is applied to a pipe routing design in the engine room space of a commercial ship. The effectiveness of this system is verified as a reasonable form of support for pipe routing design jobs. The automatic routing result of this system can serve as a good basis model in the initial stages of pipe routing design, allowing the designer to reduce their design lead time significantly. As a result, the design productivity overall can be improved with this automatic pipe routing system.

  19. The development of a practical pipe auto-routing system in a shipbuilding CAD environment using network optimization

    Science.gov (United States)

    Kim, Shin-Hyung; Ruy, Won-Sun; Jang, Beom Seon

    2013-09-01

    An automatic pipe routing system is proposed and implemented. Generally, the pipe routing design as a part of the shipbuilding process requires a considerable number of man hours due to the complexity which comes from physical and operational constraints and the crucial influence on outfitting construction productivity. Therefore, the automation of pipe routing design operations and processes has always been one of the most important goals for improvements in shipbuilding design. The proposed system is applied to a pipe routing design in the engine room space of a commercial ship. The effectiveness of this system is verified as a reasonable form of support for pipe routing design jobs. The automatic routing result of this system can serve as a good basis model in the initial stages of pipe routing design, allowing the designer to reduce their design lead time significantly. As a result, the design productivity overall can be improved with this automatic pipe routing system

  20. Theoretical and experimental study on dynamic responses of piping systems with combined dampers

    International Nuclear Information System (INIS)

    Gershtein, M.; Fridman, Ya.; Perelmiter, A.

    1996-01-01

    Vibrations of pipelines transporting fluids, gases, and granular materials are excited by the air flow, internal pressure pulsation, or seismic ground motion. The susceptibility of oil and gas pipelines to seismic damage has been demonstrated in earthquakes everywhere around the world. Devices for above-ground pipelines and piping systems vibration suppression with combination of dry friction and viscous energy dissipation are developed by AVIBRA, Shear deformation of viscous-elastic material in these devices occurs prior to interfacial slip. The way to account this phenomenon is to model the damper as an elastic-viscous element in series with an ideal Coulomb dry friction element. The harmonic balance method was applied to obtain an equivalent viscous damping constant for a combined damper. Iteration process was used to predict a dynamic response of a piping system with combined dampers subjected to sinusoidal excitation. Every iteration step was based on ANSYS procedures. Time integration of systems with hysteretic friction models presents computational difficulties. Some examples of dynamic responses of piping systems were analyzed by a time integration procedure for finite-element models. Combined dry friction-viscous dissipation dampers were tested on a piping model under harmonic excitation. It was clarified that combined dampers are very effective to reduce dynamic response. The seismic response of the piping system with combined dampers was calculated using time history finite-element analysis. The excellent effectiveness of AVIBRA combined dampers for aseismic design and retrofitting of pipelines and piping systems was confirmed by the analysis

  1. Comparative performance of passive devices for piping system under seismic excitation

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Praveen, E-mail: pra_veen74@rediffmail.com [Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India); Jangid, R.S. [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai, 400076 (India); Reddy, G.R. [Bhabha Atomic Research Centre, Trombay, Mumbai, 400085 (India)

    2016-03-15

    Highlights: • Correlated the analytical results obtained from the proposed analytical procedures with experimental results in the case of XPD. • Substantial reduction of the seismic response of piping system with passive devices is observed. • Significant increase in the modal damping of the piping system is noted. • There exist an optimum parameters of the passive devices. • Good amount of energy dissipation is observed by using passive devices. - Abstract: Among several passive control devices, X-plate damper, viscous damper, visco-elastic damper, tuned mass damper and multiple tuned mass dampers are popular and used to mitigate the seismic response in the 3-D piping system. In the present paper detailed studies are made to see the effectiveness of the dampers when used in 3-D piping system subjected to artificial earthquake with increasing amplitudes. The analytical results obtained using Wen's model are compared with the corresponding experimental results available which indicated a good match with the proposed analytical procedure for the X-plate dampers. It is observed that there is significant reduction in the seismic response of interest like relative displacement, acceleration and the support reaction of the piping system with passive devices. In general, the passive devices under particular optimum parameters such as stiffness and damping are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping system.

  2. RIMAP demonstration project. Pat. 1: Risk based life management of piping system in power plant Heilbronn

    International Nuclear Information System (INIS)

    Bareiss, J.; Puck, P.; Matschecko, B.; Jovanovic, A.; Balos, D.; Perunicic

    2003-01-01

    In the framework of EU project RIMAP [1] a new European Guideline for optimized risk based maintenance and inspection planning of industrial plants (RBLM - Risk Based Life Management) is being developed. The RIMAP project consists of the three clustered projects: development (RTD), demonstration (DEMO) and thematic network (TN). Current work and future, planned work in RIMAP demonstration project on applications of the RIMAP methodology in power plants are presented briefly in the first part of the paper. Also presented in the paper are the results of a preliminary analysis of piping system in power plant Heilbronn using the concept of risk-based monitoring as part of overall concept of risk-based life management. Shortly the following issues are discussed in the paper: identification of critical components, application of a multilevel risk analysis (..from ''screening'' to ''detailed analysis''), determination of PoF - Probability of Failure, determination of COF - Consequence of Failure and optimation of inspection and maintenance plan. (orig.)

  3. 24 CFR 3280.705 - Gas piping systems.

    Science.gov (United States)

    2010-04-01

    ... upstream of the connection. (3) The connection(s) may be made by a listed quick disconnect device which... separated. (4) The flexible connector, direct plumbing pipe, or “quick disconnect” device shall be provided... disconnect device is installed, a 3 inch by 13/4 inch minimum size tag made of etched, metal-stamped or...

  4. Evaluation of LBB margin of nuclear piping systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun [Seoul Nationl Univ., Seoul (Korea, Republic of); Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-04-15

    Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material.

  5. Evaluation of LBB margin of nuclear piping systems

    International Nuclear Information System (INIS)

    Hwang, Il Soon; Kim, Ji Hyeon; Oh, Yeong Jin; Lim, Jun; Kim, In Seob; Kim, Yong Seon; Lee, Joo Seok

    1999-04-01

    Most of previous elastic-plastic fracture studies for LBB assessment of low alloy steel piping have been focused on base metals and weld metals. In contract, the heat affected zone of welded pipe has not been studied in detail primarily because the size of heat affected zone in welded pipe os too small to make specimens for mechanical properties measurement. When structural members are joined by welding, the base metal is heated to its melting point and then cooled rapidly. As a result of this very severe thermal cycle, mechanical properties in the heat affected zone can be degraded by grain coarsening, the precipitation and the segregation of trace impurities. In this study, a thermal and microstructural analysis is performed, and mechanical properties are measured for the weld heat affected zone of SA106Gr.C low allowed piping steel. In addition, inter critical annealing treatment. in two-phase (alpha+gamma) region was performed to investigate the possibilities of improving the toughness and reducing dynamic strain aging (DSA) susceptibility for giving allowable LBB safety margins. From the results, intercritical annealing is shown to give a smaller ductility loss due to DSA than the case of as-received material. Furthermore, the intercritical annealing was able to increase the impact toughness by a factor of 1.5 compared to the as-received material

  6. Remediation System Evaluation, Northwest Pipe and Casing Site

    Science.gov (United States)

    The Northwest Pipe and Casing Site is located in Clackamas, Oregon, approximately 20 miles southeastof Portland. The site consists of approximately 53 acres, and has historically been divided into two parcels(Parcel A to the north and Parcel B to the..

  7. Welding Penetration Control of Fixed Pipe in TIG Welding Using Fuzzy Inference System

    Science.gov (United States)

    Baskoro, Ario Sunar; Kabutomori, Masashi; Suga, Yasuo

    This paper presents a study on welding penetration control of fixed pipe in Tungsten Inert Gas (TIG) welding using fuzzy inference system. The welding penetration control is essential to the production quality welds with a specified geometry. For pipe welding using constant arc current and welding speed, the bead width becomes wider as the circumferential welding of small diameter pipes progresses. Having welded pipe in fixed position, obviously, the excessive arc current yields burn through of metals; in contrary, insufficient arc current produces imperfect welding. In order to avoid these errors and to obtain the uniform weld bead over the entire circumference of the pipe, the welding conditions should be controlled as the welding proceeds. This research studies the intelligent welding process of aluminum alloy pipe 6063S-T5 in fixed position using the AC welding machine. The monitoring system used a charge-coupled device (CCD) camera to monitor backside image of molten pool. The captured image was processed to recognize the edge of molten pool by image processing algorithm. Simulation of welding control using fuzzy inference system was constructed to simulate the welding control process. The simulation result shows that fuzzy controller was suitable for controlling the welding speed and appropriate to be implemented into the welding system. A series of experiments was conducted to evaluate the performance of the fuzzy controller. The experimental results show the effectiveness of the control system that is confirmed by sound welds.

  8. PEP cooling water systems and underground piped utilities design criteria report

    International Nuclear Information System (INIS)

    Hall, F.; Robbins, D.

    1975-10-01

    This paper discusses the cooling systems required by the PEP Storage Ring. Particular topics discussed are: Cooling tower systems, RF cavity and vacuum chamber LCW cooling systems, klystron and ring magnet LLW cooling systems, Injection magnet LCW Cooling Systems; PEP interaction area detector LCW Cooling Systems; and underground piped utilities. 1 ref., 20 figs

  9. Requirements Report Computer Software System for a Semi-Automatic Pipe Handling System and Fabrication Facility

    National Research Council Canada - National Science Library

    1980-01-01

    .... This report is to present the requirements of the computer software that must be developed to create Pipe Detail Drawings and to support the processing of the Pipe Detail Drawings through the Pipe Shop...

  10. Leak before break evaluation for main steam piping system made of SA106 Gr.C

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Kyoung Mo; Jee, Kye Kwang; Pyo, Chang Ryul; Ra, In Sik [Korea Power Engineering Company, Seoul (Korea, Republic of)

    1997-04-01

    The basis of the leak before break (LBB) concept is to demonstrate that piping will leak significantly before a double ended guillotine break (DEGB) occurs. This is demonstrated by quantifying and evaluating the leak process and prescribing safe shutdown of the plant on the basis of the monitored leak rate. The application of LBB for power plant design has reduced plant cost while improving plant integrity. Several evaluations employing LBB analysis on system piping based on DEGB design have been completed. However, the application of LBB on main steam (MS) piping, which is LBB applicable piping, has not been performed due to several uncertainties associated with occurrence of steam hammer and dynamic strain aging (DSA). The objective of this paper is to demonstrate the applicability of the LBB design concept to main steam lines manufactured with SA106 Gr.C carbon steel. Based on the material properties, including fracture toughness and tensile properties obtained from the comprehensive material tests for base and weld metals, a parametric study was performed as described in this paper. The PICEP code was used to determine leak size crack (LSC) and the FLET code was used to perform the stability assessment of MS piping. The effects of material properties obtained from tests were evaluated to determine the LBB applicability for the MS piping. It can be shown from this parametric study that the MS piping has a high possibility of design using LBB analysis.

  11. 23. MPA-Seminar: Safety and reliability of plant technology with special emphasis on behaviour of pressurized components and systems at increased loading. Vol. 1. Papers 1-26

    International Nuclear Information System (INIS)

    1998-01-01

    This book is dedicated to the components of nuclear and conventional power plants with special emphasis on the behaviour of pressurized components and systems. The following topics are discussed: 1. integrity of pipes, vessels, and components and 2. fracture mechanics

  12. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  13. An overview of environmental degradation of materials in nuclear power plant piping systems

    International Nuclear Information System (INIS)

    Shack, W.J.

    1988-01-01

    Several types of environmental degradation of piping in light water reactor (LWR) power systems have already had significant economic impact on the industry. These include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping, erosion-corrosion of carbon steel piping in secondary systems, and a variety of types of fatigue failures. In addition, other problems have been identified that must be addressed in considering extended lifetimes for nuclear plants. These include the embrittlement of cast stainless steels after extended thermal aging at reactor operating temperatures and the effect of reactor environments on the design margin inherent in the ASME Section III fatigue design curves especially for carbon steel piping. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

  14. An overview of environmental degradation of materials in nuclear power plant piping systems

    International Nuclear Information System (INIS)

    Shack, W.J.

    1987-08-01

    Piping in light water reactor (LWR) power systems is affected by several types of environmental degradation: intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel piping in boiling water reactors (BWRs) has required research, inspection, and mitigation programs that will ultimately cost several billion dollars; erosion-corrosion of carbon steel piping has been observed frequently in the secondary systems of both BWRs and pressurized water reactors (PWRs); the effect of the BWR environment can greatly diminish the design margin inherent in the ASME Section III fatigue design curves for carbon steel piping; and cast stainless steels are subject to embrittlement after extended thermal aging at reactor operating temperatures. These problems are being addressed by wide-ranging research programs in this country and abroad. The purpose of this review is to highlight some of the accomplishments of these programs and to note some of the remaining unanswered questions

  15. Reactor Primary Coolant System Pipe Rupture Study. Progress report No. 32, July--December 1974

    International Nuclear Information System (INIS)

    1975-03-01

    The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue studies focused on Elastic/Plastic ASME Code Design Rules, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, and (c) studies directed at quantifying weld sensitization in T-304 stainless steel. (auth)

  16. Study on heat pipe assisted thermoelectric power generation system from exhaust gas

    Science.gov (United States)

    Chi, Ri-Guang; Park, Jong-Chan; Rhi, Seok-Ho; Lee, Kye-Bock

    2017-11-01

    Currently, most fuel consumed by vehicles is released to the environment as thermal energy through the exhaust pipe. Environmentally friendly vehicle technology needs new methods to increase the recycling efficiency of waste exhaust thermal energy. The present study investigated how to improve the maximum power output of a TEG (Thermoelectric generator) system assisted with a heat pipe. Conventionally, the driving energy efficiency of an internal combustion engine is approximately less than 35%. TEG with Seebeck elements is a new idea for recycling waste exhaust heat energy. The TEG system can efficiently utilize low temperature waste heat, such as industrial waste heat and solar energy. In addition, the heat pipe can transfer heat from the automobile's exhaust gas to a TEG. To improve the efficiency of the thermal power generation system with a heat pipe, effects of various parameters, such as inclination angle, charged amount of the heat pipe, condenser temperature, and size of the TEM (thermoelectric element), were investigated. Experimental studies, CFD simulation, and the theoretical approach to thermoelectric modules were carried out, and the TEG system with heat pipe (15-20% charged, 20°-30° inclined configuration) showed the best performance.

  17. A leak-before-break strategy for CANDU primary piping systems

    International Nuclear Information System (INIS)

    Aggarwal, M.L.; Kozluk, M.J.; Lin, T.C.; Manning, B.W.; Vijay, D.K.

    1986-01-01

    Recent advances in elastic-plastic fracture mechanics have made it possible to assess the stability of cracks in ductile piping systems. These technological developments have been used by Ontario Hydro as the nucleus of an approach for demonstrating that CANDU primary heat transport piping systems will not break catastrophically; at worst they would leak at a detectable rate. This leak-before-break approach has been taken on the Darlington nuclear generating station as a design stage alternative to the provision of pipe whip restraints on large diameter, primary heat transport system piping. Positive conclusions reached via this approach are considered sufficient to exclude the requirement to provide protective devices, such as pipe whip restraints. In arriving at the proposed leak-before-break approach a review of current and proposed leak-before-break licensing positions of other jurisdictions (particularly those in the United States and the Federal Republic of Germany) was carried out. The approach presented makes use of recent American developments in the area of elastic-plastic fracture mechanics. It also gives consideration to aspects which are unique to the pressurized heavy water (CANDU) reactors used by Ontario Hydro. The proposed leak-before-break approach is described and its use is illustrated by applying it to the Darlington generating station primary heat transport system pump suction piping. (author)

  18. Experimental Investigation of A Heat Pipe-Assisted Latent Heat Thermal Energy Storage System

    Science.gov (United States)

    Tiari, Saeed; Mahdavi, Mahboobe; Qiu, Songgang

    2016-11-01

    In the present work, different operation modes of a latent heat thermal energy storage system assisted by a heat pipe network were studied experimentally. Rubitherm RT55 enclosed by a vertical cylindrical container was used as the Phase Change Material (PCM). The embedded heat pipe network consisting of a primary heat pipe and an array of four secondary heat pipes were employed to transfer heat to the PCM. The primary heat pipe transports heat from the heat source to the heat sink. The secondary heat pipes transfer the extra heat from the heat source to PCM during charging process or retrieve thermal energy from PCM during discharging process. The effects of heat transfer fluid (HTF) flow rate and temperature on the thermal performance of the system were investigated for both charging and discharging processes. It was found that the HTF flow rate has a significant effect on the total charging time of the system. Increasing the HTF flow rate results in a remarkable increase in the system input thermal power. The results also showed that the discharging process is hardly affected by the HTF flow rate but HTF temperature plays an important role in both charging and discharging processes. The authors would like to acknowledge the financial supports by Temple University for the project.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  20. IPM Pipe

    Science.gov (United States)

    Submit A Report View Reports List [+] View Reports Map [+] CDM Alert System Sign Up For Alerts User Login Annual Epidemic Histories Annual Season Summaries Contact Us ipmPIPE User Login Web Administrator Login

  1. System identification by experimental data processing, application to turbulent transport of a tracer in pipe flow

    International Nuclear Information System (INIS)

    Burgos, Manuel; Getto, Daniel; Berne, Philippe

    2005-01-01

    System identification is the first, and probably the most important step in detecting abnormal behavior, control system design or performance improving. Data analysis is performed for studying the plant behavior, sensitivity of operation procedures and several other goals. In all these cases, the observed data is the convolution of an input function, and the system's impulse response. Practical discrete time convolutions may be performed multiplying a matrix built from the impulse response by the input vector, but for deconvolution it is necessary to invert the matrix which is singular in a causal system. Another method for deconvolution is by means of Fourier Transforms. Actual readings are usually corrupted by noise and, besides, their transform shows high low frequencies components and high frequency ones mainly due to additive noise. Subjective decisions as cut-off frequency should be taken as well. This paper proposes a deconvolution method based on parameters fitting of suitable models, where they exist, and estimation of values where analytical forms are not available. It is based on the global, non linear fitting of them, with a maximum likelihood criteria. An application of the method is shown using data from two fluid flow experiments. The experimental test rigs basically consist in a long section of straight pipe in which fluid is flowing. A pulse of tracer is injected at the entrance and detected at various locations along the pipe. An attempt of deconvolution of signals from successive probes using a classical model describing the flow of tracer as a plug moving with the average fluid velocity, plus some axial dispersion. The parameters are for instance the velocity of the plug and a dispersion coefficient. After parameter fitting, the model is found to reproduce the experimental data well. The flow rates deduced from the adjusted travel times are in very good agreement with the actual values. In addition, the flow dispersion coefficient is obtained

  2. Safety design guide for pipe rupture protection for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for pipe rupture protection identifies high-energy systems in which pipe ruptures must be postulated to occur, as well as systems that must be protected from the dynamic effects of such ruptures. Dynamic effects considered in this SDG consist of pipe whip (including missiles generated by pipe ruptures, if any) and jet impingement, Requirements for protection against the dynamic effects of a postulated pipe rupture and method of protection of essential structures, systems and components are specified for these effects. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 5 refs. (Author) .new

  3. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    International Nuclear Information System (INIS)

    Rahardjo, H.P.

    2011-01-01

    Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipment. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogeneously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occurred in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code) was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs) of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA) in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, expansion load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (re-routing) is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node 30, also a

  4. Examination of the X-ray piping diagnostic system using EGS4 (measuring the thickness of a steel pipe with rust)

    International Nuclear Information System (INIS)

    Kajiwara, G.

    2001-01-01

    In a series of papers entitled 'Examination of the X-ray piping diagnostic system using EGS4' presented the proceedings of the EGS4 users' meetings, I discussed the possibility of measuring the thickness of piping walls with rust. In the present paper, I describe, based on our earlier results, how the thickness of steel pipes with rust can be measured. I conducted EGS4 simulation to measure the thickness of a combination of steel and rust and made an energy absorption diagram for this combination. The equivalent thickness of steel was obtained through experiments and the system operation. The thickness of the steel determined by using the diagram agreed well with the actual steel thickness obtained by the experiments. In the future, we will focus on how to automate this measurement procedure and how to use the same procedure to measure the thickness of pipes filled with water. (author)

  5. Fundamentals of piping design

    CERN Document Server

    Smith, Peter

    2013-01-01

    Written for the piping engineer and designer in the field, this two-part series helps to fill a void in piping literature,since the Rip Weaver books of the '90s were taken out of print at the advent of the Computer Aid Design(CAD) era. Technology may have changed, however the fundamentals of piping rules still apply in the digitalrepresentation of process piping systems. The Fundamentals of Piping Design is an introduction to the designof piping systems, various processes and the layout of pipe work connecting the major items of equipment forthe new hire, the engineering student and the vetera

  6. Enhanced Thermal Management System for Spent Nuclear Fuel Dry Storage Canister with Hybrid Heat Pipes

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Dry storage uses the gas or air as coolant within sealed canister with neutron shielding materials. Dry storage system for spent fuel is regarded as relatively safe and emits little radioactive waste for the storage, but it showed that the storage capacity and overall safety of dry cask needs to be enhanced for the dry storage cask for LWR in Korea. For safety enhancement of dry cask, previous studies of our group firstly suggested the passive cooling system with heat pipes for LWR spent fuel dry storage metal cask. As an extension, enhanced thermal management systems for the spent fuel dry storage cask for LWR was suggested with hybrid heat pipe concept, and their performances were analyzed in thermal-hydraulic viewpoint in this paper. In this paper, hybrid heat pipe concept for dry storage cask is suggested for thermal management to enhance safety margin. Although current design of dry cask satisfies the design criteria, it cannot be assured to have long term storage period and designed lifetime. Introducing hybrid heat pipe concept to dry storage cask designed without disrupting structural integrity, it can enhance the overall safety characteristics with adequate thermal management to reduce overall temperature as well as criticality control. To evaluate thermal performance of hybrid heat pipe according to its design, CFD simulation was conducted and previous and revised design of hybrid heat pipe was compared in terms of temperature inside canister

  7. Theory and application of a three-dimensional code SHAPS to complex piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1983-01-01

    This paper describes the theory and application of a three-dimensional computer code SHAPS to the complex piping systems. The code utilizes a two-dimensional implicit Eulerian method for the hydrodynamic analysis together with a three-dimensional elastic-plastic finite-element program for the structural calculation. A three-dimensional pipe element with eight degrees of freedom is employed to account for the hoop, flexural, axial, and the torsional mode of the piping system. In the SHAPS analysis the hydrodynamic equations are modified to include the global piping motion. Coupling between fluid and structure is achieved by enforcing the free-slip boundary conditions. Also, the response of the piping network generated by the seismic excitation can be included. A thermal transient capability is also provided in SHAPS. To illustrate the methodology, many sample problems dealing with the hydrodynamic, structural, and thermal analyses of reactor-piping systems are given. Validation of the SHAPS code with experimental data is also presented

  8. Optimal Pipe Size Design for Looped Irrigation Water Supply System Using Harmony Search: Saemangeum Project Area

    Science.gov (United States)

    Lee, Ho Min; Sadollah, Ali

    2015-01-01

    Water supply systems are mainly classified into branched and looped network systems. The main difference between these two systems is that, in a branched network system, the flow within each pipe is a known value, whereas in a looped network system, the flow in each pipe is considered an unknown value. Therefore, an analysis of a looped network system is a more complex task. This study aims to develop a technique for estimating the optimal pipe diameter for a looped agricultural irrigation water supply system using a harmony search algorithm, which is an optimization technique. This study mainly serves two purposes. The first is to develop an algorithm and a program for estimating a cost-effective pipe diameter for agricultural irrigation water supply systems using optimization techniques. The second is to validate the developed program by applying the proposed optimized cost-effective pipe diameter to an actual study region (Saemangeum project area, zone 6). The results suggest that the optimal design program, which applies an optimization theory and enhances user convenience, can be effectively applied for the real systems of a looped agricultural irrigation water supply. PMID:25874252

  9. Optimal Pipe Size Design for Looped Irrigation Water Supply System Using Harmony Search: Saemangeum Project Area

    Directory of Open Access Journals (Sweden)

    Do Guen Yoo

    2015-01-01

    Full Text Available Water supply systems are mainly classified into branched and looped network systems. The main difference between these two systems is that, in a branched network system, the flow within each pipe is a known value, whereas in a looped network system, the flow in each pipe is considered an unknown value. Therefore, an analysis of a looped network system is a more complex task. This study aims to develop a technique for estimating the optimal pipe diameter for a looped agricultural irrigation water supply system using a harmony search algorithm, which is an optimization technique. This study mainly serves two purposes. The first is to develop an algorithm and a program for estimating a cost-effective pipe diameter for agricultural irrigation water supply systems using optimization techniques. The second is to validate the developed program by applying the proposed optimized cost-effective pipe diameter to an actual study region (Saemangeum project area, zone 6. The results suggest that the optimal design program, which applies an optimization theory and enhances user convenience, can be effectively applied for the real systems of a looped agricultural irrigation water supply.

  10. ELIMINATING CONSERVATISM IN THE PIPING SYSTEM ANALYSIS PROCESS THROUGH APPLICATION OF A SUITE OF LOCALLY APPROPRIATE SEISMIC INPUT MOTIONS

    International Nuclear Information System (INIS)

    Crawford, Anthony L.; Spears, Robert E.; Russell, Mark J.

    2009-01-01

    Seismic analysis is of great importance in the evaluation of nuclear systems due to the heavy influence such loading has on their designs. Current Department of Energy seismic analysis techniques for a nuclear safety-related piping system typically involve application of a single conservative seismic input applied to the entire system (1). A significant portion of this conservatism comes from the need to address the overlapping uncertainties in the seismic input and in the building response that transmits that input motion to the piping system. The approach presented in this paper addresses these two sources of uncertainty through the application of a suite of 32 input motions whose collective performance addresses the total uncertainty while each individual motion represents a single variation of it. It represents an extension of the soil-structure interaction analysis methodology of SEI/ASCE 43-05 (2) from the structure to individual piping components. Because this approach is computationally intensive, automation and other measures have been developed to make such an analysis efficient. These measures are detailed in this paper

  11. Investigations on penetration control for automated pipe welding system

    International Nuclear Information System (INIS)

    Fujiki, Daisuke; Sato, Akihiro; Funamoto, Takao; Matsumoto, Toshimi; Kobayashi, Masahiro

    1995-01-01

    We have been investigating process conditions forming sound root bead by orbital welding technique for nuclear power stations. Specimens used were stainless steel (SUS304) pipes (318.5 mm outside diameter and 15.4 mm thickness), and pulsed gas tungsten-arc (GTA) welder was adopted. We have found process conditions to form sound root bead by changing both heat input conditions and joint designs. It is found that reducing volume of molten metal is necessary to form sound root bead. And it is also found that changing joint designs is effective to reduce volume of molten metal. By selecting proper joint designs, we could form sound root bead in constant heat input conditions in every position of pipe. (author)

  12. Inelastic response of piping systems subjected to in-structure seismic excitation

    International Nuclear Information System (INIS)

    Campbell, R.D.; Kennedy, R.P.; Trasher, R.D.

    1983-01-01

    A study was undertaken to examine the inelastic response of single-degree-of-freedom systems and a simple piping system to varying levels of earthquake loading with superimposed static loading. The objective was to examine the conservatism inherent in ASME code rules for the design of piping systems by quantifying the ratio of the dynamic margin to the static margin for various degrees of inelastic strain, system frequencies and instructure time histories. Previous studies of elastic, perfectly-plastic and bilinear strain-hardening, single-degree-of-freedom models subjected to earthquake ground motion records have demonstrated the conservatism in current design methodology and design codes for earthquake resistant design of structures. This study compares response of single degree of freedom and simple piping system subjected to typical in-structure earthquake time histories and focuses on the excess margin inherent in current design criteria for piping systems. It is shown that the factor of safety against failure is variable and is dependent upon the frequency content of the loading, the dynamic characteristics of the piping system and the allowable system ductility. A recommendation is made for revision to current criteria on the basis of maintaining a constant factor of safety for dynamic and static loading

  13. Preliminary inspection of secondary cooling system piping for maintenance plan in JMTR

    International Nuclear Information System (INIS)

    Hanakawa, Hiroki; Hanawa, Yoshio; Izumo, Hironobu; Fukasaku, Akitomi; Nagao, Yoshiharu; Miyazawa, Masataka; Niimi, Motoji

    2008-01-01

    The JMTR is under the refurbishment and will start on FY 2011. The JMTR will operate for about 20 years from 2011. Before this JMTR operation, preliminary inspection of secondary cooling system piping was carried out in order to make a maintenance plan. As the results of this inspection, it was confirmed that the corrosion was reached by piping ingot, or decrease of piping thickness could hardly be observed. Therefore, it was confirmed that the strength or the functionality of the piping had been maintained by usual operation and maintenance. According to the results of this inspection, the basic date for maintenances are confirmed and it is clear to be able to make the maintenances plan in future. (author)

  14. Comparison of secondary system piping Cr content with inspection data

    International Nuclear Information System (INIS)

    Tapping, R.L.; Mitchell, A.M.

    1997-06-01

    For several years a number of Ontario Hydro and CANDU-6 stations have been sampling sections of secondary-side piping for chromium content. Several hundred of these measurements have been made, and comparisons with inspection data drawn. There is special interest in chromium concentrations in the range 0.01< Cr<0.1 wt.%, in order to better define the effect of trace chromium content on susceptibility to flow-assisted corrosion. (author)

  15. Seismic qualification of piping systems based on strain criteria

    International Nuclear Information System (INIS)

    Peters, K.; Rangette, A.

    1988-01-01

    Typical LMFBR piping is characterized by elevated temperature and low pressure levels. Taking into account operational conditions only these characteristics demand for and allow flexible piping design. The overestimation of the damage potential of seismic loading by e.g. improper failure criteria usually contradicts operational needs producing the known result of excessive ''snubberism'' and reduction of operational margins. As a matter of fact, due to its transiency seismic loading is essentially secondary provoking the natural design requirement ductility instead of stiffness and rigidity - i.e. exclusion of failure by strain control instead of stress control - and thus avoiding the LMFBR typical competition between operational needs and seismic qualification. The design requirement ductility needs judgement mechanisms, i.e. suitable load descriptions, allowed strain levels and strain evaluation tools. A simplified method for strain range estimation and the underlying basic ideas are roughly outlined. The status of verification and experience gained so far is described. The results achieved suggest that the qualification of piping based on ductility requirement controlled by strain criteria is not out of reach. (author)

  16. Survey of strong motion earthquake effects on thermal power plants in California with emphasis on piping systems. Volume 2, Appendices

    International Nuclear Information System (INIS)

    Stevenson, J.D.

    1995-11-01

    Volume 2 of the ''Survey of Strong Motion Earthquake Effects on Thermal Power Plants in California with Emphasis on Piping Systems'' contains Appendices which detail the detail design and seismic response of several power plants subjected to strong motion earthquakes. The particular plants considered include the Ormond Beach, Long Beach and Seal Beach, Burbank, El Centro, Glendale, Humboldt Bay, Kem Valley, Pasadena and Valley power plants. Included is a typical power plant piping specification and photographs of typical power plant piping specification and photographs of typical piping and support installations for the plants surveyed. Detailed piping support spacing data are also included

  17. Single-earthquake design for piping systems in advanced light water reactors

    International Nuclear Information System (INIS)

    Terao, D.

    1993-01-01

    Appendix A to Part 100 of Title 10 of the Code of Federal Regulations (10 CFR Part 100) requires, in part, that all structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public shall be designed to remain functional and within applicable stress and deformation limits when subject to an operating basis earthquake (OBE). The US Nuclear Regulatory Commission (NRC) is proposing changes to Appendix A to Part 100 to redefine the OBE at a level such that its purpose can be satisfied without the need to perform explicit response analyses. Consequently, only the safe-shutdown earthquake (SSE) would be required for the seismic design of safety-related structures, systems and components. The purpose of this paper is to discuss the proposed changes to existing seismic design criteria that the NRC staff has found acceptable for implementing the proposed rule change in the design of safety-related piping systems in the advanced light water reactor (ALWR) lead plant. These criteria apply only to the ALWR lead plant design and are not intended to replace the seismic design criteria approved by the Commission in the licensing bases of currently operating facilities. Although the guidelines described herein have been proposed for use as a pilot program for implementing the proposed rule change specifically for the ALWR lead plant, the NRC staff expects that these guidelines will also be applied to other ALWRs

  18. The evaluation of stress and piping support loads on RSG-GAS secondary cooling system

    International Nuclear Information System (INIS)

    Pustandyo, W.; Sitandung, Y. B.; Sujalmo, S.

    1998-01-01

    The evaluation of stress and piping support loads was evaluated on piping segment of secondary cooling water piping. In this paper, the analysis methods are presented with the use of computer code PS + CAEPIPE Version 3. 4. 05. W. From the selected pipe segment, the data of pipe characteristic, material properties, operation and design condition, equipment and support were used as inputs. The result of analysis show that stress and support loads if using location, kind and number of support equal with the system that have been installed for sustain load 3638 psi (node 160), thermal 13517 psi (node 90) and combination of sustain and thermal (node 90) 16747 psi. Meanwhile,if the optimization support, stress and support load for sustain load are respectively 4238 psi (node 10), thermal 13517 psi (node 90) and combination of sustain + thermal (node 90) 17350 psi. The limit values of permitted support based on Code PS+CAEPIPE of sustain load are 15000 psi, thermal 22500 psi and combination of sustain + thermal 37500 psi. The conclusion of evaluation result, that stress support load of pipe secondary cooling system are sufficiently low and using support show excessive and not economic

  19. Seismic re-evaluation of piping systems of heavy water plant, Kota

    International Nuclear Information System (INIS)

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

    2002-05-01

    Heavy Water Plant, Kota is the first indigenous heavy water plant built in India. The plant started operation in the year 1985 and it is approaching the completion of its originally stipulated design life. In view of the excellent record of plant operation for the past so many years, it has been planned to carry out various exercises for the life extension of the plant. In the first stage, evaluation of operation stresses was carried out for the process critical piping layouts and equipment, which are connected with 25 process critical nozzle locations, identified based on past history of the plant performance. Fatigue life evaluation has been carried out to fmd out the Cumulative Usage Factor, which helps in arriving at a decision regarding the life extension of the plant. The results of these exercises have been already reported separately vide BARC/200I /E/O04. In the second stage, seismic reevaluation of the plant has been carried out to assess its ability to maintain its integ:rity in case of a seismic event. The aim of this exercise is to assess the effects of the maximum probable earthquake at the plant site on the various systems and components of the plant. This exercise is further aimed at ensuring the adequacy of seismic supports to maintain the integrity of the system in case of a seismic event and to suggest some retrofitting measures, if required. Seismic re-evaluation of the piping of Heavy Water Plant, Kota has been performed taking into account the interaction effects from the connected equipment. Each layout has been qualified using the latest provisions of ASME Code Section III, Subsection ND wherein the earthquake loading has been considered as a reversing dynamic load. The maximum combined stresses for all the layouts due to pressure, weight and seismic loadings have been found to be well within the code allowable limit. Therefore, it has been concluded that during a maximum probable seismic event, the possibility of pipe rupture can be safely

  20. Experimental investigation on an integrated thermal management system with heat pipe heat exchanger for electric vehicle

    International Nuclear Information System (INIS)

    Zou, Huiming; Wang, Wei; Zhang, Guiying; Qin, Fei; Tian, Changqing; Yan, Yuying

    2016-01-01

    Highlights: • An integrated thermal management system is proposed for electric vehicle. • The parallel branch of battery chiller can supply additional cooling capacity. • Heat pipe performance on preheating mode is better than that on cooling mode. • Heat pipe heat exchanger is a feasible choice for battery thermal management. - Abstract: An integrated thermal management system combining a heat pipe battery cooling/preheating system with the heat pump air conditioning system is presented to fulfill the comprehensive energy utilization for electric vehicles. A test bench with battery heat pipe heat exchanger and heat pump air conditioning for a regular five-chair electric car is set up to research the performance of this integrated system under different working conditions. The investigation results show that as the system is designed to meet the basic cabinet cooling demand, the additional parallel branch of battery chiller is a good way to solve the battery group cooling problem, which can supply about 20% additional cooling capacity without input power increase. Its coefficient of performance for cabinet heating is around 1.34 at −20 °C out-car temperature and 20 °C in-car temperature. The specific heat of the battery group is tested about 1.24 kJ/kg °C. There exists a necessary temperature condition for the heat pipe heat exchanger to start action. The heat pipe heat transfer performance is around 0.87 W/°C on cooling mode and 1.11 W/°C on preheating mode. The gravity role makes the heat transfer performance of the heat pipe on preheating mode better than that on cooling mode.

  1. FRAMEWORK FOR STRUCTURAL ONLINE HEALTH MONITORING OF AGING AND DEGRADATION OF SECONDARY PIPING SYSTEMS DUE TO SOME ASPECTS OF EROSION

    Energy Technology Data Exchange (ETDEWEB)

    Gribok, Andrei V.; Agarwal, Vivek

    2017-06-01

    This paper describes the current state of research related to critical aspects of erosion and selected aspects of degradation of secondary components in nuclear power plants (NPPs). The paper also proposes a framework for online health monitoring of aging and degradation of secondary components. The framework consists of an integrated multi-sensor modality system, which can be used to monitor different piping configurations under different degradation conditions. The report analyses the currently known degradation mechanisms and available predictive models. Based on this analysis, the structural health monitoring framework is proposed. The Light Water Reactor Sustainability Program began to evaluate technologies that could be used to perform online monitoring of piping and other secondary system structural components in commercial NPPs. These online monitoring systems have the potential to identify when a more detailed inspection is needed using real time measurements, rather than at a pre-determined inspection interval. This transition to condition-based, risk-informed automated maintenance will contribute to a significant reduction of operations and maintenance costs that account for the majority of nuclear power generation costs. Furthermore, of the operations and maintenance costs in U.S. plants, approximately 80% are labor costs. To address the issue of rising operating costs and economic viability, in 2017, companies that operate the national nuclear energy fleet started the Delivering the Nuclear Promise Initiative, which is a 3 year program aimed at maintaining operational focus, increasing value, and improving efficiency. There is unanimous agreement between industry experts and academic researchers that identifying and prioritizing inspection locations in secondary piping systems (for example, in raw water piping or diesel piping) would eliminate many excessive in-service inspections. The proposed structural health monitoring framework takes aim at

  2. The Challenge of Providing Safe Water with an Intermittently Supplied Piped Water Distribution System

    Science.gov (United States)

    Kumpel, E.; Nelson, K. L.

    2012-12-01

    An increasing number of urban residents in low- and middle-income countries have access to piped water; however, this water is often not available continuously. 84% of reporting utilities in low-income countries provide piped water for fewer than 24 hours per day (van den Berg and Danilenko, 2010), while no major city in India has continuous piped water supply. Intermittent water supply leaves pipes vulnerable to contamination and forces households to store water or rely on alternative unsafe sources, posing a health threat to consumers. In these systems, pipes are empty for long periods of time and experience low or negative pressure even when water is being supplied, leaving them susceptible to intrusion from sewage, soil, or groundwater. Households with a non-continuous supply must collect and store water, presenting more opportunities for recontamination. Upgrading to a continuous water supply, while an obvious solution to these challenges, is currently out of reach for many resource-constrained utilities. Despite its widespread prevalence, there are few data on the mechanisms causing contamination in an intermittent supply and the frequency with which it occurs. Understanding the impact of intermittent operation on water quality can lead to strategies to improve access to safe piped water for the millions of people currently served by these systems. We collected over 100 hours of continuous measurements of pressure and physico-chemical water quality indicators and tested over 1,000 grab samples for indicator bacteria over 14 months throughout the distribution system in Hubli-Dharwad, India. This data set is used to explore and explain the mechanisms influencing water quality when piped water is provided for a few hours every 3-5 days. These data indicate that contamination occurs along the distribution system as water travels from the treatment plant to reservoirs and through intermittently supplied pipes to household storage containers, while real

  3. Development of a software for the ASME code qualification of class-I nuclear piping systems

    International Nuclear Information System (INIS)

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

    1999-11-01

    In nuclear industry, the designer often comes across the requirements of Class-1 piping systems which need to be qualified for various normal and abnormal loading conditions. In order to have quick design changes and the design reviews at various stages of design, it is quite helpful if a dedicated software is available for the qualification of Class-1 piping systems. BARC has already purchased a piping analysis software CAESAR-II and has used it for the life extension of heavy water plant, Kota. CAESAR-II facilitates the qualification of Class-2 and Class-3 piping systems among others. However, the present version of CAESAR-II does not have the capability to perform stress checks for the ASME Class-1 nuclear piping systems. With this requirement in mind and the prohibitive costs of commercially available software for the Class-1 piping analyses, it was decided to develop a separate software for this class of piping in such a way that the input and output details of the piping from the CAESAR-II software can be made use of. This report principally contains the details regarding development of a software for codal qualification of Class-1 nuclear piping as per ASME code section-III, NB-3600. The entire work was carried out in three phases. The first phase consisted of development of the routines for reading the output files obtained from the CAESAR-II software, and converting them into required format for further processing. In this phase, the nodewise informations available from the CAESAR-II output file were converted into element-wise informations. The second phase was to develop a general subroutine for reading the various input parameters such as diameter, wall thickness, corrosion allowance, bend radius and also to recognize the bend elements based on the bend radius, directly from the input file of CAESAR-II software. The third phase was regarding the incorporation of the required steps for performing the ASME codal checks as per NB-3600 for Class-1 piping

  4. The conservatism of the net-section stress procedure for predicting the failure of cracked piping systems: the effect of crack shape complexity on the degree of conservatism

    International Nuclear Information System (INIS)

    Smith, E.

    1996-01-01

    The failure of circumferentially cracked steel piping is often predicted by assuming that failure conforms to a net-section stress criterion using as input an appropriate value for the critical net-section stress together with a knowledge of the anticipated loadings. The stress at the cracked section is usually calculated via a purely elastic analysis based on the piping being uncracked. however, because the piping is built-in at the ends into a larger component, and since the onset of crack extension requires some plastic deformation, use of the net-section stress approach can give overly conservative failure predictions. In earlier work, the author has quantified the extent of this conservatism, and has shown how it depends on the geometry of the cracked section, the material ductility and the elastic flexibility of a piping system. This paper quantifies the conservatism with regard to the case where a through-wall crack extends over a prescribed fraction of the pipe circumference, while there is also an internal circumferential crack extending around the remainder of the pipe section. This is an extreme form of circumferential cracking but nevertheless, simulates the well-known Duane Arnold safe-end crack. (Author)

  5. Bayesian Belief Networks for predicting drinking water distribution system pipe breaks

    International Nuclear Information System (INIS)

    Francis, Royce A.; Guikema, Seth D.; Henneman, Lucas

    2014-01-01

    In this paper, we use Bayesian Belief Networks (BBNs) to construct a knowledge model for pipe breaks in a water zone. To the authors’ knowledge, this is the first attempt to model drinking water distribution system pipe breaks using BBNs. Development of expert systems such as BBNs for analyzing drinking water distribution system data is not only important for pipe break prediction, but is also a first step in preventing water loss and water quality deterioration through the application of machine learning techniques to facilitate data-based distribution system monitoring and asset management. Due to the difficulties in collecting, preparing, and managing drinking water distribution system data, most pipe break models can be classified as “statistical–physical” or “hypothesis-generating.” We develop the BBN with the hope of contributing to the “hypothesis-generating” class of models, while demonstrating the possibility that BBNs might also be used as “statistical–physical” models. Our model is learned from pipe breaks and covariate data from a mid-Atlantic United States (U.S.) drinking water distribution system network. BBN models are learned using a constraint-based method, a score-based method, and a hybrid method. Model evaluation is based on log-likelihood scoring. Sensitivity analysis using mutual information criterion is also reported. While our results indicate general agreement with prior results reported in pipe break modeling studies, they also suggest that it may be difficult to select among model alternatives. This model uncertainty may mean that more research is needed for understanding whether additional pipe break risk factors beyond age, break history, pipe material, and pipe diameter might be important for asset management planning. - Highlights: • We show Bayesian Networks for predictive and diagnostic management of water distribution systems. • Our model may enable system operators and managers to prioritize system

  6. A fatigue analysis including environmental effects for a pipe system in a Swedish BWR

    International Nuclear Information System (INIS)

    Steingrimsdottir, Kristin; Dahlberg, Magnus

    2011-10-01

    A BWR feed water piping system (austenitic steel) has been analyzed with two different fatigue curves and environmental factors. Original fatigue curve from ASME is compared to a new fatigue curve; ANL. The influence of environmental correction factors (Fen) is studied further for the piping system. It is noted that the results apply for this particular system, and general conclusions should be cautiously drawn. Typical for this system is that all dominant loads are within the low-cycle regime. This implies that the change of fatigue curve only leads to limited increases in usage factors. Larger changes can occur if larger number of cycles is within the high-cycle regime

  7. Numerical and experimental analysis of heat pipes with application in concentrated solar power systems

    Science.gov (United States)

    Mahdavi, Mahboobe

    Thermal energy storage systems as an integral part of concentrated solar power plants improve the performance of the system by mitigating the mismatch between the energy supply and the energy demand. Using a phase change material (PCM) to store energy increases the energy density, hence, reduces the size and cost of the system. However, the performance is limited by the low thermal conductivity of the PCM, which decreases the heat transfer rate between the heat source and PCM, which therefore prolongs the melting, or solidification process, and results in overheating the interface wall. To address this issue, heat pipes are embedded in the PCM to enhance the heat transfer from the receiver to the PCM, and from the PCM to the heat sink during charging and discharging processes, respectively. In the current study, the thermal-fluid phenomenon inside a heat pipe was investigated. The heat pipe network is specifically configured to be implemented in a thermal energy storage unit for a concentrated solar power system. The configuration allows for simultaneous power generation and energy storage for later use. The network is composed of a main heat pipe and an array of secondary heat pipes. The primary heat pipe has a disk-shaped evaporator and a disk-shaped condenser, which are connected via an adiabatic section. The secondary heat pipes are attached to the condenser of the primary heat pipe and they are surrounded by PCM. The other side of the condenser is connected to a heat engine and serves as its heat acceptor. The applied thermal energy to the disk-shaped evaporator changes the phase of working fluid in the wick structure from liquid to vapor. The vapor pressure drives it through the adiabatic section to the condenser where the vapor condenses and releases its heat to a heat engine. It should be noted that the condensed working fluid is returned to the evaporator by the capillary forces of the wick. The extra heat is then delivered to the phase change material

  8. Study (Prediction of Main Pipes Break Rates in Water Distribution Systems Using Intelligent and Regression Methods

    Directory of Open Access Journals (Sweden)

    Massoud Tabesh

    2011-07-01

    Full Text Available Optimum operation of water distribution networks is one of the priorities of sustainable development of water resources, considering the issues of increasing efficiency and decreasing the water losses. One of the key subjects in optimum operational management of water distribution systems is preparing rehabilitation and replacement schemes, prediction of pipes break rate and evaluation of their reliability. Several approaches have been presented in recent years regarding prediction of pipe failure rates which each one requires especial data sets. Deterministic models based on age and deterministic multi variables and stochastic group modeling are examples of the solutions which relate pipe break rates to parameters like age, material and diameters. In this paper besides the mentioned parameters, more factors such as pipe depth and hydraulic pressures are considered as well. Then using multi variable regression method, intelligent approaches (Artificial neural network and neuro fuzzy models and Evolutionary polynomial Regression method (EPR pipe burst rate are predicted. To evaluate the results of different approaches, a case study is carried out in a part ofMashhadwater distribution network. The results show the capability and advantages of ANN and EPR methods to predict pipe break rates, in comparison with neuro fuzzy and multi-variable regression methods.

  9. Development of bore tools for pipe inspection

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Nakahira, Masataka; Taguchi, Kou; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Reactor (ITER), replacement and maintenance on in-vessel components requires that all cooling pipes connected be cut and removed, that a new component be installed, and that all cooling pipes be rewelded. After welding is completed, welded area must be inspected for soundness. These tasks require a new work concept for securing shielded area and access from narrow ports. Tools had to be developed for nondestructive inspection and leak testing to evaluate pipe welding soundness by accessing areas from inside pipes using autonomous locomotion welding and cutting tools. A system was proposed for nondestructive inspection of branch pipes and the main pipe after passing through pipe curves, the same as for welding and cutting tool development. Nondestructive inspection and leak testing sensors were developed and the basic parameters were obtained. In addition, the inspection systems which can move inside pipes and conduct the nondestructive inspection and the leak testing were developed. In this paper, an introduction will be given to the current situation concerning the development of nondestructive inspection and leak testing machines for the branch pipes. (author)

  10. New developments of belt conveyor systems; Inclined belt systems, vertical pipe elevators, vibration belts, oscillating tubes

    Energy Technology Data Exchange (ETDEWEB)

    Bahke, E.A. (Universitaet Karlsruhe, Karlsruhe (Germany, F.R.). Inst. fuer Foerdertechnik)

    1991-03-01

    Factors that have influenced the design of belt conveyor systems are discussed - these include strength and shaping. Belt conveyor systems for inclined, steep-angle and vertical conveying are described and comparison made between cable belt and steel cord belt conveyors used in coal mines. Hose-belt or tube conveyors such as are used in the PWH/Conti-Rollgurt Conveyor System for feeding boilers in German coal fired power stations are mentioned and advantages of the pipe-belt conveyor for vertical transport discussed. Design of the vibratory conveyor for transporting solids upwards by pulses is described. 29 refs., 19 figs., 2 tabs.

  11. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  12. A state-of-the-art review on hybrid heat pipe latent heat storage systems

    International Nuclear Information System (INIS)

    Naghavi, M.S.; Ong, K.S.; Mehrali, M.; Badruddin, I.A.; Metselaar, H.S.C.

    2015-01-01

    The main advantage of latent heat thermal energy storage systems is the capability to store a large quantity of thermal energy in an isothermal process by changing phase from solid to liquid, while the most important weakness of these systems is low thermal conductivity that leads to unsuitable charging/discharging rates. Heat pipes are used in many applications – as one of the most efficient heat exchanger devices – to amplify the charging/discharging processes rate and are used to transfer heat from a source to the storage or from the storage to a sink. This review presents and critically discusses previous investigations and analysis on the incorporation of heat pipe devices into latent heat thermal energy storage with heat pipe devices. This paper categorizes different applications and configurations such as low/high temperature solar, heat exchanger and cooling systems, analytical approaches and effective parameters on the performance of hybrid HP–LHTES systems.

  13. Design and implementation of cooling system for beam pipe of BESIII

    International Nuclear Information System (INIS)

    Li Xunfeng; Zheng Lifang; Dong Sujun

    2008-01-01

    Cooling system for beam pipe is designed, based on the properties of structure, the surrounding and the required temperature of beam pipe in BESIII. The main devices are double for spare parts, and Siemens program logic control is used in the cooling system, which realize the reliability of the equipment and assure the system long time running. OPC is used to communicate between Upper computer and program logic control as the third-party communication protocol, which resolve the problem of communication for complex multi-station, the upper computer assist the program logic control to detect and control the equipment. The cooling system have reasonable structure, comprehensive function, good precision; it can take away the heat from inner wall of beam pipe in time, and control the temperature on inner wall and outer wall in the required range. (authors)

  14. Implementation of the structural integrity analysis for PWR primary components and piping

    International Nuclear Information System (INIS)

    Pellissier-Tanon, A.

    1982-01-01

    The trends on the definition, the assessment and the application of fracture strength evaluation methodology, which have arisen through experience in the design, construction and operation of French 900-MW plants are reviewed. The main features of the methodology proposed in a draft of Appendix ZG of the RCC-M code of practice for the design verification of fracture strength of primary components are presented. The research programs are surveyed and discussed from four viewpoints, first implementation of the LEFM analysis, secondly implementation of the fatigue crack propagation analysis, thirdly analysis of vessel integrity during emergency core cooling, and fourthly methodology for tear fracture analysis. (author)

  15. An experimental study of the response of the multiple support piping systems

    International Nuclear Information System (INIS)

    Chiba, T.; Koyanagi, R.

    1987-01-01

    From the test results, following remarks have been obtained. 1. Since the effect of internal pressure was not so small on the stress response, its effect should be considered in the design of piping systems. 2. The effect of the phase of excitations was fairly dominant to the response of piping systems. From this fact, the adopting of the support structures which have different dynamic characteristics may be one of the more realistic approaches to reduce the response of piping systems. 3. The acceleration responses near the support points are always underestimated because the natural modes of the analysis are zero at these support points. 4. If the pseudo-static response is dominant, the stress responses near the support points are always overestimated by the ABS method to support groups. In such case the SRSS method is recommended. 5. The 10% method to the closely spaced modes is conservative for the flexible piping. The closely spaced mode methods to these flexible piping systems should be used carefully. 6. The SRSS combination method is offered the reasonable results to the space, modes and support groups in the multiple response spectra method. (orig.)

  16. On the optimization of support positioning and stiffness for piping systems in power plants

    International Nuclear Information System (INIS)

    Collina, A.; Zanaboni, P.; Belloli, M.

    2005-01-01

    The optimal location of supports for the reduction of vibration in piping systems is an interesting structural problem, that can be approached also with the methods used in the updating parameters problems, especially when the cause of the vibration is due to a resonance or a ineffective damping level, and the shift of a critical frequency is an effective remedy. In the proposed paper a frequency domain method of Finite Element model updating is used in order to properly locate the natural frequencies of a given piping system, starting from the nominal condition. The method considers modal parameters, and enables to directly update the physical quantities of the finite element model, i.e., in the considered application, the stiffness and damping of the supporting devices. The model of the structure is made up by beam finite elements, lumped springs and masses, and rigid links. The general procedure is applied here updating the stiffness and location of the supports of the piping according to the comparison among the current frequencies of the piping systems structure and the ones that are required, according to specific requirements. Application to a system similar to a portion of an actual AP1000 (a new Advanced Passive Nuclear Reactor) piping-support layout is presented. (authors)

  17. Corrosion and fouling concerns of service water piping and heat exchanger components

    International Nuclear Information System (INIS)

    Witt, F.J.

    1991-01-01

    In this paper, the author would like to discuss NRC generic communications issued to all operating nuclear power plants for service water systems relative to corrosion (including microbiologically influenced corrosion (MIC) and typical operational events involving system failure and degradations as a result of corrosion and fouling. Lessons can be learned from the unforeseen experience of others, and knowing and understanding these experiences can improve the performance of safety-related service water systems

  18. High temperature heat pipe experiments in low earth orbit

    International Nuclear Information System (INIS)

    Woloshun, K.; Merrigan, M.A.; Sena, J.T.; Critchley, E.

    1993-01-01

    Although high temperature, liquid metal heat pipe radiators have become a standard component on most high power space power system designs, there is no experimental data on the operation of these heat pipes in a zero gravity or micro-gravity environment. Experiments to benchmark the transient and steady state performance of prototypical heat pipe space radiator elements are in preparation for testing in low earth orbit. It is anticipated that these heat pipes will be tested aborad the Space Shuttle in 1995. Three heat pipes will be tested in a cargo bay Get Away Special (GAS) canister. The heat pipes are SST/potassium, each with a different wick structure; homogeneous, arterial, and annular gap, the heat pipes have been designed, fabricated, and ground tested. In this paper, the heat pipe designs are specified, and transient and steady-state ground test data are presented

  19. Water hammer and cavitational hammer in process plant pipe systems

    International Nuclear Information System (INIS)

    Dudlik, A.; Schoenfeld, S.B.H.; Hagemann, O.; Fahlenkamp, H.

    2003-01-01

    Fast acting valves are often applied for quick safety shut-down of pipelines for liquids and gases in the chemical and petrochemical industry as well as in power plants and state water supplies. The fast deceleration of the liquid leads to water hammer upstream the valve and to cavitational hammer downstream the fast closing valve. The valve characteristics given by manufacturers are usually measured at steady state flow conditions of the liquid. In comparison, the dynamic characteristics depend on the initial liquid velocity, valve closing velocity, the absolute pipe pressure and the pipe geometry. Fraunhofer UMSICHT conducts various test series examining valve dynamic characteristics in order of the dynamic analysis of pressure surges in fast closing processes. Therefore a test rig is used which consists of two pipelines of DN 50 and DN 100 with an approximate length of 230 m each. In this paper the results of performed pressure surge experiments with fast closing and opening valves will be compared to calculations of commercial software programs such as MONA, FLOWMASTER 2. Thus the calculation software for water supply, power plants oil and gas and chemical industry can be permanently improved. (orig.)

  20. Application of new developments in coupled seismic analysis of piping systems

    International Nuclear Information System (INIS)

    Gupta, A.; Gupta, A.K.

    1995-01-01

    The current practice of calculating the seismic response is to perform the analysis of the primary structure (buildings) and the secondary systems (piping) separately. Earthquake input to the primary system in terms of a design response spectrum. An acceleration time history compatible with the design response spectrum is developed (a non-unique process) and primary system is analyzed to obtain the acceleration histories at the desired floors. Floor time histories are used for generating the corresponding instructure response spectrum (IRIS). The instructure response spectra are used as input at the supports of secondary systems. Further, in case of multiple supports, an envelope spectrum (introducing conservatism) is obtained from the individual support IRS. The effect of relative support motion is incorporated by a worst-case separate static analysis (adding to the conservatism). In the above method, mass interaction between the secondary and primary system is ignored, which may have significant effect at resonant frequencies (further adding to the conservatism). The calculated response may be an order of magnitude higher than they should be. Two computer programs, CREST and CREST-IRIS, were developed at Center for NUclear Power Plant Structures, Equipment and Piping. Any one of the two computer programs together with a piping analysis program can be used to perform an accurate coupled seismic analysis of piping systems. The two computer programs have been validated against the time history analysis for simple problems. In the present study, we have applied CREST to analyze two real-life piping systems. The piping analysis program used in this research is the commercial software PIPESTRESS, developed by DST Computer Services of Geneva, Switzerland. (author). 4 refs., 3 figs., 2 tabs

  1. Investigation of noninvasive healing of damaged piping system using electro-magneto-mechanical methods

    KAUST Repository

    Mukherjee, Debanjan

    2014-01-01

    Virtually all engineering applications involve the use of piping, conduits and channels. In the petroleum industry, piping systems are extensively employed in upstream and downstream processes. These piping systems often carry fluids that are corrosive, which leads to wear, cavitation and cracking. The replacement of damaged piping systems can be quite expensive, both in terms of capital costs, as well as in operational downtime. This motivates the present research on noninvasive healing of cracked piping systems. In this investigation, we propose to develop computational models for characterizing noninvasive repair strategies involving electromagnetically guided particles. The objective is to heal industrial-piping systems noninvasively, from the exterior of the system, during operation, resulting in no downtime, with minimal relative cost. The particle accumulation at a target location is controlled by external electro-magneto-mechanical means. There are two primary effects that play a role for guiding the particles to the solid-fluid interface/wall: mechanical shear due to the fluid flow, and an electrical or magnetic force. In this work we develop and study a relationship that characterizes contributions of both, and ascertain how this relationship scales with characteristic physical parameters. Characteristic non-dimensional parameters that describe system behavior are derived and their role in design is illustrated. A detailed, fully 3-dimensional discrete element simulation framework is presented, and illustrated using a model problem of magnetically guided particles. The detailed particle behavior is considered to be regulated by three effects: (1) the field strength (2) the mass flow rate and (3) the wall interactions.

  2. Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning

    International Nuclear Information System (INIS)

    Kastner, W.; Erve, M.; Henzel, N.; Stellwag, B.

    1990-01-01

    Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs

  3. High Temperatures Health Monitoring of the Condensed Water Height in Steam Pipe Systems

    Science.gov (United States)

    Lih, Shyh-Shiuh; Bar-Cohen, Yoseph; Lee, Hyeong Jae; Badescu, Mircea; Bao, Xiaoqi; Sherrit, Stewart; Takano, Nobuyuki; Ostlund, Patrick; Blosiu, Julian

    2013-01-01

    Ultrasonic probes were designed, fabricated and tested for high temperature health monitoring system. The goal of this work was to develop the health monitoring system that can determine the height level of the condensed water through the pipe wall at high temperature up to 250 deg while accounting for the effects of surface perturbation. Among different ultrasonic probe designs, 2.25 MHz probes with air backed configuration provide satisfactory results in terms of sensitivity, receiving reflections from the target through the pipe wall. A series of tests were performed using the air-backed probes under irregular conditions, such as surface perturbation and surface disturbance at elevated temperature, to qualify the developed ultrasonic system. The results demonstrate that the fabricated air-backed probes combined with advanced signal processing techniques offer the capability of health monitoring of steam pipe under various operating conditions.

  4. Leak-before-break due to fatigue cracks in the cold leg piping system

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Collier, R.P.

    1984-01-01

    This review paper presents the results of a deterministic assessment of the margin of safety against a large break in the cold leg piping system of pressurized water reactors. The paper focuses on the computation of leak rates resulting from fatigue cracks that penetrate the full wall thickness. Results are presented that illustrate the sensitivity of the leak rate to stress level, crack shape and crack orientation. Further, the leak rates for specific conditions are contrasted to detection levels, shutdown criteria, make-up capacity and the leak rate associated with final failure of the piping system. The results of these computations indicate that, in general, leaks far in excess of the present detection sensitivities would result at crack sizes well below the critical crack sizes for the upset loadings on the cold leg piping system

  5. A simplified dynamic analysis for reactor piping systems under blowdown conditions

    International Nuclear Information System (INIS)

    Chen, M.M.

    1975-01-01

    In the design of pipelines in a nuclear power plant for blowdown conditions, is it customary to conduct dynamic analysis of the piping system to obtain the responses and the resulting stresses. Calculations are repeated for each design modification in piping geometry or supporting system until the design codes are met. The numerical calculations are, in general, very costly and time consuming. Until now, there have been no simple means for calculating the dynamic responses for the design. The proposed method reduces the dynamic calculation to a quasi-static one, and can be beneficially used for the preliminary design. The method is followed by a complete dynamical analysis to improve the final results. The new formulations greatly simplify the numerical computation and provide design guides. When used to design a given piping system, the method saved approximately one order of magnitude of computer time. The approach can also be used for other types of structures

  6. A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, Robert A.

    2013-09-18

    Hydrogen explosions may occur simultaneously with water hammer accidents in nuclear facilities, and a theoretical mechanism to relate water hammer to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in pipe systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the pipe system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany. Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism.

  7. Gyrotron: an ECH system component

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Baldiri Salvado

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

  9. Experimental study of a photovoltaic solar-assisted heat-pump/heat-pipe system

    International Nuclear Information System (INIS)

    Fu, H.D.; Pei, G.; Ji, J.; Long, H.; Zhang, T.; Chow, T.T.

    2012-01-01

    A practical design for a heat pump with heat-pipe photovoltaic/thermal (PV/T) collectors is presented. The hybrid system is called the photovoltaic solar-assisted heat-pump/heat-pipe (PV-SAHP/HP) system. To focus on both actual demand and energy savings, the PV-SAHP/HP system was designed to be capable of operating in three different modes, namely, the heat-pipe, solar-assisted heat pump, and air-source heat-pump modes. Based on solar radiation, the system operates in an optimal mode. A series of experiments were conducted in Hong Kong to study the performance of the system when operating in the heat-pipe and the solar-assisted heat-pump modes. Moreover, energy and exergy analyses were used to investigate the total PV/T performance of the system. - Highlights: ► A novel PV-SAHP/HP system with three different operating modes was proposed. ► Performance of the PV-SAHP/HP system was studied experimentally. ► A optimal operating mode of the PV-SAHP/HP system was suggested in this paper.

  10. 46 CFR 108.437 - Pipe sizes and discharge rates for enclosed ventilation systems for rotating electrical equipment.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pipe sizes and discharge rates for enclosed ventilation systems for rotating electrical equipment. 108.437 Section 108.437 Shipping COAST GUARD, DEPARTMENT OF... Systems Fixed Carbon Dioxide Fire Extinguishing Systems § 108.437 Pipe sizes and discharge rates for...

  11. Evaluations of the piping system inelastic analysis computer program PIRAX2

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1977-01-01

    The report contains two sets of comparisons of inelastic test data with PIRAX2-Theory; i.e., ORNL beam tests and ORNL elbow tests. The purpose of these comparisons is to evaluate the accuracy of the simplified analytical techniques used in PIRAX2. The test data are on structures that are much simpler than piping systems but provide a fundamental basis for comparison. The report includes an analysis of a 3-anchor piping system to illustrate the relative simplicity of PIRAX2 input/output data and relatively small computer running time. Some areas of needed improvements in PIRAX2 are discussed

  12. 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

  13. Shuttle Ku-band bent-pipe implementation considerations. [for Space Shuttle digital communication systems

    Science.gov (United States)

    Batson, B. H.; Seyl, J. W.; Huth, G. K.

    1977-01-01

    This paper describes an approach for relay of data-modulated subcarriers from Shuttle payloads through the Shuttle Ku-band communications subsystem (and subsequently through a tracking and data relay satellite system to a ground terminal). The novelty is that a channel originally provided for baseband digital data is shown to be suitable for this purpose; the resulting transmission scheme is referred to as a narrowband bent-pipe scheme. Test results demonstrating the validity of the narrowband bent-pipe mode are presented, and limitations on system performance are described.

  14. Seismic response analysis of a piping system subjected to multiple support excitations in a base isolated NPP building

    International Nuclear Information System (INIS)

    Surh, Han-Bum; Ryu, Tae-Young; Park, Jin-Sung; Ahn, Eun-Woo; Choi, Chul-Sun; Koo, Ja Choon; Choi, Jae-Boong; Kim, Moon Ki

    2015-01-01

    Highlights: • Piping system in the APR 1400 NPP with a base isolation design is studied. • Seismic response of piping system in base isolated building are investigated. • Stress classification method is examined for piping subjected to seismic loading. • Primary stress of piping is reduced due to base isolation design. • Substantial secondary stress is observed in the main steam piping. - Abstract: In this study, the stress response of the piping system in the advanced power reactor 1400 (APR 1400) with a base isolation design subjected to seismic loading is addressed. The piping system located between the auxiliary building with base isolation and the turbine building with a fixed base is considered since it can be subjected to substantial relative support movement during seismic events. First, the support responses with respect to the base characteristic are investigated to perform seismic analysis for multiple support excitations. Finite element analyses are performed to predict the piping stress response through various analysis methods such as the response spectrum, seismic support movement and time history method. To separately evaluate the inertial effect and support movement effect on the piping stress, the stress is decomposed into a primary and secondary stress using the proposed method. Finally, influences of the base isolation design on the piping system in the APR 1400 are addressed. The primary stress based on the inertial loading is effectively reduced in a base isolation design, whereas a considerable amount of secondary stress is generated in the piping system connecting a base isolated building with a fixed base building. It is also confirmed that both the response spectrum analysis and seismic support movement analysis provide more conservative estimations of the piping stress compared to the time history analysis

  15. Ultrasonic Measurement of Erosion/corrosion Rates in Industrial Piping Systems

    Science.gov (United States)

    Sinclair, A. N.; Safavi, V.; Honarvar, F.

    2011-06-01

    Industrial piping systems that carry aggressive corrosion or erosion agents may suffer from a gradual wall thickness reduction that eventually threatens pipe integrity. Thinning rates could be estimated from the very small change in wall thickness values measured by conventional ultrasound over a time span of at least a few months. However, measurements performed over shorter time spans would yield no useful information—minor signal distortions originating from grain noise and ultrasonic equipment imperfections prevent a meaningful estimate of the minuscule reduction in echo travel time. Using a Model-Based Estimation (MBE) technique, a signal processing scheme has been developed that enables the echo signals from the pipe wall to be separated from the noise. This was implemented in a laboratory experimental program, featuring accelerated erosion/corrosion on the inner wall of a test pipe. The result was a reduction in the uncertainty in the wall thinning rate by a factor of four. This improvement enables a more rapid response by system operators to a change in plant conditions that could pose a pipe integrity problem. It also enables a rapid evaluation of the effectiveness of new corrosion inhibiting agents under plant operating conditions.

  16. Integrity assessment of the cold leg piping system in a PWR

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Leis, B.N.

    1981-01-01

    The purpose of this paper is to examine the integrity of a nuclear piping system, designed in accordance with Section III, in the context of a damage tolerance analysis procedure. Such a procedure directly addresses the defects and cyclic loadings that are responsible for the above noted exceptions. The analysis and results reported here are for a fatigue life analysis of the Cold Leg piping in a PWR. This piping system is particularly important from a safety standpoint since a large break is a possible initiator of a core meltdown accident. The analysis employs LEFM concepts to determine the time between the initial start-up and (1) formation of a leak, (2) detection of the leak, and (3) the final fracture of the piping. Both longitudinal and circumferential defects are considered. The defects are assumed to propagate from the pipe I.D. in a self-similar manner. Inputs to the analysis were derived from information supplied by plant operators and vendors, published data, and 'expert opinions'. The life was computed using a linear damage accumulation. (orig./GL)

  17. Experimental research of heat recuperators in ventilation systems on the basis of heat pipes

    Directory of Open Access Journals (Sweden)

    Matveev Andrey

    2017-01-01

    Full Text Available The paper presents the results of experimental studies of heat pipes and their thermo-technical characteristics (heat power, conductivity, heat transfer resistance, heat-transfer coefficient, temperature level and differential, etc.. The theoretical foundations and the experimental methods of the research of ammonia heat pipes made of aluminum section АS – КRА 7.5 – R1 (made of the alloy AD - 31 are explained. The paper includes the analysis of the thermo-technical characteristics of heat pipes as promising highly efficient heat transfer devices, which may be used as the basic elements of heat exchangers - heat recuperators for exhaust ventilation air, capable of providing energy-saving technologies in ventilation systems for housing and public utilities and for various branches of industry. The thermo-technical characteristics of heat pipes (HP as the basic elements of a decentralized supply-extract ventilation system (DSEVS and energy-saving technologies are analyzed. As shown in the test report of the ammonia horizontal HP made of the section АS-КRА 7,5-R1-120, this pipe ensures safe operation under various loads.

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

    International Nuclear Information System (INIS)

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

    2003-12-01

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

  19. Use of the modal superposition technique for piping system blowdown analyses

    International Nuclear Information System (INIS)

    Ware, A.G.; Macek, R.W.

    1983-01-01

    A standard method of solving for the seismic response of piping systems is the modal superposition technique. Only a limited number of structural modes are considered (typically those up to 33 Hz in the U.S.), since the effect on the calculated response due to higher modes is generally small, and the method can result in considerable computer cost savings over the direct integration method. The modal superposition technique has also been applied to piping response problems in which the forcing functions are due to fluid excitation. Application of the technique to this case is somewhat more difficult, because a well defined cutoff frequency for determining structural modes to be included has not been established. This paper outlines a method for higher mode corrections, and suggests methods to determine suitable cutoff frequencies for piping system blowdown analyses. A numerical example illustrates how uncorrected modal superposition results can produce erroneous stress results

  20. Determination of limits for smallest detectable and largest subcritical leakage cracks in piping systems

    International Nuclear Information System (INIS)

    Bieselt, R.; Wolf, M.

    1995-01-01

    Nuclear power plant piping systems - those still in their original as-built condition as well as upgraded designs - are subject to safety analysis. In order to limit the consequences of postulated piping failures, the basic safety concept incorporating rupture preclusion criteria is applied to specific high-energy piping systems. Leak-before-break analyses are also conducted within the framework of this concept. These analyses serve to determine the potential consequences of jet and reaction forces due to maximum subcritical leakage cracks while also establishing the minimum crack sizes that would be reliably detectable by the leakage rates resulting from these cracks. The boundary conditions for these analyses are not clearly defined. Using various examples as a basis, this paper presents and discusses how the leak-before-break concept can be applied. (orig.)

  1. Practical considerations from construction of the FFTF pipe support system

    International Nuclear Information System (INIS)

    Hartsell, J.O.

    1979-04-01

    Late this year the FFTF is scheduled to initiate criticality and power operation, an event signifying complete resolution of construction problems and compliance with applicable requirements. One of the tasks accomplished to achieve this position was the conduct of cell completion walkthroughs, problem identification, and resolution. Heavily oriented toward the final verification of pipe supports, the cell completion process identified and resolved a series of problems not readily apparent during the detailed design phases of plant construction. It is the objective of this presentation to establish a degree of awareness of these types of problems and of the FFTF techniques employed in resolution. All of the problems depicted herein have been solved and the installations corrected

  2. Water Hammer Mitigation on Postulated Pipe Break of Feed Water System

    International Nuclear Information System (INIS)

    Seong, Ho Je; Woo, Kab Koo; Cho, Keon Taek

    2008-01-01

    The Feed Water (FW) system supplies feedwater from the deaerator storage tank to the Steam Generators(S/G) at the required pressure, temperature, flow rate, and water chemistry. The part of FW system, from the S/G to Main Steam Valve House just outside the containment building wall, is designed as safety grade because of its safety function. According to design code the safety related system shall be designed to protect against dynamic effects that may results from a pipe break on high energy lines such as FW system. And the FW system should be designed to minimize blowdown volume of S/G secondary side during the postulated pipe break. Also the FW system should be designed to prevent the initiation or to minimize the effects of water hammer transients which may be induced by the pipe break. This paper shows the results of the hydrodynamic loads induced by the pipe break and the optimized design parameters to mitigate water hammer loads of FW system for Shin-Kori Nuclear Power Plant Unit 3 and 4 (SKN 3 and 4)

  3. Numerical and experimental analysis of the vibratory behavior of a nuclear power plant piping system excitated by a pump

    International Nuclear Information System (INIS)

    Vatin, E.; Guillou, J.; Tephany, F.; Trollat, C.

    1993-08-01

    This paper presents a study on the dynamic response of piping systems installed in the French 1300 MWe Nuclear Power Plants. Variations in pressure are generated by a multi-staged centrifugal pump mounted on the piping system and provide a dynamic excitation of the pipe. This type of dynamic loading has led to nozzle cracks for some of the pipes, whereas, for other installations, it has not be found detrimental. This study presents an explanation of the different dynamic behavior observed at the various plants. To this end, a numerical model, calibrated with on-site measurements, is impleted. (authors). 8 figs., 1 tab., 5 refs

  4. Heat Pipes

    Science.gov (United States)

    1990-01-01

    Bobs Candies, Inc. produces some 24 million pounds of candy a year, much of it 'Christmas candy.' To meet Christmas demand, it must produce year-round. Thousands of cases of candy must be stored a good part of the year in two huge warehouses. The candy is very sensitive to temperature. The warehouses must be maintained at temperatures of 78-80 degrees Fahrenheit with relative humidities of 38- 42 percent. Such precise climate control of enormous buildings can be very expensive. In 1985, energy costs for the single warehouse ran to more than 57,000 for the year. NASA and the Florida Solar Energy Center (FSEC) were adapting heat pipe technology to control humidity in building environments. The heat pipes handle the jobs of precooling and reheating without using energy. The company contacted a FSEC systems engineer and from that contact eventually emerged a cooperative test project to install a heat pipe system at Bobs' warehouses, operate it for a period of time to determine accurately the cost benefits, and gather data applicable to development of future heat pipe systems. Installation was completed in mid-1987 and data collection is still in progress. In 1989, total energy cost for two warehouses, with the heat pipes complementing the air conditioning system was 28,706, and that figures out to a cost reduction.

  5. An appraisal of procedures used to give the criterion for instability of a through-wall circumferential crack in a stainless steel piping system

    International Nuclear Information System (INIS)

    Smith, E.

    1989-01-01

    Against the background of the problem of intergranular stress corrosion cracking of 304 stainless steel in Boiling Water Reactor piping systems, this paper presents a critical appraisal of procedures that are currently used to give the criterion for instability of a through-wall circumferential crack in a stainless steel piping system. Particular attention is focussed on a simple procedure developed by Cotter, Chang and Zahoor, which has been applied to specific piping systems, the objective being to underpin its viability. The considerations are applicable to not only Boiling Water Reactor piping systems, but to other piping systems where pipe failure due to circumferential cracking is a potential problem. (author)

  6. Strain Limits within the Scope of the Integrity Assessment of Piping Systems

    International Nuclear Information System (INIS)

    Mutz, Alexander

    2008-01-01

    Allowable stresses in nuclear power plant piping resulting from loading conditions to be considered in Germany are determined on the basis of the German Safety Standards of the Nuclear Safety Standards Commission, KTA. The limitation of the different stress categories within the analysis of the mechanical behaviour is based on a linear elastic material behaviour. Because of the ductile material used in high energy nuclear piping, a more realistic assessment can be performed on the basis of allowable strains using elastic plastic material behaviour. In the present work comparison between the analysis of piping systems considering the elastic material model and the actual elastic plastic material behaviour is performed. The possibilities of allocating plastic strains to calculated elastic stresses is discussed. A parametric study on straight pipes with the actual elastic plastic material model under pure bending is the basis of deriving the elastic plastic strains for the calculated elastic stresses. Strain limits are suggested which correspond to the different stress categories. The aim is to utilize the deformation possibilities of ductile materials used in German nuclear piping and the allocation of maximum strains to different load categories. Keywords: strain limit, ductile material, stress category. (author)

  7. Strain Limits within the Scope of the Integrity Assessment of Piping Systems

    Energy Technology Data Exchange (ETDEWEB)

    Mutz, Alexander [EnBW, Durlacher Allee 93, Karlsruhe 76131 (Germany)

    2008-07-01

    Allowable stresses in nuclear power plant piping resulting from loading conditions to be considered in Germany are determined on the basis of the German Safety Standards of the Nuclear Safety Standards Commission, KTA. The limitation of the different stress categories within the analysis of the mechanical behaviour is based on a linear elastic material behaviour. Because of the ductile material used in high energy nuclear piping, a more realistic assessment can be performed on the basis of allowable strains using elastic plastic material behaviour. In the present work comparison between the analysis of piping systems considering the elastic material model and the actual elastic plastic material behaviour is performed. The possibilities of allocating plastic strains to calculated elastic stresses is discussed. A parametric study on straight pipes with the actual elastic plastic material model under pure bending is the basis of deriving the elastic plastic strains for the calculated elastic stresses. Strain limits are suggested which correspond to the different stress categories. The aim is to utilize the deformation possibilities of ductile materials used in German nuclear piping and the allocation of maximum strains to different load categories. Keywords: strain limit, ductile material, stress category. (author)

  8. Hybrid solar lighting distribution systems and components

    Science.gov (United States)

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

    2011-07-05

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

  9. Field measurement of the piping system vibration of Ko-Ri unit 4 during the load-following operation

    International Nuclear Information System (INIS)

    Chung, Tae-Young; Hong, Sung-Yull; Kim, Bum-Nyun.

    1989-01-01

    During the load-following operation of nuclear power plants, flow rate, temperature, and pressure in the piping system can be varied by changing the electric power output level, and these variations can cause different vibration phenomena in the piping system. The piping system vibration is important because it is directly related to the dynamic stress of the piping system and can affect the life of the piping system through structural fatigue. An assessment of vibration levels for the classes II and III piping systems of the Ko-Ri Unit 4950-MW nuclear power plant was performed according to the given pattern of the load-following operation to study its feasibility from the viewpoint of piping system vibration. The classes II and III piping system vibration of the Ko-Ri Unit 4 may not cause any potential problem under the given pattern of the load-following operation; however, it is recommended that long-term operation in the 85 to 95% power range be avoided as much as possible

  10. Comparison of fracture toughness values from large-scale pipe system tests and C(T) specimens

    International Nuclear Information System (INIS)

    Olson, R.; Scott, P.; Marschall, C.; Wilkowski, G.

    1993-01-01

    Within the International Piping Integrity Research Group (IPIRG) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe and the experiments were conducted at 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment of standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria. 4 refs., 14 figs., 1 tab

  11. Management of manufacture and installation of plant pipings by bar code system

    International Nuclear Information System (INIS)

    Suwa, Minoru

    1995-01-01

    As for the piping system of nuclear power plants, the number of parts is very large, and the mill sheet is attached to each part, therefore, it is necessary to manage them individually, and large man power is required. In order to resolve the delay of mechanization in the factory, bar code system was adopted on full scale. At the time of taking parts out from the store, bar code labels are stuck to all piping parts. By this means, all the processes of manufacture and inspection are managed with a computer, and it is useful for labor saving and the prevention of mistaken input. This system is centering around the system of the progress management for piping manufacture, and is operated by being coupled with respective systems of production design, order and inventory, mill sheet management and installation management. The management of production design, manufacture, inspection and installation is explained. There is the problem of sticking bar code labels again as the labels become dirty or parts pass through coating and pickling processes. The direct carving of bar codes on parts by laser marker was tried, and it was successful for stainless steel, but in carbon steel pipes, it was hard to read. It is desirable to develop the bar codes which endure until the end of plant life. (K.I.)

  12. Uncertainty analysis technique of dynamic response and cumulative damage properties of piping system

    International Nuclear Information System (INIS)

    Suzuki, Kohei; Aoki, Shigeru; Hara, Fumio; Hanaoka, Masaaki; Yamashita, Tadashi.

    1982-01-01

    It is a technologically important subject to establish the method of uncertainty analysis statistically examining the variation of the earthquake response and damage properties of equipment and piping system due to the change of input load and the parameters of structural system, for evaluating the aseismatic capability and dynamic structural reliability of these systems. The uncertainty in the response and damage properties when equipment and piping system are subjected to excessive vibration load is mainly dependent on the irregularity of acting input load such as the unsteady vibration of earthquakes, and structural uncertainty in forms and dimensions. This study is the basic one to establish the method for evaluating the uncertainty in the cumulative damage property at the time of resonant vibration of piping system due to the disperse of structural parameters with a simple model. First, the piping models with simple form were broken by resonant vibration, and the uncertainty in the cumulative damage property was evaluated. Next, the response analysis using an elasto-plastic mechanics model was performed by numerical simulation. Finally, the method of uncertainty analysis for response and damage properties by the perturbation method utilizing equivalent linearization was proposed, and its propriety was proved. (Kako, I.)

  13. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  14. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

  15. Heat pipes in modern heat exchangers

    International Nuclear Information System (INIS)

    Vasiliev, Leonard L.

    2005-01-01

    Heat pipes are very flexible systems with regard to effective thermal control. They can easily be implemented as heat exchangers inside sorption and vapour-compression heat pumps, refrigerators and other types of heat transfer devices. Their heat transfer coefficient in the evaporator and condenser zones is 10 3 -10 5 W/m 2 K, heat pipe thermal resistance is 0.01-0.03 K/W, therefore leading to smaller area and mass of heat exchangers. Miniature and micro heat pipes are welcomed for electronic components cooling and space two-phase thermal control systems. Loop heat pipes, pulsating heat pipes and sorption heat pipes are the novelty for modern heat exchangers. Heat pipe air preheaters are used in thermal power plants to preheat the secondary-primary air required for combustion of fuel in the boiler using the energy available in exhaust gases. Heat pipe solar collectors are promising for domestic use. This paper reviews mainly heat pipe developments in the Former Soviet Union Countries. Some new results obtained in USA and Europe are also included

  16. CaSO4 Scale Formation on Vibrated Piping System in the Presence Citric Acid

    Science.gov (United States)

    Mangestiyono, W.; Jamari, J.; Muryanto, S.; Bayuseno, A. P.

    2018-02-01

    Vibration in many industries commonly generated by the operation mechanical equipment such as extruder, mixer, blower, compressor, turbine, generator etc. Vibration propagates into the floor and attacks the pipe around those mechanical equipment. In this paper, the influence of vibration in a pipe on the CaSO4 scale formation was investigated to understand the effect of vibration on the kinetics, mass of scale, crystal phases and crystal polymorph. To generate vibration force, mechanical equipment was prepared consisted of electrical motor, crankshaft, connecting rod and a vibration table at where test pipe section mounted. Deposition rate increased significantly when the vibration affected to the system i.e. 0.5997 and 1.6705 gr/hr for vibration frequency 4.00 and 8.00 Hz. The addition 10.00 ppm of citric acid declined the deposition rate of 8 Hz experiment from 3.4599 gr/hr to 2.2865 gr/hr.

  17. A modeling approach for district heating systems with focus on transient heat transfer in pipe networks

    DEFF Research Database (Denmark)

    Mohammadi, Soma; Bojesen, Carsten

    2015-01-01

    the temperature in DH systems. The main focus is on modeling transient heat transfer in pipe networks regarding the time delays between the heat supply unit and the consumers, the heat loss in the pipe networks and the consumers’ dynamic heat loads. A pseudo-dynamic approach is adopted and also the implicit...... district heating networks [DHN] characteristics. This paper is presenting a new developed model, which reflects the thermo-dynamic behavior of DHN. It is designed for tree network topologies. The purpose of the model is to serve as a basis for applying a variety of scenarios towards lowering...... finite element method is applied to simulate transient temperature changes in pipe networks. The model is calculating time series data related to supply temperature to the DHN from heat production units, heat loads and return temperature related to each consumer to calculate dynamic temperature changes...

  18. Optimization of distribution piping network in district cooling system using genetic algorithm with local search

    International Nuclear Information System (INIS)

    Chan, Apple L.S.; Hanby, Vic I.; Chow, T.T.

    2007-01-01

    A district cooling system is a sustainable means of distribution of cooling energy through mass production. A cooling medium like chilled water is generated at a central refrigeration plant and supplied to serve a group of consumer buildings through a piping network. Because of the substantial capital investment involved, an optimal design of the distribution piping configuration is one of the crucial factors for successful implementation of the district cooling scheme. In the present study, genetic algorithm (GA) incorporated with local search techniques was developed to find the optimal/near optimal configuration of the piping network in a hypothetical site. The effect of local search, mutation rate and frequency of local search on the performance of the GA in terms of both solution quality and computation time were investigated and presented in this paper

  19. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    International Nuclear Information System (INIS)

    Kaliatka, A.; Valinčius, M.

    2012-01-01

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  20. Provisions and active measures to preclude guillotine breaks in piping systems of pressurized water reactors

    International Nuclear Information System (INIS)

    Dorner, H.; Michel, E.

    1983-01-01

    The conditions and active measures which preclude a spontaneous failure of pipings are shown. With the basic safety concept a quality standard is achieved characterized by high-grade material properties, a structure that is adequate to the loads to which the components will be subjected in service and is amenable to inspection, precise load and stress evaluation, optimized manufacturing and operation monitoring. The possible failure types are described and the safety against failure is assessed. (author) [pt

  1. The development of the design method of nuclear piping system supported by elasto-plastic support structures (Part 1)

    International Nuclear Information System (INIS)

    Endo, R.; Murota, M.; Kawahata, J.-I.; Sato, T.; Mekomoto, Y.; Takayama, Y.; Kobayashi, H.; Hirose, J.

    1993-01-01

    The conventional aseismic design method of nuclear piping system is very conservative because of the accumulation of various safety factors in the design process, and nuclear piping systems are thought to have a large safety margin. Considering this situation, we promoted research to further rationalize nuclear power plants by reducing the amount of support structures and reducing the piping seismic response through vibration energy absorption resulting from the elasto-plastic behavior of piping support structures. The research has the following three stages. In the first stage, we select conventional piping support structures in Japanese light-water reactors that exhibit elasto-plastic behavior, and study the displacement dependency and the vibration frequency dependency on the stiffness and the energy absorption by testing their model. In the second stage, we make a piping test model with support structures whose characteristics have already been obtained, and perform vibration tests on a shaking table. In this way, we analyze the piping vibration characteristics by sinusoidal wave sweep tests and the piping response characteristics by seismic wave vibration tests, when the support structures are in an elasto-plastic condition. In the third stage, a general method is developed to evaluate the characteristics of the support structures obtained in the tests and it is applied to the evaluation of the characteristics of general support structures. A simplified analysis method is developed to evaluate the piping seismic response using the piping model test result. To expand the results mentioned above, we are developing a seismic design method of piping systems that allows support structures to have elasto-plastic behaviour. This paper reports the results of experiments conducted under the joint research program of Japanese electric power companies with support elements in the first stage and those with piping models in the second stage

  2. Response of piping system with semi-active variable stiffness damper under tri-directional seismic excitation

    International Nuclear Information System (INIS)

    Praveen Kumar; Jangid, R.S.; Reddy, G.R.

    2013-01-01

    Highlights: ► Piping system with semi-active variable stiffness damper is investigated under different seismic excitations. ► Switching control law and modified switching control law are adopted. ► There exist an optimum parameters of the SAVSD. ► Substantial reduction of the seismic response of piping system with SAVSD is observed. ► Good amount of energy dissipation is observed. -- Abstract: Seismic loads on piping system due to earthquakes can cause excessive vibrations, which can lead to serious instability resulting in damage or complete failure. In this paper, semi-active variable stiffness dampers (SAVSDs) have been studied to mitigate seismic response and vibration control of piping system used in the process industries, fossil and fissile fuel power plant. The SAVSD changes its stiffness depending upon the piping response and accordingly adds the control forces in the piping system. A study is conducted on the performance of SAVSD due to variation in device stiffness ratios in the switching control law and modified switching control law, which plays an important role in the present control algorithm of the damper. The effectiveness of the SAVSD in terms of reduction in the responses, namely, displacements, accelerations and base shear of the piping system is investigated by comparing uncontrolled responses under four different artificial earthquake motions with increasing amplitudes. The analytical results demonstrate that the SAVSDs under particular optimum parameters are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping systems

  3. Response of piping system with semi-active variable stiffness damper under tri-directional seismic excitation

    Energy Technology Data Exchange (ETDEWEB)

    Praveen Kumar, E-mail: praveen@barc.gov.in [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jangid, R.S. [Department of Civil Engineering, Indian Institute of Technology Bombay, Powai, Mumbai 400076 (India); Reddy, G.R. [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2013-05-15

    Highlights: ► Piping system with semi-active variable stiffness damper is investigated under different seismic excitations. ► Switching control law and modified switching control law are adopted. ► There exist an optimum parameters of the SAVSD. ► Substantial reduction of the seismic response of piping system with SAVSD is observed. ► Good amount of energy dissipation is observed. -- Abstract: Seismic loads on piping system due to earthquakes can cause excessive vibrations, which can lead to serious instability resulting in damage or complete failure. In this paper, semi-active variable stiffness dampers (SAVSDs) have been studied to mitigate seismic response and vibration control of piping system used in the process industries, fossil and fissile fuel power plant. The SAVSD changes its stiffness depending upon the piping response and accordingly adds the control forces in the piping system. A study is conducted on the performance of SAVSD due to variation in device stiffness ratios in the switching control law and modified switching control law, which plays an important role in the present control algorithm of the damper. The effectiveness of the SAVSD in terms of reduction in the responses, namely, displacements, accelerations and base shear of the piping system is investigated by comparing uncontrolled responses under four different artificial earthquake motions with increasing amplitudes. The analytical results demonstrate that the SAVSDs under particular optimum parameters are very effective and practically implementable for the seismic response mitigation, vibration control and seismic requalification of piping systems.

  4. Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

    International Nuclear Information System (INIS)

    Wang Chenglong; Tian Wenxi; Su Guanghui; Zhang Dalin; Wu Yingwei; Qiu Suizheng

    2013-01-01

    The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

  5. Qualification of a Method to Calculate the Irrecoverable Pressure Loss in High Reynolds Number Piping Systems

    Energy Technology Data Exchange (ETDEWEB)

    Sigg, K. C.; Coffield, R. D.

    2002-09-01

    High Reynolds number test data has recently been reported for both single and multiple piping elbow design configurations at earlier ASME Fluid Engineering Division conferences. The data of these studies ranged up to a Reynolds number of 42 x 10[sup]6 which is significantly greater than that used to establish design correlations before the data was available. Many of the accepted design correlations, based on the lower Reynolds number data, date back as much as fifty years. The new data shows that these earlier correlations are extremely conservative for high Reynolds number applications. Based on the recent high Reynolds number information a new recommended method has been developed for calculating irrecoverable pressure loses in piping systems for design considerations such as establishing pump sizing requirements. This paper describes the recommended design approach and additional testing that has been performed as part of the qualification of the method. This qualification testing determined the irrecoverable pressure loss of a piping configuration that would typify a limiting piping section in a complicated piping network, i.e., multiple, tightly coupled, out-of-plane elbows in series under high Reynolds number flow conditions. The overall pressure loss measurements were then compared to predictions, which used the new methodology to assure that conservative estimates for the pressure loss (of the type used for pump sizing) were obtained. The recommended design methodology, the qualification testing and the comparison between the predictions and the test data are presented. A major conclusion of this study is that the recommended method for calculating irrecoverable pressure loss in piping systems is conservative yet significantly lower than predicted by early design correlations that were based on the extrapolation of low Reynolds number test data.

  6. Nonlinear dynamic analysis of piping systems using the pseudo force method

    International Nuclear Information System (INIS)

    Prachuktam, S.; Bezler, P.; Hartzman, M.

    1979-01-01

    Simple piping systems are composed of linear elastic elements and can be analyzed using conventional linear methods. The introduction of constraint springs separated from the pipe with clearance gaps to such systems to cope with the pipe whip or other extreme excitation conditions introduces nonlinearities to the system, the nonlinearities being associated with the gaps. Since these spring-damper constraints are usually limited in number, descretely located, and produce only weak nonlinearities, the analysis of linear systems including these nonlinearities can be carried out by using modified linear methods. In particular, the application of pseudo force methods wherein the nonlinearities are treated as displacement dependent forcing functions acting on the linear system were investigated. The nonlinearities induced by the constraints are taken into account as generalized pseudo forces on the right-hand side of the governing dynamic equilibrium equations. Then an existing linear elastic finite element piping code, EPIPE, was modified to permit application of the procedure. This option was inserted such that the analyses could be performed using either the direct integration method or via a modal superposition method, the Newmark-Beta integration procedure being employed in both methods. The modified code was proof tested against several problems taken from the literature or developed with the nonlinear dynamics code OSCIL. The problems included a simple pipe loop, cantilever beam, and lumped mass system subjected to pulsed and periodic forcing functions. The problems were selected to gage the overall accuracy of the method and to insure that it properly predicted the jump phenomena associated with nonlinear systems. (orig.)

  7. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  8. A study on prediction of metal loss by flow-accelerated corrosion in the CANDU NPP secondary piping systems

    International Nuclear Information System (INIS)

    Shim, S. H.; Song, J. S.; Yoon, K. B.; Hwang, K. M.; Jin, T. E.; Lee, S. H.; Kim, W. S.

    2001-01-01

    Flow-Accelerated Corrosion(FAC) is a phenomenon that results in metal loss from piping, vessels, and equipment made of carbon steel. FAC occurs only under certain conditions of flow, chemistry, geometry, and material. Unfortunately, those conditions are in much of the high-energy piping in nuclear and fossil-fueled power plants. Also, for domestic NPP secondary pipings whose operating time become longer, more evidences of FAC have been reported. The authors are studying on FAC management using CHECWORKS, computer code developed by EPRI. This paper is on the prediction results of metal loss by FAC in the one of CANDU type NPP secondary piping systems

  9. Review of the use of rigid and high-impact PVC pipes in natural gas distribution systems in the Netherlands

    Energy Technology Data Exchange (ETDEWEB)

    Mutter, F; Benjamin, P

    1974-08-01

    Because of a number of instances of stress corrosion cracking or crazing occurring in PVC pipes used in Dutch gas distribution systems, VEG-GASINSTITUUT began an intensive investigation of rigid PVC pipes and high-impact pipes in distribution use under various conditions and with varying service lives. The work led to an investigation of laboratory testing techniques in which the stress-cracking phenomenon found in practice could be duplicated under controllable conditions. Pipes of various materials were examined for their resistance to stress cracking, then this resistance was compared with other long- and short-term physical properties of the material.

  10. Fire protection in Angra-2 nuclear power plant. The use of fire protection collars on plastic piping systems

    International Nuclear Information System (INIS)

    Oliveira Segabinaze, R. de

    1994-01-01

    The object of this paper is to briefly the use of fire protection collars on plastic piping systems passing through wall and floor penetration. The fire protection collars consist of a stainless steel housing, in which the leading edges of two pivoting plates are in constant pressure contact with the pipe. In case of fire these plates react on the softened pipe with a guillotine action, thereby stopping the flow; within the housing a foam material expands to fill the space when subject to the heat of the fire. The piping project has to be modified to permit the fixing of the collars to walls and floor penetrations. (author). 2 refs, 9 figs

  11. OCSEGen: Open Components and Systems Environment Generator

    Science.gov (United States)

    Tkachuk, Oksana

    2014-01-01

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

  12. Sound transmission through pipe systems and into plate structures in buildings. A simplified sea model

    NARCIS (Netherlands)

    Bron-van der Jagt, G.S.

    2007-01-01

    In the study presented, it has been investigated whether Statistical Energy Analysis (SEA) could be applied in a simplified way as a framework for a prediction model regarding noise due to service equipment, specifically sound transmission within (plastic wastewater) pipe systems and between these

  13. 46 CFR 28.815 - Bilge pumps, bilge piping, and dewatering systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Bilge pumps, bilge piping, and dewatering systems. 28... REQUIREMENTS FOR COMMERCIAL FISHING INDUSTRY VESSELS Aleutian Trade Act Vessels § 28.815 Bilge pumps, bilge... fixed, self priming, powered, bilge pump, having a minimum capacity rating of 50 gallons per minute...

  14. Nuclear power plant steam pipes repairing with TIRANT 3 robot system

    International Nuclear Information System (INIS)

    Soto Tomas, Marcelo; Curiel Nieva, Marceliano; Monzo Blasco, Enrique; Rodriguez, Salvador Pineda; Vaquer Perez, Juan I.

    2011-01-01

    A typical application functions covering the steam pipes inner surface in coal-fired power station and nuclear power plants. The results of this process are spectacular in terms of protection against corrosion and abrasion, but its application has conditioning factors, such as: Severe application conditions for workers. Due to the postural position (usually kneeling) in small diameter pipes and working with fireproof clothing and masks with outdoor air supplying, due to fumes, sparks and molten metal particles, radiological contamination, confined space, poor lighting... Coating uniformity. As metallization is a manual process, the carried out measurements show small variations in the thickness of the coating, always within the tolerance limits established by the applicable regulations and quality assurance. For all these reasons, Grupo Dominguis has developed the TIRANT 3 robot, a worldwide innovative system, for metallization of steam pipes inner surface. TIRANT 3 robot is teleoperated from outside of the pipe, so that human intervention is reduced to the operations of robot positioning and change of metallization wire. As it is an independent system of the human factor, metallization process performance is significantly increased by reducing rest periods due only to the robot maintenance. Likewise, TIRANT 3 system permits to increase resulting coating uniformity, and thus its resistance, keeping selected parameters constant depending on required type and thickness of wire. TIRANT 3 system has successfully worked in 2010 during the stops refueling of the Units I and II of Laguna Verde nuclear power plant in Mexico. (author)

  15. Piping Stress analysis for primary system of nuclear power plant AP-600

    International Nuclear Information System (INIS)

    Tjahjono, Hendro; Arhatari, B.D.; W, Pustandyo; Sitandung, J.B; Sudarmaji, Djoko

    1999-01-01

    Piping stress analysis for AP-600 primary system has been done using software CAEPIPE and PS-CAEPIPE. The loading applied to the system are static and seismic category I and II piping in reactor building have been analysed, those are PXS-900, CVS-110, PCS-030, CAS-700 and CCS-050. These system contain pipes with the normal diameter of 1 , 2 , 4 a nd 8 . The design pressures are in the range of 150oF to 300oF. The acceleration taken as input in PS-CAEPIPE is based on seismic response spectra of floor the piping is located. In CAEPIPE, the acceleration taken from the peak of response spectra multiplied by 1.7 all of the acceleration in this case are no more than 0.36g. The result shows that after locating some supports, all system are acceptable without snubbers. The maximum stress are 11210 psi for deadweight load and 35593 psi for total load (the allowable values are 15000 psi and 45000 psi). The maximum displacement are 0.123 in for deadweight load, 1.474 in for hot load seismic load (the allowable values are 0.125 in for deadweight and 2.5 in for total load). The difference results of the both software is mainly in seismic calculation where mare parameters can be evaluated by PS-CAEPIPE including to evaluate valves acceleration in seismic condition

  16. ANALYSIS OF MATERIALS IN AN EXPERIMENTAL TESTING PIPE SYSTEM FOR AN INHIBITOR OF MUSSEL KILL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel P. Molloy

    2003-06-04

    A comprehensive series of 16 laboratory experiments demonstrated that the presence of vinyl tubing within a recirculating pipe system was responsible for lowering zebra mussel kill following treatment with the bacterium Pseudomonas fluorescens. All vinyl tubing was replaced in all testing units with silicone tubing, and high mussel kill (>95%) was then obtained.

  17. High frequency statistical energy analysis applied to fluid filled pipe systems

    NARCIS (Netherlands)

    Beek, P.J.G. van; Smeulers, J.P.M.

    2013-01-01

    In pipe systems, carrying gas with high velocities, broadband turbulent pulsations can be generated causing strong vibrations and fatigue failure, called Acoustic Fatigue. This occurs at valves with high pressure differences (i.e. chokes), relief valves and obstructions in the flow, such as sharp

  18. Analysis of nuclear piping system seismic tests with conventional and energy absorbing supports

    International Nuclear Information System (INIS)

    Park, Y.; DeGrassi, G.; Hofmayer, C.; Bezler, P.; Chokshi, N.

    1997-01-01

    Large-scale models of main steam and feedwater piping systems were tested on the shaking table by the Nuclear Power Engineering Cooperation (NUPEC) of Japan, as part of the Seismic Proving Test Program. This paper describes the linear and nonlinear analyses performed by NRC/BNL and compares the results to the test data

  19. Digital Components in Swedish NPP Power Systems

    International Nuclear Information System (INIS)

    Karlsson, Mattias; Eriksson, Tage

    2015-01-01

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

  20. The spatial distribution of pollutants in pipe-scale of large-diameter pipelines in a drinking water distribution system.

    Science.gov (United States)

    Liu, Jingqing; Chen, Huanyu; Yao, Lingdan; Wei, Zongyuan; Lou, Liping; Shan, Yonggui; Endalkachew, Sahle-Demessie; Mallikarjuna, Nadagouda; Hu, Baolan; Zhou, Xiaoyan

    2016-11-05

    In large-diameter drinking water pipelines, spatial differences in hydraulic and physiochemical conditions may also result in spatial variations in pipe corrosion, biofilm growth and pollutant accumulation. In this article, the spatial distributions of various metals and organic contaminants in two 19-year-old grey cast iron pipes which had an internal diameter of 600mm (DN600), were investigated and analyzed by Atomic Absorption Spectrometry, Gas Chromatography-Mass Spectrometry, Energy Dispersive Spectrometer, X-ray Diffraction, etc. The spatial distribution of heavy metals varied significantly across the pipe section, and iron, manganese, lead, copper, and chromium were highest in concentration in the upper portion pipe-scales. However, the highest aluminum and zinc content was detected in the lower portion pipe-scales. Apart from some common types of hydrocarbons formed by microbial metabolites, there were also some microalgae metabolites and exogenous contaminants accumulated in pipe-scale, which also exhibited high diversity between different spatial locations. The spatial distributions of the physical and chemical properties of pipe-scale and contaminants were quite different in large-diameter pipes. The finding put forward higher requirements on the research method about drinking water distribution system chemical safety. And the scientific community need understand trend and dynamics of drinking water pipe systems better. Copyright © 2016 Elsevier B.V. All rights reserved.

  1. Structured Performance Analysis for Component Based Systems

    OpenAIRE

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

    2012-01-01

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

  2. Application study of the heat pipe to the passive decay heat removal system of the modular HTR

    International Nuclear Information System (INIS)

    Ohashi, K.; Okamoto, F.; Hayakawa, H.; Hayashi, T.

    2001-01-01

    To investigate the applicability of the heat pipe to the decay hat removal (DHR) system of the modular HTRs, preliminary study of the Heat Pipe DHR System was performed. The results show that the Heat Pipe DHR System is applicable to the modular HTRs and its heat removal capability is sufficient. Especially by applying the variable conductance heat pipe, the possibility of a fully passive DHR system with lower heat loss during normal operation is suggested. The experiments to obtain the fundamental characteristics data of the variable conductance heat pipe were carried out. The experimental results show very clear features of self-control characteristics. The experimental results and the experimental analysis results are also shown. (author)

  3. Design evaluation on sodium piping system and comparison of the design codes

    International Nuclear Information System (INIS)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

    2015-01-01

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  4. Design evaluation on sodium piping system and comparison of the design codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

    2015-03-15

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  5. Design Evaluation of a Piping System in the SELFA Sodium Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seok-Kwon; Jo, Young-Chul; Lee, Hyeong-Yeon; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, design evaluations on the SELFA piping system has been conducted according to the ASME B31.1 and RCC-MRx RD-3600. The conservatism of the two codes was quantified based on the evaluation results. It was shown that B31.1 was more conservative for the sustained loads while less conservative for thermal expansion loads when compare with those of RD-3600. However, all the evaluation results according to the two codes were within the code allowables. There are two main piping systems in the SELFA test loop. In this study, the integrity of the SELFA piping system has been evaluated according to the two design-by-rule (DBR) codes of ASME B31.1 and RCC-MRx RD-3600. B31.1 is an industry design code for power piping while RD-3600 is a class 3 nuclear DBR code. The conservatism of the two codes was quantified based on the evaluation results as per the two DBR codes. The sodium test facility of the SELFA is under construction at KAERI for the investigation of thermo-hydraulic behavior of finned-tube sodium-to-air heat exchanger.

  6. Comparison study of time history and response spectrum responses for multiply supported piping systems

    International Nuclear Information System (INIS)

    Wang, Y.K.; Subudhi, M.; Bezler, P.

    1983-01-01

    In the past decade, several investigators have studied the problem of independent support excitation of a multiply supported piping system to identify the real need for such an analysis. This approach offers an increase in accuracy at a small increase in computational costs. To assess the method, studies based on the response spectrum approach using independent support motions for each group of commonly connected supports were performed. The results obtained from this approach were compared with the conventional envelope spectrum and time history solutions. The present study includes a mathematical formulation of the independent support motion analysis method suitable for implementation into an existing all purpose piping code PSAFE2 and a comparison of the solutions for some typical piping system using both Time History and Response Spectrum Methods. The results obtained from the Response Spectrum Methods represent the upper bound solution at most points in the piping system. Similarly, the Seismic Anchor Movement analysis based on the SRP method over predicts the responses near the support points and under predicts at points away from the supports

  7. Remaining life case history studies for high energy piping systems using equivalent stress

    International Nuclear Information System (INIS)

    Cohn, M.J.

    1987-01-01

    As the development of plant life extension for high energy piping systems is progressing, conventional piping system design methodologies are also being reevaluated. Traditional guidelines such as American National Standard Institute/American Society of Mechanical Engineers B31.1 (ANSI/ASME) were developed for plants having design lives in the 25- to 30-year regime based upon relatively short-term base metal creep data. These guidelines use a simplified approach for the piping analysis. Two types of stress criteria must be satisfied. The first type is longitudinal plus torsion stress checks for several types of loading conditions versus the material allowable stresses. The second type is an independent minimum wall thickness check which considers the hoop stress versus the material allowable stress. Seven case histories have been evaluated to estimate the minimum piping system creep life based on the current ANSI/ASME B31.1 finite element type of analysis, which is a traditional approach, versus a multiaxial stress state type of analysis. In nearly every case, the equivalent stress methodology predicted significantly higher stresses. Consequently, the equivalent stress methodology resulted in 11 to 96% lower time to rupture values as compared to the values predicted using ANSI/ASME B31.1 stresses

  8. Effect of PVRC damping with independent support motion response spectrum analysis of piping systems

    International Nuclear Information System (INIS)

    Wang, Y.K.; Bezler, P.; Shteyngart, S.

    1986-01-01

    The Technical Committee for Piping Systems of the Pressure Vessel Research Committee (PVRC) has recommended new damping values to be used in the seismic analyses of piping systems in nuclear power plants. To evaluate the effects of coupling these recommendations with the use of independent support motion analyses methods, two sets of seismic analyses have been carried out for several piping systems. One set based on the use of uniform damping as specified in Regulatory Guide 1.61, the other based on the PVRC recommendations. In each set the analyses were performed using independent support motion time history and response spectrum methods as well as the envelope spectrum method. In the independent response spectrum analyses, 14 response estimates were in fact obtained by considering different combination procedures between the support group contributions and all sequences of combinations between support groups, modes and directions. For each analysis set, the response spectrum results were compared with time history estimates of those results. Comparison tables were then prepared depicting the percentage by which the response spectrum estimates exceeded the time history estimates. By comparing the result tables between both analysis sets, the impact of PVRC damping can be observed. Preliminary results show that the degree of exceedance of the response spectrum estimates based on PVRC damping is less than that based on uniform damping for the same piping problem. Expressed differently the results obtained if ISM methods are coupled with PVRC damping are not as conservative as those obtained using uniform damping

  9. Investigation on method of elasto-plastic analysis for piping system (benchmark analysis)

    International Nuclear Information System (INIS)

    Kabaya, Takuro; Kojima, Nobuyuki; Arai, Masashi

    2015-01-01

    This paper provides method of an elasto-plastic analysis for practical seismic design of nuclear piping system. JSME started up the task to establish method of an elasto-plastic analysis for nuclear piping system. The benchmark analyses have been performed in the task to investigate on method of an elasto-plastic analysis. And our company has participated in the benchmark analyses. As a result, we have settled on the method which simulates the result of piping exciting test accurately. Therefore the recommended method of an elasto-plastic analysis is shown as follows; 1) An elasto-plastic analysis is composed of dynamic analysis of piping system modeled by using beam elements and static analysis of deformed elbow modeled by using shell elements. 2) Bi-linear is applied as an elasto-plastic property. Yield point is standardized yield point multiplied by 1.2 times, and second gradient is 1/100 young's modulus. Kinematic hardening is used as a hardening rule. 3) The fatigue life is evaluated on strain ranges obtained by elasto-plastic analysis, by using the rain flow method and the fatigue curve of previous studies. (author)

  10. Design demonstrations for Category B tank system piping at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1994-02-01

    Demonstration of the design of the tank systems described in this report is stipulated by the Federal Facility Agreement (FFA) between the U.S. Environmental Protection Agency-Region IV, the Tennessee Department of Environment and Conservation, and the U.S. Department of Energy. This report provides a design demonstration of the secondary containment and ancillary equipment of 30 piping systems listed in the FFA as Category B (i.e., existing tank systems with secondary containment). The design demonstrations were developed using information obtained from design drawings (as-built when available), construction specifications, and interviews with facility operators. Each design demonstration addresses system conformance to the requirements of the FFA (Appendix F, Section C). Deficiencies or restrictions regarding the ability to demonstrate that each of the containment systems conforms to FFA requirements are noted in the discussion of each piping system

  11. Lifetime analysis of the THTR steam generator and piping system

    International Nuclear Information System (INIS)

    Kemter, F.; Gloeckner, H.J.; Fritz, H.U.; Koenig, H.

    1989-01-01

    For the life monitoring during operation of the water / steam circuit operated in the high temperature area and the steam-raising units in the THTR, the life monitoring program SLAP was developed. For highly loaded components the current components exploitation and the remaining available life can be determined during operation. A survey is given of the procedure in determining the life exploitation and of the program structure of SLAP. (DG) [de

  12. Heat pipe based cold energy storage systems for datacenter energy conservation

    International Nuclear Information System (INIS)

    Singh, Randeep; Mochizuki, Masataka; Mashiko, Koichi; Nguyen, Thang

    2011-01-01

    In the present paper, design and economics of the novel type of thermal control system for datacenter using heat pipe based cold energy storage has been proposed and discussed. Two types of cold energy storage system namely: ice storage system and cold water storage system are explained and sized for datacenter with heat output capacity of 8800 kW. Basically, the cold energy storage will help to reduce the chiller running time that will save electricity related cost and decrease greenhouse gas emissions resulting from the electricity generation from non-renewable sources. The proposed cold energy storage system can be retrofit or connected in the existing datacenter facilities without major design changes. Out of the two proposed systems, ice based cold energy storage system is mainly recommended for datacenters which are located in very cold locations and therefore can offer long term seasonal storage of cold energy within reasonable cost. One of the potential application domains for ice based cold energy storage system using heat pipes is the emergency backup system for datacenter. Water based cold energy storage system provides more compact size with short term storage (hours to days) and is potential for datacenters located in areas with yearly average temperature below the permissible cooling water temperature (∼25 o C). The aforesaid cold energy storage systems were sized on the basis of metrological conditions in Poughkeepsie, New York. As an outcome of the thermal and cost analysis, water based cold energy storage system with cooling capability to handle 60% of datacenter yearly heat load will provide an optimum system size with minimum payback period of 3.5 years. Water based cold energy storage system using heat pipes can be essentially used as precooler for chiller. Preliminary results obtained from the experimental system to test the capability of heat pipe based cold energy storage system have provided satisfactory outcomes and validated the proposed

  13. A heat pipe solar collector system for winter heating in Zhengzhou city, China

    Directory of Open Access Journals (Sweden)

    Zheng Hui-Fan

    2017-01-01

    Full Text Available A heat pipe solar collector system for winter heating is investigated both experimentally and theoretically. The hourly heat collecting capacity, water temperature and contribution rate of solar collector system based on Zhengzhou city typical sunshine are calculated. The study reveals that the heat collecting capacity and water temperature increases initially and then decreases, and the solar collector system can provide from 40% to 78% heating load for a 200 m2 villa with in Zhengzhou city from November to March.

  14. Check valve slam waterhammer in piping systems equipped with multiple parallel pumps

    International Nuclear Information System (INIS)

    Sponsel, J.; Bird, E.; Zarechnak, A.

    1993-01-01

    The low pressure safety injection system at the calvert cliff's plant is designed to provide cooling water to the reactor in the event of a postulated accident and for reactor cool-down and decay heat removal during normal maintenance and refueling. This system experienced repeated damage to the axial piping supports on the pump section and the discharge headers due to the check valve phenomenon. To determine the cause, testing was performed in both the LPSI and CCW systems

  15. Systems with randomly failing repairable components

    DEFF Research Database (Denmark)

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

    2005-01-01

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

  16. Design and Integrity Evaluation of High-temperature Piping Systems in the STELLA-2 Sodium Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seok-Kwon; Lee, Hyeong-Yeon; Eoh, JaeHyuk; Kim, Jong-Bum; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ju, Yong-Sun [KOASIS Inc., Daejeon (Korea, Republic of)

    2016-09-15

    In this study, elevated temperature design and integrity evaluation have been conducted using two different piping design codes for the high-temperature piping systems of sodium integral effect test loop for safety simulation and assessment(STELLA-2) being developed by KAERI(Korea Atomic Energy Research Institute). The design code of ASME B31.1 for power piping and French nuclear grade piping design guideline, RCC-MRx RD-3600 were applied, and conservatism of those codes was quantified based on the piping integrity evaluation results. The piping system of Model DHRS, Model IHTS and PSLS are to be installed in STELLA-2. The integrity evaluation results for the three piping systems according to the two design codes showed that integrity of the piping system was confirmed. As a code comparison result, ASME B31.1 was shown to be more conservative for sustained loads while RD-3600 was more conservative for thermal loads compared to B31.1.

  17. Survey on application of probabilistic fracture mechanics approach to nuclear piping

    International Nuclear Information System (INIS)

    Kashima, Koichi

    1987-01-01

    Probabilistic fracture mechanics (PFM) approach is newly developed as one of the tools to evaluate the structural integrity of nuclear components. This report describes the current status of PFM studies for pressure vessel and piping system in light water reactors and focuses on the investigations of the piping failure probability which have been undertaken by USNRC. USNRC reevaluates the double-ended guillotine break (DEGB) of rector coolant piping as a design basis event for nuclear power plant by using the PFM approach. For PWR piping systems designed by Westinghouse, two causes of pipe break are considered: pipe failure due to the crack growth and pipe failure indirectly caused by failure of component supports due to an earthquake. PFM approach shows that the probability of DEGB from either cause is very low and that the effect of earthquake on pipe failure can be neglected. (author)

  18. The development of design method of nuclear piping system supported by elasto-plastic support structures (part 2)

    International Nuclear Information System (INIS)

    Endo, R.; Murota, M.; Kawabata, J-I.; Hirose, J.; Nekomoto, Y.; Takayama, Y.; Kobayashi, H.

    1995-01-01

    The conventional seismic design method of nuclear piping system is very conservative because of the accumulation of various safety factors in the design process, and nuclear piping systems are thought to have a large safety margin. Considering this situations, research program was promoted to furthermore rationalize nuclear power plants by reducing the amount of support structures and reducing the piping's seismic response through vibration energy absorption resulting from the elasto-plastic behavior of piping support structures. The research had the following three stages. In the first stage, we selected conventional piping support structures in light-water reactors that exhibited elasto-plastic behavior, and studied the effect of displacement and the vibration frequency on the stiffness and on the energy absorption by testing these models. In the second stage, vibration tests were performed using piping models with support structures on shaking tables. The piping vibration characteristics were clarified by sinusoidal sweep tests and the piping response characteristics by seismic wave vibration tests when the support structures were in an elasto-plastic condition. In the third stage, a general method was developed to evaluate the characteristics of a variety of support structures in the tests. A simplified analysis method was also developed to evaluate the piping seismic response using the piping model test result. To expand the results mentioned above, we also established a new seismic design method of piping systems that allowed support structures to have elasto-plastic behavior. This paper reports the newly developed seismic design method based on the results of experiments conducted under the joint research program of Japanese electric power companies (The Japan Atomic Power Co., Hokkaido EPC, Tohoku EPC, Tokyo EPC, Chubu EPC, Hokuriku EPC, Kansai EPC, Chugoku EPC, Shikoku EPC, Kyushu EPC) and nuclear plant makers (Hitachi Ltd., Toshiba Co., MHI Ltd., HEC Ltd

  19. Development of total systems of piping stress analysis and evaluation: ISAPPS

    International Nuclear Information System (INIS)

    Oki, Teizaburo; Koyanagi, Ryoichi; Fukuda, Masanao

    1978-01-01

    IHI has developed the systems of piping stress analysis and evaluation: ISAPPS (IHI Stress Analysis Program for Piping Systems), which are further described in this paper. In addition, the results of structural analysis and heat transfer analysis were confirmed. An example of stress evaluation in accordance with the modified ASME Code Sec. III is shown. ISAPPS consists of the following seven parts, and is designed for easy adoption of other programs by making modifications. 1. Piping design oriented language programs 2. Structural analysis programs 3. Isometric plotting programs 4. Multi-file dumping program 5. Load combination program 6. Heat transfer program 7. Stress evaluation programs As one of the examples of structural analysis programs, IHI make use of the modified SAP IV developed by the University of California. Evaluations of stresses are performed in accordance with: 1. ASME Boiler and Pressure Vessel Code, Sec. III Class 1, 2 and 3 2. ANSI Code, B31.1 and B31.3 3. MITI (Ministry of International Trade and Industry ) Code ISAPPS is very useful for design of nuclear and chemical pipings and so on. (author)

  20. Design of cylindrical pipe automatic welding control system based on STM32

    Science.gov (United States)

    Chen, Shuaishuai; Shen, Weicong

    2018-04-01

    The development of modern economy makes the demand for pipeline construction and construction rapidly increasing, and the pipeline welding has become an important link in pipeline construction. At present, there are still a large number of using of manual welding methods at home and abroad, and field pipe welding especially lacks miniature and portable automatic welding equipment. An automated welding system consists of a control system, which consisting of a lower computer control panel and a host computer operating interface, as well as automatic welding machine mechanisms and welding power systems in coordination with the control system. In this paper, a new control system of automatic pipe welding based on the control panel of the lower computer and the interface of the host computer is proposed, which has many advantages over the traditional automatic welding machine.