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Sample records for asme power test

  1. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    Energy Technology Data Exchange (ETDEWEB)

    Seniuk, P.J.

    1996-12-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

  2. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    International Nuclear Information System (INIS)

    Seniuk, P.J.

    1996-01-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, open-quotes Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plantsclose quotes. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O ampersand M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O ampersand M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion

  3. ASME power test code ptc 4.1 for steam generators; Codigo de pruebas de potencia ASME ptc 4.1 para generadores de vapor

    Energy Technology Data Exchange (ETDEWEB)

    Plauchu Alcantara, Jorge Alberto [Plauchu Consultores, Morelia, Michoacan (Mexico)

    2001-07-01

    This presentation is oriented towards those who in this subject have experience in the design and equipment specification, plant projects, factory and field testing, operation or result analyses. An important fraction of the national energy supply, approximately 13%, is applied to the steam generation in the different aspects of the industrial activity, in the electrical industry of public service and in the commercial and services sector. The development of the national programs of energy efficiency verifies this when dedicating to this use of the energy important projects, some of them with support of the USAID. The measurement of the energy utilization or the efficiency of steam generators (or boilers) is made applying some procedure agreed by the parts and the one of greater acceptance and best known in Mexico and internationally is the ASME Power Test Code PTC 4.1 for Steam Generators. The purpose and formality in the determination of efficiency and of steam generation capacity behavior, thermal basic regime or fulfillment of guarantees, radically changes the exigencies of strict attachment to the PTC 4.1 This definition will determine the importance of the test method selected, the deviations and convened exceptions, the influence of the precision and the measurement errors, the consideration of auxiliary equipment, etc. An interpretation or incorrect application of the Test Code has lead and will lead to results and nonreliable decisions. [Spanish] Esta exposicion se orienta a quienes en este tema cuenta con experiencia en diseno y especificacion de equipo, proyecto de planta, pruebas en fabrica y campo, operacion o analisis de resultados. Una fraccion importante de la oferta nacional de energia, 13% aproximadamente, se aplica a la generacion de vapor en diferentes giros de actividad industrial, en la industria electrica, de servicio publico y en el sector de servicios y comercial. El desarrollo de los programas nacionales de eficiencia energetica comprueba

  4. Nondestructive testing standards and the ASME code

    International Nuclear Information System (INIS)

    Spanner, J.C.

    1991-04-01

    Nondestructive testing (NDT) requirements and standards are an important part of the ASME Boiler and Pressure Vessel Code. In this paper, the evolution of these requirements and standards is reviewed in the context of the unique technical and legal stature of the ASME Code. The coherent and consistent manner by which the ASME Code rules are organized is described, and the interrelationship between the various ASME Code sections, the piping codes, and the ASTM Standards is discussed. Significant changes occurred in ASME Sections 5 and 11 during the 1980s, and these are highlighted along with projections and comments regarding future trends and changes in these important documents. 4 refs., 8 tabs

  5. ASME section XI: rules for inservice inspection of nuclear power plants -an introspection

    International Nuclear Information System (INIS)

    John, P.K.; Anto, Y.; Mungikar, C.P.; Wagh, P.M.

    1994-01-01

    Section XI of the ASME BPV code is addressed to the examination, test and inspection requirements of the components of nuclear power plants (NPPs). Since its inception in 1970, this code section has undergone vast changes -probably the most among other ASME BPV code sections. Section XI is full fledged and lays down requirements with regard to all preservice inspections, inservice inspection, repair and replacement of components, tests of system etc. Tarapur Atomic Power Station (TAPS) has the distinction of being one of the earliest BWR type NPPs in the world that has an inservice inspection programme organised in line with the ASME section XI requirements. This paper summarises the experiences gained from time to time using this code section and a few suggestions to prospective users. An effort is also made to explain the philosophy of inservice inspection from ASME section XI point of view. 3 refs

  6. ASME section XI: rules for inservice inspection of nuclear power plants -an introspection

    Energy Technology Data Exchange (ETDEWEB)

    John, P K; Anto, Y; Mungikar, C P; Wagh, P M [Nuclear Power Corporation of India Ltd., Tarapur (India). Tarapur Atomic Power Station

    1994-12-31

    Section XI of the ASME BPV code is addressed to the examination, test and inspection requirements of the components of nuclear power plants (NPPs). Since its inception in 1970, this code section has undergone vast changes -probably the most among other ASME BPV code sections. Section XI is full fledged and lays down requirements with regard to all preservice inspections, inservice inspection, repair and replacement of components, tests of system etc. Tarapur Atomic Power Station (TAPS) has the distinction of being one of the earliest BWR type NPPs in the world that has an inservice inspection programme organised in line with the ASME section XI requirements. This paper summarises the experiences gained from time to time using this code section and a few suggestions to prospective users. An effort is also made to explain the philosophy of inservice inspection from ASME section XI point of view. 3 refs.

  7. Parametric study of emerging high power accelerator applications using Accelerator Systems Model (ASM)

    International Nuclear Information System (INIS)

    Berwald, D.H.; Mendelsohn, S.S.; Myers, T.J.; Paulson, C.C.; Peacock, M.A.; Piaszczyk, CM.; Rathke, J.W.; Piechowiak, E.M.

    1996-01-01

    Emerging applications for high power rf linacs include fusion materials testing, generation of intense spallation neutrons for neutron physics and materials studies, production of nuclear materials and destruction of nuclear waste. Each requires the selection of an optimal configuration and operating parameters for its accelerator, rf power system and other supporting subsystems. Because of the high cost associated with these facilities, economic considerations become paramount, dictating a full evaluation of the electrical and rf performance, system reliability/availability, and capital, operating, and life cycle costs. The Accelerator Systems Model (ASM), expanded and modified by Northrop Grumman during 1993-96, provides a unique capability for detailed layout and evaluation of a wide variety of normal and superconducting accelerator and rf power configurations. This paper will discuss the current capabilities of ASM, including the available models and data base, and types of trade studies that can be performed for the above applications. (author)

  8. Technical requirements for the ASME PRA standard for nuclear power plant applications

    International Nuclear Information System (INIS)

    Fleming, Karl N.; Bernsen, Sidney A.; Simard, Ronald L.

    2000-01-01

    In 1998 the American Society of Mechanical Engineers (ASME) formed the Committee on Nuclear Risk Management (CNRM) and a Project Team to develop a standard on PRAs for use in risk informed applications. This ASME standard is being developed to help provide an adequate level of quality in PRAs that are being used to support ASME initiatives to risk informed in-service inspection (ISI) and in-service testing (IST) of nuclear power plant components. A related need supported by the industry and the U.S. Nuclear Regulatory Commission is to reduce the level of effort that is being expended in pilot applications of risk informed initiatives to address questions about the sufficiency of quality in the supporting PRA models. The purpose of this paper is to discuss the authors' views on some of the technical issues that were encountered in the effort to develop the ASME PRA standard. Draft 12 of this standard has been issued for comment, and is currently being finalized with the aim of releasing the standard in early 2001. (author)

  9. Interpretation, with respect to ASME code Case N-318, of limit moment and fatigue tests of lugs welded to pipe

    International Nuclear Information System (INIS)

    Foster, D.C.; Van Duyne, D.A.; Budlong, L.A.; Muffett, J.W.; Wais, E.A.; Streck, G.; Rodabaugh, E.C.

    1990-01-01

    Two nonmandatory ASME code cases have been used often in the evaluation of lugs on nuclear-power- plant piping systems. ASME Code Case N-318 provides guidance for evaluation of the design of rectangular cross-section attachments on Class 2 or 3 piping, and ASME Code Case N-122 provides guidance for evaluation of lugs on Class 1 piping. These code cases have been reviewed and evaluated based on available test data. The results indicate that the Code cases are overly conservative. Recommendations for revisions to the cases are presented which, if adopted, will reduce the overconservatism

  10. Risk-informed technology developments for nuclear power plants within the ASME in 2000-2001

    International Nuclear Information System (INIS)

    Wesley Rowley, C.; Balkey, K.R.

    2001-01-01

    The purpose of this paper is to provide information on developments within the ASME to support risk-informing NRC regulations for nuclear power plants. This paper builds on a publication at ICONE-8 that discussed ASME risk-informed nuclear power plant initiatives, both in Research and in Codes and Standards, particularly those related to risk-informing Part 50 of the 10 CFR (Code of federal regulations). During the past year, the ASME BNCS formed a Task Force to focus the Society's efforts to support risk-informing 10 CFR Part 50. Key efforts underway that are guided by the task force include finalizing the ASME PRA (probability risk assessment) Standard, developing a Code Case to risk-inform the repair, replacement, and modification activities for ASME components, and developing a Code Case to risk-inform the safety classification of pressure boundary components. Several other initiatives are also under investigation such as introducing risk insights into other ASME nuclear codes and standards supported by appropriate research and technical basis information. Supplementary information will also be provided to update an initial high level plan of ASME risk-informed initiatives for nuclear power plants that was presented at ICONE-8, including plans to communicate these risk-informed technology developments to the public. The authors included and acknowledged contributions from several other cognizant members of the ASME BNCS (board on nuclear codes standards) Task Group on RIP50 in the paper. (authors)

  11. Description of comprehensive pump test change to ASME OM code, subsection ISTB

    International Nuclear Information System (INIS)

    Hartley, R.S.

    1994-01-01

    The American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Main Committee and Board on Nuclear Codes and Standards (BNCS) recently approved changes to ASME OM Code-1990, Subsection ISTB, Inservice Testing of Pumps in Light-Water Reactor Power Plants. The changes will be included in the 1994 addenda to ISTB. The changes, designated as the comprehensive pump test, incorporate a new, improved philosophy for testing safety-related pumps in nuclear power plants. An important philosophical difference between the open-quotes old codeclose quotes inservice testing (IST) requirements and these changes is that the changes concentrate on less frequent, more meaningful testing while minimizing damaging and uninformative low-flow testing. The comprehensive pump test change establishes a more involved biannual test for all pumps and significantly reduces the rigor of the quarterly test for standby pumps. The increased rigor and cost of the biannual comprehensive tests are offset by the reduced cost of testing and potential damage to the standby pumps, which comprise a large portion of the safety-related pumps at most plants. This paper provides background on the pump testing requirements, discusses potential industry benefits of the change, describes the development of the comprehensive pump test, and gives examples and reasons for many of the specific changes. This paper also describes additional changes to ISTB that will be included in the 1994 addenda that are associated with, but not part of, the comprehensive pump test

  12. The application of RCM to ASME code requirements for in-service testing

    International Nuclear Information System (INIS)

    Rowley, C.W.

    1990-01-01

    This paper reports that the high reliability of nuclear power plant systems and components is highly important for both nuclear safety and electrical power production economics. The optimum operating performance of these plant systems and components is heavily dependent on the original or modified design for its inherent reliability and the appropriate trade-off in preventive and corrective maintenance for its developed reliability. In developing this optimum operating performance goal, the plant staff could rely solely on the experience of its personnel. However using this internal subjective approach, the average nuclear power availability has been far less than 80%. Obviously the production economics of a nuclear power plant is the province of the owner-operator, but the safety system and component performance impacts the entire industry. Hence the nuclear industry needs to have in-service testing requirements that maintain the necessary safety standards. Historically the ASME Inservice Testing Code has been a vehicle for defining some of those necessary safety standards, such as inservice testing of pumps, valves, and snubbers. The nuclear industry needs to expand the code testing to include all the systems that affect these necessary safety standards

  13. Design and test of ASME strainer for primary cooling system in HANARO

    International Nuclear Information System (INIS)

    Park, Yong-Chul; Ryu, Jeong-Soo

    1999-01-01

    The ASME strainers have been newly installed at the suction side of each reactor coolant pump to get rid of the foreign materials which may damage the pump impeller or interfere with the coolant path of fuel flow tube or primary plate type heat exchanger. The strainer was designed in accordance with ASME SEC. III, DIV. 1, ND and the structural integrity was verified by seismic analysis. The screen was designed in accordance with the effective void area from the result of flow analysis for T-type strainer. After installation of the strainer, it was confirmed through the field test that the flow characteristics of primary cooling system were not adversely affected. The pressure loss coefficient was calculated by Darcy equation using the pressure difference through each strainer and the flow rate measured during the strainer performance test. And these are useful data to predict flow variations by the pressure difference. (author)

  14. ASME N510 test results for Savannah River Site AACS filter compartments

    Energy Technology Data Exchange (ETDEWEB)

    Paul, J.D.; Punch, T.M. [Westinghouse Savannah River Company, Aiken, SC (United States)

    1995-02-01

    The K-Reactor at the Savannah River Site recently implemented design improvements for the Airborne Activity Confinement System (AACS) by procuring, installing, and testing new Air Cleaning Units, or filter compartments, to ASME AG-11, N509, and N510 requirements. Specifically, these new units provide documentable seismic resistance to a Design Basis Accident earthquake, provide 2 inch adsorber beds with 0.25 second residence time, and meet all AG-1, N509, and N510 requirements for testability and maintainability. This paper presents the results of the Site acceptance testing and discusses an issue associated with sample manifold qualification testing.

  15. Code cases for implementing risk-based inservice testing in the ASME OM code

    Energy Technology Data Exchange (ETDEWEB)

    Rowley, C.W.

    1996-12-01

    Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices.

  16. Code cases for implementing risk-based inservice testing in the ASME OM code

    International Nuclear Information System (INIS)

    Rowley, C.W.

    1996-01-01

    Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices

  17. Activated sludge models ASM1, ASM2, ASM2d and ASM3

    DEFF Research Database (Denmark)

    Henze, Mogens; Gujer, W.; Mino, T.

    This book has been produced to give a total overview of the Activated Sludge Model (ASM) family at the start of 2000 and to give the reader easy access to the different models in their original versions. It thus presents ASM1, ASM2, ASM2d and ASM3 together for the first time.Modelling of activated...... sludge processes has become a common part of the design and operation of wastewater treatment plants. Today models are being used in design, control, teaching and research.ContentsASM3: Introduction, Comparison of ASM1 and ASM3, ASM3: Definition of compounds in the model, ASM3: Definition of processes...... in the Model, ASM3: Stoichiometry, ASM3: Kinetics, Limitations of ASM3, Aspects of application of ASM3, ASM3C: A Carbon based model, Conclusion ASM 2d: Introduction, Conceptual Approach, ASM 2d, Typical Wastewater Characteristics and Kinetic and Stoichiometric Constants, Limitations, Conclusion ASM 2...

  18. Report on ANSI/ASME nuclear air and gas treatment standards for nuclear power plants

    International Nuclear Information System (INIS)

    Fish, J.F.

    1979-01-01

    Original N Committee, N45-8, has completed and published through the approved American National Standards Institute process two Standards, N-509 and N-510. This committee has been dissolved and replaced by ASME Committee on Nuclear Air and Gas Treatment with expanded scope to cover not only air cleaning, but thermal treatment equipment. Current efforts are directed to produce Code documents rather than Standards type publications. This report summarizes changed scope, current organization and sub-committee coverage areas

  19. Proceedings of the 4th NRC/ASME symposium on valve and pump testing

    International Nuclear Information System (INIS)

    1996-01-01

    The 1996 Symposium on Valve and Pump Testing, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the U.S. Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. Individual papers of this Proceedings have been cataloged separately

  20. Proceedings of the 4th NRC/ASME symposium on valve and pump testing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The 1996 Symposium on Valve and Pump Testing, jointly sponsored by the Board on Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the U.S. Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. Individual papers of this Proceedings have been cataloged separately.

  1. More on fatigue verification of Class 1 nuclear power piping according to ASME BPV III NB-3600

    International Nuclear Information System (INIS)

    Zeng, Lingfu; Dahlström, Lars; Jansson, Lennart G.

    2011-01-01

    In this paper, fatigue verification of Class 1 nuclear power piping according to ASME Boiler and Pressure Vessel Code, Section III, NB-3600, and relevant issues that are often discussed in connection to the power uprate of several Swedish BWR reactors in recent years, are dealt with. Key parameters involved in the fatigue verification, i.e. the alternating stress intensity S alt , the penalty factor K e and the cumulative damage factor U, and relevant computational procedures applicable for the assessment of low-cycle fatigue failure using strain-controlled data, are particularly addressed. A so-called simplified elastic-plastic discontinuity analysis for alternative verification when basic fatigue requirements found unsatisfactory, and the procedures provided in NB-3600 for evaluating the alternating stress intensity S alt , are reviewed in detail. Our emphasis is placed on other procedures alternative to the simplified elastic-plastic discontinuity analysis. A more in-depth discussion is given to an alternative suggested earlier by the authors using nonlinear finite element analyses. This paper is a continuation of our work presented in ICONE16/17/18, which attempted to categorize design rules in the code into linear design rules and non-linear design rules and to clarify corresponding design requirements and finite element analyses, in particular, those non-linear ones. (author)

  2. Cisco ASM Router

    CERN Multimedia

    2001-01-01

    One of the two "ASM/2-32EM" boxes installed in 1988, from "Cisco Systems Inc." - then an unknown 20-employee company in Menlo Park, California (USA). This is one of the first two Cisco boxes to appear in Switzerland, and possibly Europe. The 220v power supply was a special modification made for use at CERN. They supported IP address filtering, which seemed just what CERN needed to help protect the new Cray XMP-48 super computer from network hackers. The two ASM boxes were both routers and terminal servers. They protected a secure private Ethernet segment used by the Cray project, as well as providing secure terminal connections to that segment, including CERN's first dialback terminal service, which allowed Cray and CERN system analysts to work on the machine from home, using another Cisco feature called TACACS. (Kindly offered by B. Segal who discovered this company while at a Usenix Conference in Phoenix, Arizona in June 1987.)

  3. ASME development of risk-based inspection guidelines for nuclear power plants

    International Nuclear Information System (INIS)

    Gore, B.F.; Balkey, K.R.

    1992-08-01

    A methodology has been developed for use in preparing guidelines for the in-service inspection of nuclear power plant pressure boundary and structural components. This methodology is a further development of a general methodology previously published by the task force for application to any industry. It is unique in that it utilizes probabilistic risk assessment (PRA) information to improve the quantification of risks associated with component ruptures. A procedure has also been recommended for using the resulting quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Characteristics which an inspection strategy must possess in order to maintain target rupture probabilities are being determined by structural risk and reliability analysis (SRRA) calculations. The major features of this methodology are described in this paper, and example applications are briefly discussed. Plans for additional work in developing and applying the methodology are described

  4. Status of ASME risk-based inspection guidelines development for nuclear power plants

    International Nuclear Information System (INIS)

    Gore, B.F.; Vo, T.V.; Balkey, K.R.

    1991-01-01

    The general methodology published by the task force is being applied to develop guidelines for the inspection of nuclear power plant components. The methodology has been expanded to include the use of information from probabilistic risk assessments to improve the quantification of risks associated with component ruptures. In addition, a procedure has been recommended for using these quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Finally, the method for determining the characteristics which an inspection strategy must possess in order to maintain these target rupture probabilities has been described. This paper presents the major features of this methodology, along with a discussion of the status of efforts to develop and to apply it further

  5. Scope and implementation of standards ASME N510 / N511 in air treatment system (HVAC) of the Asco nuclear power plant; Alcance e implementacion de las normas ASME N511 en el sistema de tratamiento de aire (HVAC) de la central nuclear de Asco

    Energy Technology Data Exchange (ETDEWEB)

    Jaimot Jimenez, J. J.

    2013-07-01

    With the ITC for renewal of license units 1 and 2 of Asco, the CSN It required the commissioning tests underway in the air, according to ASME N510 filter units. It is required that, for safety-related units, to undertake preventive inspections according to ASME N511. All these requirements, in tight deadlines, have represented a great challenge for the organizations of maintenance and engineering of ANAV.

  6. Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

    Energy Technology Data Exchange (ETDEWEB)

    Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P. [and others

    1996-12-01

    This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising.

  7. Optimized periodic verification testing blended risk and performance-based MOV inservice test program an application of ASME code case OMN-1

    International Nuclear Information System (INIS)

    Sellers, C.; Fleming, K.; Bidwell, D.; Forbes, P.

    1996-01-01

    This paper presents an application of ASME Code Case OMN-1 to the GL 89-10 Program at the South Texas Project Electric Generating Station (STPEGS). Code Case OMN-1 provides guidance for a performance-based MOV inservice test program that can be used for periodic verification testing and allows consideration of risk insights. Blended probabilistic and deterministic evaluation techniques were used to establish inservice test strategies including both test methods and test frequency. Described in the paper are the methods and criteria for establishing MOV safety significance based on the STPEGS probabilistic safety assessment, deterministic considerations of MOV performance characteristics and performance margins, the expert panel evaluation process, and the development of inservice test strategies. Test strategies include a mix of dynamic and static testing as well as MOV exercising

  8. Relaxation of inservice test frequency requirement for Kori 1 ASME code pumps

    International Nuclear Information System (INIS)

    Sohn, Gap Heon; Choi, Hae Yoon; Min, Kyung Sung; Rim, Nam Jin

    1994-08-01

    The objective of this investigation is to evaluate the technical and regulational requirements to justify the relaxation of the test frequency of Kori 1 pumps through reviewing the related rules and codes and standards, technical specifications of Kori 1 and other similar plants, standard technical specifications, research results for tech. spec. improvements and site test records. It is concluded that the relaxation of test frequency to quarterly be justified based on the conformance with rules and codes and standard, quarterly test cases in similar plants and standard tech. spec., recommendations of research result and stable site test records. (Author) 16 refs., 26 figs., 13 tabs

  9. ASME XI stroke time testing of solenoid valves at Connecticut Yankee Station

    Energy Technology Data Exchange (ETDEWEB)

    Martin, C.W.

    1996-12-01

    Connecticut Yankee Atomic Power Company has developed the capability of measuring the stroke times of AC and DC solenoid valves. This allows the station to measure the stroke time of any solenoid valve in the plant, even those valves which do not have valve stem position indicators. Connecticut Yankee has adapted the ITI MOVATS Checkmate 3 system, using a signal input from a Bruel and Kjaer (B&K) Model 4382 acoustic accelerometer and the Schaumberg Campbell Associates (SCA) Model SCA-1148 dual sensor, which is a combined accelerometer and gaussmeter.

  10. ASME XI stroke time testing of solenoid valves at Connecticut Yankee Station

    International Nuclear Information System (INIS)

    Martin, C.W.

    1996-01-01

    Connecticut Yankee Atomic Power Company has developed the capability of measuring the stroke times of AC and DC solenoid valves. This allows the station to measure the stroke time of any solenoid valve in the plant, even those valves which do not have valve stem position indicators. Connecticut Yankee has adapted the ITI MOVATS Checkmate 3 system, using a signal input from a Bruel and Kjaer (B ampersand K) Model 4382 acoustic accelerometer and the Schaumberg Campbell Associates (SCA) Model SCA-1148 dual sensor, which is a combined accelerometer and gaussmeter

  11. 9 July 2012 - Academy of Sciences Malaysia (ASM), Chairman, Mathematical and Physical Sciences Discipline Group M. Yahaya FASc and his delegation visiting the LHC superconducting magnet test hall with Technology Department G. De Rijk.

    CERN Multimedia

    Maximilien Brice

    2012-01-01

    9 July 2012 - Academy of Sciences Malaysia (ASM), Chairman, Mathematical and Physical Sciences Discipline Group M. Yahaya FASc and his delegation visiting the LHC superconducting magnet test hall with Technology Department G. De Rijk.

  12. Relationship between IEEE Std. 7-4.3.2-1993 and ASME NQA-1, parts I and II revisions and the impact on nuclear power generating stations

    International Nuclear Information System (INIS)

    Blauw, R.J.

    1996-01-01

    Clear understanding of software related design control requirements is key to growth in the use of computers in nuclear power generating stations. Inconsistent terminology within the nuclear and software standards arena has impacted the ability of both nuclear station system engineers (i.e., the domain expert) to clearly communicate with the software/computer hardware experts. In order for computer development to occur both groups need to have a common terminology basis. Without this commonality, inappropriate application of requirements could result. This paper will present a overview of ongoing efforts within the Institute of Electrical and Electronics Engineers Nuclear Power Engineering Committee (IEEE NPEC) and the American Society of Mechanical Engineers Nuclear Quality Assurance (ASME NQA) Committee to develop this commonality

  13. A practical approach for implementing risk-based inservice testing of pumps at nuclear power plants

    International Nuclear Information System (INIS)

    Hartley, R.S.; Maret, D.; Seniuk, P.; Smith, L.

    1996-01-01

    The American Society of Mechanical Engineers (ASME) Center for Research and Technology Development's (CRTD) Research Task Force on Risk-Based Inservice Testing has developed guidelines for risk-based inservice testing (IST) of pumps and valves. These guidelines are intended to help the ASME Operation and Maintenance (OM) Committee to enhance plant safety while focussing appropriate testing resources on critical components. This paper describes a practical approach for implementing those guidelines for pumps at nuclear power plants. The approach, as described in this paper, relies on input, direction, and assistance from several entities such as the ASME Code Committees, United States Nuclear Regulatory Commission (NRC), and the National Laboratories, as well as industry groups and personnel with applicable expertise. Key parts of the risk-based IST process that are addressed here include: identification of important failure modes, identification of significant failure causes, assessing the effectiveness of testing and maintenance activities, development of alternative testing and maintenance strategies, and assessing the effectiveness of alternative testing strategies with present ASME Code requirements. Finally, the paper suggests a method of implementing this process into the ASME OM Code for pump testing

  14. A practical approach for implementing risk-based inservice testing of pumps at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hartley, R.S. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Maret, D.; Seniuk, P.; Smith, L.

    1996-12-01

    The American Society of Mechanical Engineers (ASME) Center for Research and Technology Development`s (CRTD) Research Task Force on Risk-Based Inservice Testing has developed guidelines for risk-based inservice testing (IST) of pumps and valves. These guidelines are intended to help the ASME Operation and Maintenance (OM) Committee to enhance plant safety while focussing appropriate testing resources on critical components. This paper describes a practical approach for implementing those guidelines for pumps at nuclear power plants. The approach, as described in this paper, relies on input, direction, and assistance from several entities such as the ASME Code Committees, United States Nuclear Regulatory Commission (NRC), and the National Laboratories, as well as industry groups and personnel with applicable expertise. Key parts of the risk-based IST process that are addressed here include: identification of important failure modes, identification of significant failure causes, assessing the effectiveness of testing and maintenance activities, development of alternative testing and maintenance strategies, and assessing the effectiveness of alternative testing strategies with present ASME Code requirements. Finally, the paper suggests a method of implementing this process into the ASME OM Code for pump testing.

  15. 76 FR 36231 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases

    Science.gov (United States)

    2011-06-21

    ...The NRC is amending its regulations to incorporate by reference the 2005 Addenda (July 1, 2005) and 2006 Addenda (July 1, 2006) to the 2004 ASME Boiler and Pressure Vessel Code, Section III, Division 1; 2007 ASME Boiler and Pressure Vessel Code, Section III, Division 1, 2007 Edition (July 1, 2007), with 2008a Addenda (July 1, 2008); 2005 Addenda (July 1, 2005) and 2006 Addenda (July 1, 2006) to the 2004 ASME Boiler and Pressure Vessel Code, Section XI, Division 1; 2007 ASME Boiler and Pressure Vessel Code, Section XI, Division 1, 2007 Edition (July 1, 2007), with 2008a Addenda (July 1, 2008); and 2005 Addenda, ASME OMa Code-2005 (approved July 8, 2005) and 2006 Addenda, ASME OMb Code-2006 (approved July 6, 2006) to the 2004 ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The NRC is also incorporating by reference (with conditions on their use) ASME Boiler and Pressure Vessel Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials, Section XI, Division 1,'' Supplement 8, ASME approval date: January 26, 2009, and ASME Boiler and Pressure Vessel Code Case N-770-1, ``Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Section XI, Division 1,'' ASME approval date: December 25, 2009.

  16. Application procedures and analysis examples of the SIE ASME-NH program

    International Nuclear Information System (INIS)

    Kim, Seok Hoon; Koo, G. H.; Kim, J. B.

    2010-12-01

    In this report, the design rule of the ASME-NH Code was briefly summarized and the application procedures of SIE ASME-NH program were analysed, the analysis examples were described. The SIE ASME-NH program was developed according to the ASME Code Section III Subsection NH rules to perform the primary stress limits, the accumulated inelastic strain limits and the creep fatigue damage evaluations in the structural design of nuclear power plants operating with high temperatures over creep temperature at normal operating conditions. In the analysis examples, the benchmark problem for the high temperature reactor vessel which was discussed in the SIE ASME-NH user's seminar was described. Also, the preliminary structural analysis of an Advanced Burner Test Reactor internal structure was described. Considering the load combinations of the various cycle types submitted from significant operating conditions, the integrity of a reactor internal structure was reviewed according to the stress and strain limits of the ASME-NH rules and the analysis and evaluation results were summarized

  17. The ASME Code today -- Challenges, threats, opportunities

    International Nuclear Information System (INIS)

    Canonico, D.A.

    1995-01-01

    Since its modest beginning as a single volume in 1914 the ASME Code, or some of its parts, is recognized today in 48 of the United States and all providence's of Canada. The ASME Code today is composed of 25 books including two Code Case books. These books cover the new construction of boilers and pressure vessels and the new construction and In-Service-Inspection of Nuclear Power Plant components. The ASME accredits all manufacturers of boilers and pressure vessels built to the ASME Code. There are approximately 7650 symbol stamps issued throughout the world. Over 23% of the symbol stamps have been issued outside the USA and Canada. The challenge to the ASME Code is to be accepted as the world standard for pressure boundary components. There are activities underway to achieve that goal. The ASME Code is being revised to make it a more friendly document to entities outside of North America. To achieve that end there are specific tasks underway which are described here

  18. On-plant test of TUV HCM12 and ASME T23 alloys for use as waterwall materials

    DEFF Research Database (Denmark)

    Karlsson, Asger; Rasmussen, Frands; Montgomery, Melanie

    2000-01-01

    to the stress' which appear in the water walls of an USC plant. The exposure was performed in the stress region close to the yield strength limit and cracks appeared in one of the materials. The tests have revealed that HCM12 is difficult to weld but develops no thermal fatigue crack whereas HCM2S is easy......The development of Ultra Super Critical power plants has emphasised the need for stronger alloys for use as water walls. For this purpose HCM12 and HCM2S have been tested at 2 Danish power plants. HCM12 and HCM2S have been tested for their weldability, for thermal fatigue properties...... and for their oxidation properties under steam conditions 435-520º C, 26-27 MPa. The test included application of different weld methods and testing of the mechanical properties of the weld seams. The weld parameters giving the optimal properties were chosen to manufacture two pieces of water wall (1 times 1.5 meter...

  19. Scheduling the powering tests

    CERN Document Server

    Barbero-Soto, E; Casas-Lino, M P; Fernandez-Robles, C; Foraz, K; Pojer, M; Saban, R; Schmidt, R; Solfaroli-Camillocci, M; Vergara-Fernandez, A

    2008-01-01

    The Large Hadron Collider is now entering in its final phase before receiving beam, and the activities at CERN between 2007 and 2008 have shifted from installation work to the commissioning of the technical systems ("hardware commissioning"). Due to the unprecedented complexity of this machine, all the systems are or will be tested as far as possible before the cool-down starts. Systems are firstly tested individually before being globally tested together. The architecture of LHC, which is partitioned into eight cryogenically and electrically independent sectors, allows the commissioning on a sector by sector basis. When a sector reaches nominal cryogenic conditions, commissioning of the magnet powering system to nominal current for all magnets can be performed. This paper briefly describes the different activities to be performed during the powering tests of the superconducting magnet system and presents the scheduling issues raised by co-activities as well as the management of resources.

  20. Evaluation of the ICET Test Stand to Assess the Performance of a Range of Ceramic Media Filter Elements in Support of ASME AG-1 Subsection FO

    Energy Technology Data Exchange (ETDEWEB)

    Schemmel, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-04-26

    High Efficiency Particulate Air (HEPA) filters are defined as extended-medium, dry-type filters with: (1) a minimum particle removal efficiency of no less than 99.97 percent for 0.3 micrometer particles, (2) a maximum, clean resistance of 1.0 inch water column (in. WC) when operated at 1,000 cubic feet per minute (CFM), and (3) a rigid casing that extends the full depth of the medium. Specifically, ceramic media HEPA filters provide better performance at elevated temperatures, are moisture resistant and nonflammable, can perform their function if wetted and exposed to greater pressures, and can be cleaned and reused. This paper describes the modification and design of a large scale test stand which properly evaluates the filtration characteristics of a range of ceramic media filters challenged with a nuclear aerosol agent in order to develop Section FO of ASME AG-1.

  1. Safety Analysis Report for the KRI-ASM Transport Package

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Kim, D. H.; Park, H. Y.; Kim, J. B.; Kim, H. J.; Seo, K. S

    2005-11-15

    Safety evaluation for the KRI-ASM transport package to transport safely I-131, which is produced at HANARO research reactor in KAERI, was carried out. In the safety analyses results for the KRI-ASM transport package, all the maximum stresses as well as the maximum temperature of the surface are lower than their allowable limits. The safety tests were performed by using the test model of the KRI-ASM transport package. Leak Test was performed after drop test and penetration test, the measured leakage rate was lower than allowable leakage rate. It is revealed that the containment integrity of the KRI-ASM transport package is maintained. Therefore, it shows that the integrity of the KRI-ASM transport package is well maintained.

  2. Statistical analysis of the ASME KIc database

    International Nuclear Information System (INIS)

    Sokolov, M.A.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) K Ic curve is a function of test temperature (T) normalized to a reference nil-ductility temperature, RT NDT , namely, T-RT NDT . It was constructed as the lower boundary to the available K Ic database. Being a lower bound to the unique but limited database, the ASME K Ic curve concept does not discuss probability matters. However, a continuing evolution of fracture mechanics advances has led to employment of the Weibull distribution function to model the scatter of fracture toughness values in the transition range. The Weibull statistic/master curve approach was applied to analyze the current ASME K Ic database. It is shown that the Weibull distribution function models the scatter in K Ic data from different materials very well, while the temperature dependence is described by the master curve. Probabilistic-based tolerance-bound curves are suggested to describe lower-bound K Ic values

  3. Round Robin Test for Performance Demonstration System of Ultrasound Examination Personnel in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Young Ho; Yang, Seung Han; Kim, Yong Sik; Yoon, Byung Sik; Lee, Hee Jong

    2005-01-01

    Ultrasound testing performance during in-service inspection for the main components of NPPs is strongly affected by each examination person. Therefore, ASME established a more strict qualification requirement in Sec. XI Appendix VIII for the ultrasound testing personnel in nuclear power plants. The Korean Performance Demonstration (KPD) System according to the ASME code for the ultrasonic testing personnel, equipments, and procedures to apply to the Class 1 and 2 piping ultrasound examination of nuclear power plants in Korea was established. And a round robin test was conducted in order to verify the effectiveness of PD method by comparing the examination results from the method of Performance Demonstration (PD) and a traditional ASME code dB-drop method. The round robin test shows that the reliability of the PD method is better than that of the dB-drop method. As a result, application of the PD method to the in-service inspection of the nuclear power plants will improve the performance of ultrasound testing

  4. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    International Nuclear Information System (INIS)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

    2016-01-01

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

  5. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  6. 75 FR 24323 - American Society of Mechanical Engineers (ASME) Codes and New and Revised ASME Code Cases

    Science.gov (United States)

    2010-05-04

    ...The NRC proposes to amend its regulations to incorporate by reference the 2005 Addenda through 2008 Addenda of Section III, Division 1, and the 2005 Addenda through 2008 Addenda of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code (ASME B&PV Code); and the 2005 Addenda and 2006 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). The NRC also proposes to incorporate by reference ASME Code Case N-722-1, ``Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials Section XI, Division 1,'' and Code Case N-770, ``Alternative Examination Requirements and Acceptance Standards for Class 1 PWR [Pressurized- Water Reactor] Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material with or without Application of Listed Mitigation Activities.''

  7. AIS ASM Operational Integration Plan

    Science.gov (United States)

    2013-08-01

    Rack mount computer AIS Radio Interface Ethernet Switch 192.168.0.x Firewall Cable Modem 192.168.0.1 VTS Accred. Boundary AIS ASM Operational... AIS ASM Operational Integration Plan Distribution Statement A: Approved for public release; distribution is unlimited. August 2013 Report No...CD-D-07-15 AIS ASM Operational Integration Plan ii UNCLAS//Public | CG-926 R&DC | I. Gonin, et al. | Public August 2013 N O T I C

  8. Meeting the difficulties of an ASME calibration for pipe welds

    International Nuclear Information System (INIS)

    Ginzel, E.; Ginzel, R.; Buchholz, J.

    2013-01-01

    The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code Section V deals with Non-Destructive Testing (NDT). In North America the use of ASME Sec. V is not always limited to just ASME regulated projects. Even such seemingly unrelated Standards as CSA Z662 (the Canadian construction standard for Oil and Gas Pipeline Systems) references this document for some aspects of the NDT. ASME has several codes that are dedicated to piping welds, for example ASME 31.1, 31.3 and 31.8. All of them reference back to ASME Section V Article 4 for the accepted techniques to use when UT is the examination option, but sometimes ASME is not very helpful. As they try to leave options open in one area they close the doors in others, or so it seems when reading the clauses as mandatory where you see the word 'shall'. This paper will attempt to layout the issues with pipe weld inspections as per Article 4, 2011 edition. Like all other codes and standards, ASME is a regulated standard that is regularly reviewed and updated, and one must be careful to reference the year of the edition being used. For the purposes of this paper we have used the 2010/11 edition. (author)

  9. Meeting the difficulties of an ASME calibration for pipe welds

    Energy Technology Data Exchange (ETDEWEB)

    Ginzel, E., E-mail: eginzel@mri.on.ca [Materials Research Inst., Waterloo, Ontario (Canada); Ginzel, R.; Buchholz, J., E-mail: rginzel@eclipsescientific.com [Eclipse Scientific, Waterloo, Ontario (Canada)

    2013-11-15

    The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code Section V deals with Non-Destructive Testing (NDT). In North America the use of ASME Sec. V is not always limited to just ASME regulated projects. Even such seemingly unrelated Standards as CSA Z662 (the Canadian construction standard for Oil and Gas Pipeline Systems) references this document for some aspects of the NDT. ASME has several codes that are dedicated to piping welds, for example ASME 31.1, 31.3 and 31.8. All of them reference back to ASME Section V Article 4 for the accepted techniques to use when UT is the examination option, but sometimes ASME is not very helpful. As they try to leave options open in one area they close the doors in others, or so it seems when reading the clauses as mandatory where you see the word 'shall'. This paper will attempt to layout the issues with pipe weld inspections as per Article 4, 2011 edition. Like all other codes and standards, ASME is a regulated standard that is regularly reviewed and updated, and one must be careful to reference the year of the edition being used. For the purposes of this paper we have used the 2010/11 edition. (author)

  10. Review of design analysis and site installation method of ASME vessel of fuel test loop of Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung Won; Kim, Jun Yeon

    1997-02-01

    The major goal of this report is to take the advantages under control, and provide basic solutions for field installation. In order to access to technical availability, the scope, limitation, and possibilities of design requirements are carefully considered in this. The chapter 3 deals with seismic stress analysis of vessels, using manufacturers` finite element analysis data. The test of manufactured equipment is scheduled to hold near future. The evaluation criteria and inspection specifications for the quality release are well illustrated in the chapter IV to efficiently check out the items. The full considerations for field installation are also reviewed with that in terms of space limit. Following the inspection and test of manufacturer`s shop, the results will be promptly reported. Based on installation principles and the analysis in the report, updated procedure and methodology for the installation will be applied to the field in more details and better breakdown precision in the coming years. (author). 14 refs., 23 tabs., 26 figs.

  11. Review of design analysis and site installation method of ASME vessel of fuel test loop of Hanaro

    International Nuclear Information System (INIS)

    Cho, Sung Won; Kim, Jun Yeon.

    1997-02-01

    The major goal of this report is to take the advantages under control, and provide basic solutions for field installation. In order to access to technical availability, the scope, limitation, and possibilities of design requirements are carefully considered in this. The chapter 3 deals with seismic stress analysis of vessels, using manufacturers' finite element analysis data. The test of manufactured equipment is scheduled to hold near future. The evaluation criteria and inspection specifications for the quality release are well illustrated in the chapter IV to efficiently check out the items. The full considerations for field installation are also reviewed with that in terms of space limit. Following the inspection and test of manufacturer's shop, the results will be promptly reported. Based on installation principles and the analysis in the report, updated procedure and methodology for the installation will be applied to the field in more details and better breakdown precision in the coming years. (author). 14 refs., 23 tabs., 26 figs

  12. Technical justification for ASME code section xi crack detection by visual examination

    International Nuclear Information System (INIS)

    Nickell, R.E.; Rashid, Y.R.

    2001-01-01

    A critical technical element of nuclear power plant license renewal in the United States is the demonstration that the effects of aging do not compromise the intended safety function(s) of a system, structure, or component during the extended term of operation. The demonstration may take either of two forms. First, it can be shown that the design basis for the system, structure, or component is sufficiently robust that the aging effects have been insignificant through the current license term, and will continue to be insignificant through the extended term. Alternatively, it can be shown that, while the aging effects may be potentially significant, those effects can be managed and functionality maintained by defined programmatic activities during the extended term of operation. The first of the two approaches is generally provided by the construction basis, such as construction in accordance with the ASME Code Section III and other consensus codes and standards. The second of the two approaches is often provided by periodic inservice inspection and testing, in accordance with the ASME Code Section XI. The purpose of the ASME Section XI inspections and tests is to assure that systems, components, and structures are fit for continued service until the next scheduled inspection or test. The purpose of this paper is to document the effectiveness of the current ASME Code Section XI visual examination procedures in detecting the effects of aging for systems, structures, and components that are tolerant of mature cracks. (author)

  13. ASME factory authorization system and the situation in Japan

    International Nuclear Information System (INIS)

    Futagawa, Kiyoshi

    1978-01-01

    Since about three or four years ago, the enterprises of machinery, iron and steel and welding materials in Japan are paying much attention to the acquisition of ASME (American Society of Mechanical Engineers) certificates or authorization to stamp the code symbols. That is, over 70 factories in Japan have undergone ASME examination, and consequently acquired the authorization or certificates. Such authorization is divided into over 20 kinds, of which about 7 are possessed by the companies in Japan. In nuclear field, the kinds of authorization are N (nuclear vessel), NPT (nuclear vessel parts), NV (nuclear vessel safety valve), and MM (material manufacturing). In non-nuclear fields, they are S (power boilers), U (pressure vessels, in Div. 1), and U2 (pressure vessels in Div. 2). The following matters are described: ASME setup, authorization procedures of ASME for factories, the kinds of authorization, factories in Japan holding the authorization or certificates, and renewal of the authorization. (Mori, K.)

  14. Assessment of induction elbows for an ASME Code application

    International Nuclear Information System (INIS)

    Panesar, J.S.; Soliman, M.

    1991-01-01

    The ASME Nuclear Codes impose some specific requirements on the wall thickness uniformity and the out-of-roundness of cross sections of the elbows for Nuclear Power Plant applications. Due to some of these requirements, manufacturing and installation of these elbows can be time consuming and quite expensive. This paper explores the feasibility of using induction elbows for nuclear application from the stress analysis point of view. To this end, three different sizes of 90deg elbows have been analyzed based on the geometry of an 'ASME Code' elbow and an elbow formed by induction bending. The analysis is carried out for internal pressure, in-plane and out-of-plane loads. Based on the results of these three carbon steel elbows, the use of induction elbows in some of the CANDU-PHW (CANadian Deuterium Uranium-Pressurized Heavy Water) power plant applications seems encouraging. However, before the feasibility can be fully confirmed analysis and induction bending tests over a wider range of geometries, loading conditions, and materials are required. (author)

  15. Techniques of the snubber testing in a nuclear power plant

    International Nuclear Information System (INIS)

    Hong, Soon Sin; Kim, Byung Chul

    1994-05-01

    The performance of PSI/ISI is one of the important inspections which improve the safety, reliability and operability of nuclear power plant. Therefore, KAERI has performed the inspections for the safety-related piping systems, pressure vessel and support components of NPP by the NDT methods in accordance with the related technical codes and standards. ASME Sec.XI of 1989 requires to perform the visual inspection and functional test of snubbers in accordance with ASME/ANSI OM-Part 4 additionally. Especially, the inspection and functional test of heavy duty hydraulic snubbers attached to the steam generator and RC pump have also been added. The main role of snubbers protects the piping system and pressure vessel from the transient events such as water hammer and earthquake during operation. Throughout the analysis of the status of snubbers in NPP, the snubber failure cases in domestic and foreign countries, the requirements and technical standards of the inspection and test, problems of snubbers in NPP, and snubber reduction program, this report requires to follow the technical codes and standards for the inspection and functional test of snubbers now being insufficient in Korean NPP. As a following action, the safety verification analysis for the reliability of piping systems and components of NPP due to snubber failure causes will be studied, and also the snubber reduction program of NPP will simplify the control and maintenance of the expensive snubbers. The objective of this report is to provide the problems of snubber in Korean NPP and contribute to the integrity assurance of the piping systems and components by the inspection and test of snubbers. 17 figs., 9 tabs., 19 refs. (Author)

  16. Inelasticity and the ASME code

    International Nuclear Information System (INIS)

    Berman, I.

    1978-01-01

    Although it may have more general applicability, this paper is specifically concerned with some aspects of plasticity for class I nuclear components that are contained in section III of the ASME Boiler and Pressure Vessel Code. It directly addresses design for components at temperatures at which creep is not a factor. A review is made of the relationship of plasticity to each of the three failure modes that the stress limits are intended to prevent. It is found that the prevention of bursting and gross distortion from a single application of pressure and the prevention of fatigue failure caused by repeated cycles of peak stresses are well supported by experimental results. The experimental verification for the rules to show that the primary plus secondary stresses shakedown to elastic behavior is not clear. Various directed efforts which could lead to greater assurance of shakedown to elastic behavior are suggested. The major approach should be a massive program to develop a test matrix of experimental information for various portions of each component of interest in the Code. (Auth.)

  17. The current state of inservice testing programs at U.S. Nuclear Power Plants - a regulatory overview

    International Nuclear Information System (INIS)

    Campbell, P.; Colaccino, J.

    1994-01-01

    Information is provided on inservice testing (IST) of pumps and valves at U.S. nuclear power plants to provide consistency in the implementation of regulatory requirements and to enhance communications among utility licensees who may have, like NSSS vendors, similar kinds and numbers of components or comparable IST programs. Documents discussed include the ASME Operation and Maintenance Standards Parts 6 and 10 (covering inservice testing of pumps and valves in light water reactor power plants), the draft NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants (including review comments by Nuclear Management and Resource Council), and applicable Licensee Event Reports including summaries of several reports relating to IST

  18. Application of ASME code AG-1 to YGN 3 ampersand 4 plants, South Korea

    International Nuclear Information System (INIS)

    Kim, Y.K.; Porco, R.D.; York, Y.D.

    1993-01-01

    Yonggwang Nuclear Power Plant Units 3 ampersand 4 are located on the southwestern coast of South Korea on the Yellow Sea. The plant is owned by Korea Electric Power Corp. (KEPCO), with the engineering being performed by Korea Power Engineering Co., Inc. (KOPEC) and Sargent and Lundy under a technology transfer agreement. The plants are both 950 Megawatt (electric) pressurized water reactors of US design. Under contract to KEPCO, Korea Heavy Industries and Construction Co., Ltd. and Ellis and Watts, Division of Dynamics Corporation of America, Batavia, Ohio, supplied major components to the YGN plants in compliance to ASME AG-1. These components included safety related Air Cleaning Units, Reactor containment Fan Cooler Units, Air Handling Units, Cubicle Coolers, Duct Electric Heaters, and fans. This paper details the extent of applicability of ASME Code AG-1 to the specific equipment, description of the equipment, conformance, testing, and design required. The paper also discusses the problems encountered in implementing ASME AG-1, working around Code sections that were not complete at contract inception, conflicts in project documents and related problems. Also discussed are the logistics problems, material availability, and quality assurance aspects complicating the applications of ASME AG-1, due to the required Korean content for some components. Based on successfully supplying the equipment referenced above, it has been concluded that AG-1 is a working document and can be successfully implemented. It provides the requirements necessary for performance, design, construction, acceptance testing, and quality assurance of equipment used as components in nuclear air and gas treatment systems in nuclear facilities. The paper also addresses lessons learned and aspects of mixing US design and US built components in Korean built assemblies

  19. Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

    International Nuclear Information System (INIS)

    1994-07-01

    The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately

  20. Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately.

  1. Evaluation of practicality of ASME code, Section XI

    International Nuclear Information System (INIS)

    Mattu, R.K.; Lauderdale, J.R.; Liu, S.N.; Lance, J.J.

    2004-01-01

    Many nuclear power plants have found that it is impractical or unduly burdensome to comply with some ASME Boiler and Pressure Code provisions and have sought relief from those provisions from the Nuclear Regulatory Commission. An Electric Power Research Institute (EPRI) project is evaluating such Code provisions and alternatives to them that will meet the safety intent of the Code with less burden on utilities. The methodology is to extract data from an on-line data base of relief requests since 1980, analyse the data to identify burdensome provisions for which there are satisfactory alternatives, and recommend changes in the Code to the ASME. (author)

  2. ASME method for particle reconstruction

    International Nuclear Information System (INIS)

    Ierusalimov, A.P.

    2009-01-01

    The method of approximate solution of motion equation (ASME) was used to reconstruct the parameters for charged particles. It provides a good precision for momentum, angular and space parameters of particles in coordinate detectors. The application of the method for CBM, HADES and MPD/NICA setups is discussed

  3. ASME nuclear codes and standards risk management strategic planning

    International Nuclear Information System (INIS)

    Hill, Ralph S. III; Balkey, Kenneth R.; Erler, Bryan A.; Wesley Rowley, C.

    2007-01-01

    This paper is prepared in honor and in memory of the late Professor Emeritus Yasuhide Asada to recognize his contributions to ASME Nuclear Codes and Standards initiatives, particularly those related to risk-informed technology and System Based Code developments. For nearly two decades, numerous risk-informed initiatives have been completed or are under development within the ASME Nuclear Codes and Standards organization. In order to properly manage the numerous initiatives currently underway or planned for the future, the ASME Board on Nuclear Codes and Standards (BNCS) has an established Risk Management Strategic Plan (Plan) that is maintained and updated by the ASME BNCS Risk Management Task Group. This paper presents the latest approved version of the plan beginning with a background of applications completed to date, including the recent probabilistic risk assessment (PRA) standards developments for nuclear power plant applications. The paper discusses planned applications within ASME Nuclear Codes and Standards that will require expansion of the ASME PRA Standard to support new advanced light water reactor and next generation reactor developments, such as for high temperature gas-cooled reactors. Emerging regulatory developments related to risk-informed, performance- based approaches are summarized. A long-term vision for the potential development and evolution to a nuclear systems code that adopts a risk-informed approach across a facility life-cycle (design, construction, operation, maintenance, and closure) is also summarized. Finally, near term and long term actions are defined across the ASME Nuclear Codes and Standards organizations related to risk management, including related U.S. regulatory activities. (author)

  4. CONAGT's place in ASME's centennial year

    International Nuclear Information System (INIS)

    Miller, W.H. Jr.

    1985-01-01

    A status report on ASME's Committee on Nuclear Air and Gas Treatment (CONAGT) is presented. This year ASME celebrates its centennial while CONAGT issues its first code sections covering fans, blowers, and refrigeration equipment. Significant code related CONAGT activities are covered as well as an explanation of CONAGT's place in the ASME organization

  5. 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

  6. ASME nuclear codes and standards risk management strategic plan

    International Nuclear Information System (INIS)

    Balkey, Kenneth R.

    2003-01-01

    Over the past 15 years, several risk-informed initiatives have been completed or are under development within the ASME Nuclear Codes and Standards organization. In order to better manage the numerous initiatives in the future, the ASME Board on Nuclear Codes and Standards has recently developed and approved a Risk Management Strategic Plan. This paper presents the latest approved version of the plan beginning with a background of applications completed to date, including the recent issuance of the ASME Standard for Probabilistic Risk Assessment (PRA) for Nuclear Power Plant Applications. The paper discusses potential applications within ASME Nuclear Codes and Standards that may require expansion of the PRA Standard, such as for new generation reactors, or the development of new PRA Standards. A long-term vision for the potential development and evolution to a nuclear systems code that adopts a risk-informed approach across a facility life-cycle (design, construction, operation, maintenance, and closure) is summarized. Finally, near term and long term actions are defined across the ASME Nuclear Codes and Standards organizations related to risk management, and related U.S. regulatory activities are also summarized. (author)

  7. Inservice inspection procedures and training according to the ASME code

    International Nuclear Information System (INIS)

    Greenwald, S.M.; Chockie, L.J.

    1987-01-01

    Mandatory training of the technical staff at a nuclear power plant is of paramount importance if we are to avoid costly plant shutdowns. This training should include the requirements for both Preservice and Inservice Inspection, in addition to Quality Assurance procedures as required by the American Society of Mechanical Engineers (ASME) Code. The training is best accomplished by utilizing instructors who are thoroughly familiar with plant operations and the ASME Code, as well as serving on one of the Code committees. This paper focuses on the Inservice Inspection procedures and the results of an intensive training effort to implement such procedures. (author)

  8. Future direction of ASME nuclear codes and standards

    International Nuclear Information System (INIS)

    Ennis, Kevin; Sheehan, Mark E.

    2003-01-01

    While the nuclear power industry in the US is in a period of stasis, there continues to be a great deal of activity in the ASME nuclear standards development arena. As plants age, the need for new approaches in standardization changes with the changing needs of the industry. New tools are becoming available in the form of risk analysis, and this is finding its way into more and more of ASME's standards activities. This paper will take a look at the direction that ASME nuclear Codes and Standards are heading in this and other areas, as well as taking a look at some advance reactor concepts and plans for standards to address new technologies

  9. Dynamic fracture toughness of ASME SA508 Class 2a ASME SA533 grade A Class 2 base and heat affected zone material and applicable weld metals

    International Nuclear Information System (INIS)

    Logsdon, W.A.; Begley, J.A.; Gottshall, C.L.

    1978-03-01

    The ASME Boiler and Pressure Vessel Code, Section III, Article G-2000, requires that dynamic fracture toughness data be developed for materials with specified minimum yield strengths greater than 50 ksi to provide verification and utilization of the ASME specified minimum reference toughness K/sub IR/ curve. In order to qualify ASME SA508 Class 2a and ASME SA533 Grade A Class 2 pressure vessel steels (minimum yield strengths equal 65 kip/in. 2 and 70 kip/in. 2 , respectively) per this requirement, dynamic fracture toughness tests were performed on these materials. All dynamic fracture toughness values of SA508 Class 2a base and HAZ material, SA533 Grade A Class 2 base and HAZ material, and applicable weld metals exceeded the ASME specified minimum reference toughness K/sub IR/ curve

  10. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2012-09-01

    In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This report discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.

  11. ASME Code Efforts Supporting HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    D.K. Morton

    2011-09-01

    In 1999, an international collaborative initiative for the development of advanced (Generation IV) reactors was started. The idea behind this effort was to bring nuclear energy closer to the needs of sustainability, to increase proliferation resistance, and to support concepts able to produce energy (both electricity and process heat) at competitive costs. The U.S. Department of Energy has supported this effort by pursuing the development of the Next Generation Nuclear Plant, a high temperature gas-cooled reactor. This support has included research and development of pertinent data, initial regulatory discussions, and engineering support of various codes and standards development. This report discusses the various applicable American Society of Mechanical Engineers (ASME) codes and standards that are being developed to support these high temperature gascooled reactors during construction and operation. ASME is aggressively pursuing these codes and standards to support an international effort to build the next generation of advanced reactors so that all can benefit.

  12. ASME Section VIII Recertification of a 33,000 Gallon Vacuum-jacketed LH2 Storage Vessel for Densified Hydrogen Testing at NASA Kennedy Space Center

    Science.gov (United States)

    Swanger, Adam M.; Notardonato, William U.; Jumper, Kevin M.

    2015-01-01

    The Ground Operations Demonstration Unit for Liquid Hydrogen (GODU-LH2) has been developed at NASA Kennedy Space Center in Florida. GODU-LH2 has three main objectives: zero-loss storage and transfer, liquefaction, and densification of liquid hydrogen. A cryogenic refrigerator has been integrated into an existing, previously certified, 33,000 gallon vacuum-jacketed storage vessel built by Minnesota Valley Engineering in 1991 for the Titan program. The dewar has an inner diameter of 9.5 and a length of 71.5; original design temperature and pressure ranges are -423 F to 100 F and 0 to 95 psig respectively. During densification operations the liquid temperature will be decreased below the normal boiling point by the refrigerator, and consequently the pressure inside the inner vessel will be sub-atmospheric. These new operational conditions rendered the original certification invalid, so an effort was undertaken to recertify the tank to the new pressure and temperature requirements (-12.7 to 95 psig and -433 F to 100 F respectively) per ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. This paper will discuss the unique design, analysis and implementation issues encountered during the vessel recertification process.

  13. The ASME research task force on risk-based in-service inspection

    International Nuclear Information System (INIS)

    Balkey, K.R.; Chapman, O.J.V.

    1997-01-01

    The use of risk-based methods in the development of in-service inspection (ISI) and in-service testing (IST) programs for nuclear power plant and other industrial applications has been studied for the last several years through the American Society of Mechanical Engineers Centre for Research and Technology Development (ASME 1991, 1992, 1994, 1996). The results of this work are being used as a foundation to develop specific requirements for implementation of risk-based technology in ASME Codes and Standards, regulatory requirements and industry programs both in the U.S. and other countries. This paper provides a brief overview of the ASME Research Methodology and how it has been adapted for application to the inspection of piping within the USA. It also relates how the reliability of nondestructive examination (NDE) methods for pressure boundary components can impact the risk and discusses the relationship between this and NDE qualification/demonstration now being implemented in Europe and the USA. (orig.)

  14. Guideline on in-service testing (IST) of ECCS for nuclear power stations in Japan

    International Nuclear Information System (INIS)

    Yamashita, N.; Honjin, M.; Yamazaki, M.; Iwami, H.

    2001-01-01

    In Japan, METI (Ministry of Economy, Trade and Industry) periodic inspections are conducted every 13 months in accordance with the EUIL (Electric Utility Industry Law) for light water reactors. On the other hand, during operation, many kinds of function tests are conducted in accordance with Technical Specifications or utilities voluntary test procedures. However, these maintenance activities during outages are in other respects considered too costly and more than necessary because of its predetermined maintenance interval and its old-fashioned full scope maintenance menu. In these circumstances, the ASME O and M Codes and Standards, which are applied to nuclear power plants in the United States, came to be one of the issues to be considered among Japanese utilities. This is because the ASME O and M Codes and Standards, which are developed to verify the operational readiness of equipment or systems by condition based maintenance or performance testing without overhauls, will be considered to suggest a rational alternative to our existing maintenance activities. Because there had been no standard for function tests other than the METI inspection procedures in Japan, activities started for the development of Japanese Operation and Maintenance Guideline (hereinafter called the Japanese OM Guideline or the Guideline) for function tests and condition monitoring. The Japanese OM Guideline is not proclaimed for regulatory use, but the long-term objective is to provide a substantial basis to change the current maintenance activities including regulatory inspection. The Guideline is composed of five parts, which are general requirements, pumps, valves, snubbers and ECCS systems. This paper gives a summary of the ECCS part of the Japanese OM Guideline, in particular, focusing on the differences between the Guideline and the ASME O and M Standards, and topics or discussion during their establishment. (author)

  15. A comparative study for SMART steam generator sizing based on ASME and Russian standard

    International Nuclear Information System (INIS)

    Kim, Y. W.; Kim, J. I.; Jang, M. H.

    2000-01-01

    A systematic comparison of ASME and Russian standard with respect to the design of SMART steam generator has been carried out. Classification of allowable stress in the Russian standard is quite different from that of ASME. Allowable stress of Russian standard and stress intensity defined in ASME were compared for various steam generator tube material as a function of design temperature. Equations and methodology of determining the thickness for the important parts of steam generator have been analyzed. For the tube subjected to internal and/or external pressure, Russian standard use the same equation in the sizing of tube with different allowable stress. However, ASME use different equations with the same value of allowable stress intensity. The hydraulic test pressure of ASME was also compared with that of Russian standard. In general, hydraulic test pressure determined by Russian standard is higher since it considers difference between allowable stress of test temperature and that of design temperature

  16. 46 CFR 52.01-2 - Adoption of section I of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section I of the ASME Boiler and Pressure...) MARINE ENGINEERING POWER BOILERS General Requirements § 52.01-2 Adoption of section I of the ASME Boiler and Pressure Vessel Code. (a) Main power boilers and auxiliary boilers shall be designed, constructed...

  17. Passport of global nuclear business. ASME code certificate acquirement and inspection practices

    International Nuclear Information System (INIS)

    Kawabata, Hiroyuki; Terajima, Makoto; Anami, Kazuhiro

    2010-01-01

    There are possibilities of Japanese nuclear industries to participate in global business such as new and additional construction of nuclear power plants in US and also Asian and other developing countries in the world. It is requisite to acquire ASME code certificate for global business participation, just as passport. This article consists of five papers on present status of ASME code certificate acquirement and inspection practices of nuclear components vendors in the area of Japanese nuclear business. Activities of JSME Committee on Power Generation Facility Codes to make JSME codes corresponded to ASME nuclear codes and standards for their international deployment are also described. (T. Tanaka)

  18. ASME nuclear codes and standards: Recent technical initiatives

    International Nuclear Information System (INIS)

    Feigel, R. E.

    1995-01-01

    Although nuclear power construction is currently in a hiatus in the US, ASME and its volunteer committees remain committed to continual improvements in the technical requirements in its nuclear codes. This paper provides an overview of several significant recent revisions to ASME' s nuclear codes. Additionally, other important initiatives currently being addressed by ASME committees will be described. With the largest population of operating light water nuclear plants in the world and worldwide use of its nuclear codes, ASME continues to support technical advancements in its nuclear codes and standards. While revisions of various magnitude are an ongoing process, several recent revisions embody significant changes based on state of the art design philosophy and substantial industry experience. In the design area, a significant revisions has recently been approved which will significantly reduce conservatisms in seismic piping design as well as provide simplified design rules. Major revisions have also been made to the requirements for nuclear material manufacturers and suppliers, which should result in clearer understanding of this difficult administrative area of the code. In the area of Section XI inservice rules, substantial studies are underway to investigate the application of probabilistic, risked based inspection in lieu of the current deterministic inspection philosophy. While much work still is required in this area, it is an important potential application of the emerging field of risk based inspection

  19. Tevatron HTS power lead test

    International Nuclear Information System (INIS)

    Feher, S.; Carcagno, R.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.; Tompkins, J.C.

    2004-01-01

    Two pairs of ASC 6 kA power leads developed for the Tevatron were successfully tested at Fermilab at over-current conditions. Stable operation was achieved while operating at a current of 9.56 kA for five hours and while continuously ramping between 0-9.56 kA at a ramp rate of 200 A/s for one hour. The minimum required liquid nitrogen flow rate was measured to be 1.5 g/s at 10 kA. After ramping up to 10 kA at 200A/s, it took only 15 minutes to stabilize the upper copper section of the lead with a flow of 1.8 g/s of liquid nitrogen vapor. Testing under extreme operating conditions--270-370 kPa liquid nitrogen vapor pressure and over 0.1 T external magnetic field--demonstrated that the HTS part of the lead can safely operate in the current sharing mode and that this design has large operating margin

  20. A strategy for implementation of experience based seismic equipment qualification in IEEE and ASME industry standards

    International Nuclear Information System (INIS)

    Adams, T.M.

    1996-01-01

    In the past 20 years, extensive data on the performance of mechanical and electric equipment during actual strong motion earthquakes and seismic qualification tests has been accumulated. Recognizing that an experience based approach provides a technically sound and cost effective method for the seismic qualification of some or certain equipment, the IEEE Nuclear Power Engineering Committee and the ASME Committee on Qualification of Mechanical Equipment established a Special Working Group to investigate the incorporation of experienced based methods into the industry consensus codes and standards currently used in the seismic qualification of Seismic Category Nuclear Power Plant equipment. This paper presents the strategy (course of action) which was developed by the Special Working Group for meeting this objective of incorporation of experience based seismic qualification standards used in the design and seismic qualification of seismic category nuclear power plant equipment. This strategy was recommended to both chartering organizations, the IEEE Nuclear Power Engineering Committee and the ASME Committee on Qualification of Mechanical Equipment for their consideration and implementation. The status of the review and implementation of the Special Working Group's recommended strategy by the sponsoring organization is also discussed

  1. Energy Systems Test Area (ESTA). Power Systems Test Facilities

    Science.gov (United States)

    Situ, Cindy H.

    2010-01-01

    This viewgraph presentation provides a detailed description of the Johnson Space Center's Power Systems Facility located in the Energy Systems Test Area (ESTA). Facilities and the resources used to support power and battery systems testing are also shown. The contents include: 1) Power Testing; 2) Power Test Equipment Capabilities Summary; 3) Source/Load; 4) Battery Facilities; 5) Battery Test Equipment Capabilities Summary; 6) Battery Testing; 7) Performance Test Equipment; 8) Battery Test Environments; 9) Battery Abuse Chambers; 10) Battery Abuse Capabilities; and 11) Battery Test Area Resources.

  2. Operating nuclear plant feedback to ASME and French codes

    International Nuclear Information System (INIS)

    Journet, J.; O'Donnell, W.J.

    1996-01-01

    The French have an advantage in nuclear plant operating experience feedback due to the highly centralized nature of their nuclear industry. There is only one utility in charge of design as well as operations (EDF) and only one reactor vendor (Framatome). The ASME Code has played a key role in resolving technical issues in the design and operation of nuclear plants since the inception of nuclear power. The committee structure of the Code brings an ideal combination of senior technical people with both broad and specialized experience to bear on complex how safe is safe enough technical issues. The authors now see an even greater role for the ASME Code in a proposed new regulatory era for the US nuclear industry. The current legalistic confrontational regulatory era has been quite destructive. There now appears to be a real opportunity to begin a new era of technical consensus as the primary means for resolving safety issues. This change can quickly be brought about by having the industry take operating plant problems and regulatory technical issues directly to the ASME Code for timely resolution. Surprisingly, there is no institution in the US nuclear industry with such a mandate. In fact, the industry is organized to feedback through the Nuclear Regulatory Commission issues which could be far better resolved through the ASME Code. Major regulatory benefits can be achieved by closing this loop and providing systematic interaction with the ASME Code. The essential elements of a new regulatory era and ideas for organizing US institutional industry responsibilities, taken from the French experience, are described in this paper

  3. ASME N511-19XX, Standard for periodic in-service testing of nuclear air treatment, heating, ventilating and air conditioning systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    A draft version of the Standard is presented in this document. The Standard covers the requirements for periodic in-service testing of nuclear safety-related air treatment, heating, ventilating, and air conditioning systems in nuclear facilities. The Standard provides a basis for the development of test programs and does not include acceptance criteria, except in cases where the results of one test influence the performance of other tests. The Standard covers general inspection and test requirements, reference values, inspection and test requirements, generic tests, acceptance criteria, in-service test requirements, testing following an abnormal incident, corrective action requirements, and quality assurance. Mandatory appendices provide a visual inspection checklist and four test procedures. Non-mandatory appendices provide additional information and guidance on mounting frame pressure leak test procedure, corrective action, challenge gas substitute selection criteria, and test program development. 8 refs., 10 tabs.

  4. ASME N511-19XX, Standard for periodic in-service testing of nuclear air treatment, heating, ventilating and air conditioning systems

    International Nuclear Information System (INIS)

    1997-01-01

    A draft version of the Standard is presented in this document. The Standard covers the requirements for periodic in-service testing of nuclear safety-related air treatment, heating, ventilating, and air conditioning systems in nuclear facilities. The Standard provides a basis for the development of test programs and does not include acceptance criteria, except in cases where the results of one test influence the performance of other tests. The Standard covers general inspection and test requirements, reference values, inspection and test requirements, generic tests, acceptance criteria, in-service test requirements, testing following an abnormal incident, corrective action requirements, and quality assurance. Mandatory appendices provide a visual inspection checklist and four test procedures. Non-mandatory appendices provide additional information and guidance on mounting frame pressure leak test procedure, corrective action, challenge gas substitute selection criteria, and test program development. 8 refs., 10 tabs

  5. ASM Inaugural Lecture 2010: Single crystal X-ray structural determination: A powerful technique for natural products research and drug discovery

    International Nuclear Information System (INIS)

    Fun Hoong Kun; Chantrapromma, S.; Boonnak, N.; Lee, V.S.

    2010-01-01

    Drug discovery from natural products resources have been extensively studied worldwide because natural products with their great structural diversity have traditionally provided most of the drugs in use. They offer major opportunities for finding novel low molecular weight leading-structures that are active against a wide range of assay targets. The most important step in the discovery process is the identification of compounds with interesting biological activity. Single crystal X-ray structure determination is a powerful technique for natural products research and drug discovery. The detailed three-dimensional structures that emerge can be co-related to the activities to these structures. In this article the following is presented: (i) co-crystal and disorder structures; (ii) determination of absolute configuration and (iii) the ability to distinguish between whether a natural product compound is a natural product or a natural product artifact. All these three properties are unique to the technique of single crystal X-ray structure determination. Case (iii) was demonstrated with a compound containing a chromene ring, namely macluraxanthone (which was isolated from Cratoxylum formosum subsp. pruniflorum, a Thai medicinal plant). (author)

  6. Preliminary research on eddy current bobbin quantitative test for heat exchange tube in nuclear power plant

    Science.gov (United States)

    Qi, Pan; Shao, Wenbin; Liao, Shusheng

    2016-02-01

    For quantitative defects detection research on heat transfer tube in nuclear power plants (NPP), two parts of work are carried out based on the crack as the main research objects. (1) Production optimization of calibration tube. Firstly, ASME, RSEM and homemade crack calibration tubes are applied to quantitatively analyze the defects depth on other designed crack test tubes, and then the judgment with quantitative results under crack calibration tube with more accuracy is given. Base on that, weight analysis of influence factors for crack depth quantitative test such as crack orientation, length, volume and so on can be undertaken, which will optimize manufacture technology of calibration tubes. (2) Quantitative optimization of crack depth. Neural network model with multi-calibration curve adopted to optimize natural crack test depth generated in in-service tubes shows preliminary ability to improve quantitative accuracy.

  7. Acceptance test report: Backup power system

    International Nuclear Information System (INIS)

    Cole, D.B.

    1996-01-01

    Acceptance Test Report for construction functional testing of Project W-030 Backup Power System. Project W-030 provides a ventilation upgrade for the four Aging Waste Facility tanks. Backup power includes a single 125 KW diesel generator, three 10-kva uninterruptible power supply units, and all necessary control

  8. Revision of the ASME nuclear quality assurance standard and its historical background

    International Nuclear Information System (INIS)

    Suzuki, Tetsuya

    2009-01-01

    ASME NQA-1-2008 'Quality Assurance Requirements for Nuclear Facility Applications' will be endorsed by US NRC by the end of 2009. This standard will apply to design, construction and operation of nuclear power plants newly erected in USA. It is important to Japanese vendors developing nuclear business in USA. Historical background, significance of revision and main revised points of the ASME nuclear quality assurance standard are described in the present paper. (T. Tanaka)

  9. Adoption of ASME Code Section XI for ISI to Research Reactors

    International Nuclear Information System (INIS)

    Tawfik, Y.E.; El-sesy, I.A.; Shaban, H.I.; Ibrahim, M.M.

    2002-01-01

    ETRR-2 (Second Egyptian thermal research reactor) is a multi-purpose, pool- type reactor with an open water surface and variable core arrangement. The core power is 22 MWth, cooled and moderated by light water and with beryllium reflectors. It contains plate- type fuel elements (MTR type, 19.7% enriched uranium) with aluminum clad. The ETRR-2 reactor consist of 57 systems and around 200 subsystems. These systems contain many mechanical components such as tanks, pipes, valves, pumps, heat exchangers, cooling tower, air compressors, and supports. In this present work, a trial was made to adopt the general requirements of ASME code, section XI to ETRR-2 research reactor. ASME (American Society of Mechanical Engineers) boiler and pressure vessel Code, section XI, provides requirements for in-service inspection (ISI) and in-service testing (IST) of components and systems, and repair/replacement activities in a nuclear power plant. Also, IAEA (International Atomic Energy Authority) has published some recommendations for ISI for research reactors similar to that rules and requirements specified in ASME. The complete ISI program requires several steps that have to be performed in sequence. These steps are described in many logic flow charts (LFC's). These logic flow charts include; the general LFC's for all steps required to complete ISI program, the LFC's for examination requirements, the LFC's for flaw evaluation modules, and the LFC's for acceptability of welds for class 1 components. This program includes, also, the inspection program for welded parts of the reactor components during its lifetime. This inspection program is applied for each system and subsystem of ETRR-2 reactor. It includes the examination area type, the component type, the part to be examined, the weld type, the examination method, the inspection program schedule, and the detailed figures of the welded components. (authors)

  10. ASME nuclear codes and standards: Scope of coverage and current initiatives

    International Nuclear Information System (INIS)

    Eisenberg, G. M.

    1995-01-01

    The objective of this paper is to address the broad scope of coverage of nuclear codes, standards and guides produced and administered by the American Society of Mechanical Engineers (ASME). Background information is provided regarding the evolution of the present activities. Details are provided on current initiatives intended to permit ASME to meet the needs of a changing nuclear industry on a worldwide scale. During the early years of commercial nuclear power, ASME produced a code for the construction of nuclear vessels used in the reactor coolant pressure boundary, containment and auxiliary systems. In response to industry growth, ASME Code coverage soon broadened to include rules for construction of other nuclear components, and inservice inspection of nuclear reactor coolant systems. In the years following this, the scope of ASME nuclear codes, standards and guides has been broadened significantly to include air cleaning activities for nuclear power reactors, operation and maintenance of nuclear power plants, quality assurance programs, cranes for nuclear facilities, qualification of mechanical equipment, and concrete reactor vessels and containments. ASME focuses on globalization of its codes, standards and guides by encouraging and promoting their use in the international community and by actively seeking participation of international members on its technical and supervisory committees and in accreditation activities. Details are provided on current international representation. Initiatives are underway to separate the technical requirements from administrative and enforcement requirements, to convert to hard metric units, to provide for non-U. S. materials, and to provide for translations into non-English languages. ASME activity as an accredited ISO 9000 registrar for suppliers of mechanical equipment is described. Rules are being developed for construction of containment systems for nuclear spent fuel and high-level waste transport packagings. Intensive

  11. Evaluation of flaws in ferritic piping: ASME Code Appendix J, Deformation Plasticity Failure Assessment Diagram (DPFAD)

    International Nuclear Information System (INIS)

    Bloom, J.M.

    1991-08-01

    This report summarizes the methods and bases used by an ASME Code procedure for the evaluation of flaws in ferritic piping. The procedure is currently under consideration by the ASME Boiler and Pressure Vessel Code Committee of Section 11. The procedure was initially proposed in 1985 for the evaluation of the acceptability of flaws detected in piping during in-service inspection for certain materials, identified in Article IWB-3640 of the ASME Boiler and Pressure Vessel Code Section 11 ''Rules for In-service Inspection of Nuclear Power Plant Components.'' for which the fracture toughness is not sufficiently high to justify acceptance based solely on the plastic limit load evaluation methodology of Appendix C and IWB-3641. The procedure, referred to as Appendix J, originally included two approaches: a J-integral based tearing instability (J-T) analysis and the deformation plasticity failure assessment diagram (DPFAD) methodology. In Appendix J, a general DPFAD approach was simplified for application to part-through wall flows in ferritic piping through the use of a single DPFAD curve for circumferential flaws. Axial flaws are handled using two DPFAD curves where the ratio of flaw depth to wall thickness is used to determine the appropriate DPFAD curve. Flaws are evaluated in Appendix J by comparing the actual pipe applied stress with the allowable stress with the appropriate safety factors for the flaw size at the end of the evaluation period. Assessment points for circumferential and axial flaws are plotted on the appropriate failure assessment diagram. In addition, this report summarizes the experimental test predictions of the results of the Battelle Columbus Laboratory experiments, the Eiber experiments, and the JAERI tests using the Appendix J DPFAD methodology. Lastly, this report also provides guidelines for handling residual stresses in the evaluation procedure. 22 refs., 13 figs., 5 tabs

  12. Estimates of margins in ASME Code strength values for stainless steel nuclear piping

    International Nuclear Information System (INIS)

    Ware, A.G.

    1995-01-01

    The margins in the ASME Code stainless steel allowable stress values that can be attributed to the variations in material strength are evaluated for nuclear piping steels. Best-fit curves were calculated for the material test data that were used to determine allowable stress values for stainless steels in the ASME Code, supplemented by more recent data, to estimate the mean stresses. The mean yield stresses (on which the stainless steel S m values are based) from the test data are about 15 to 20% greater than the ASME Code yield stress values. The ASME Code yield stress values are estimated to approximately coincide with the 97% confidence limit from the test data. The mean and 97% confidence limit values can be used in the probabilistic risk assessments of nuclear piping

  13. Analysis of preservice inspection relief requests and recommendations for ASME code changes

    International Nuclear Information System (INIS)

    Cook, J.F.

    1985-05-01

    NRC regulations require that preservice inspection (PSI) of nuclear plants be performed in accordance with referenced editions and addenda of Division 1 rules of Section XI, ''Rules for Inservice Inspection of Nuclear Power Plant Components'', of the ASME Boiler and Pressure Vessel Code (ASME Code). The regulations permit applicants to request and obtain relief from the NRC from specific ASME Code requirements that are determined to be impractical. Applicant requests for relief from preservice inspection (PSI) requirements were compiled and analyzed. From this data, covering a total of 178 relief requests, common problems with examination requirements were identified. Changes to examination requirements to solve selected problems are proposed. By following later ASME Code requirements, 46 out of the 178 relief requests can be eliminated. Implementing proposed Code changes would eliminate another 25 relief requests, leaving 107 relief requests out of the original 178 relief requests covered by this survey

  14. The ASME section XI code, past, present, and future

    International Nuclear Information System (INIS)

    Anderson, W.F.; Bush, S.H.; Chockie, L.J.

    1982-01-01

    A little over a decade has passed since the first drafts of the Codes effecting the program of inservice inspections in the United States have been implemented and enforced in operating power plants. During the ensuing years, advantage has been taken of results of the application of the program of inspections and tests. In several instances, the Code has been revised to correct deficiencies, and additional rules have been added to accommodate additional systems and portions of the plant which were deemed to also be important to safety. Additional rules have now been published to recognize the liquid metal cooled reactor systems and the gas cooled reactor systems along with the water cooled reactor systems. While the past and the present can be stated with considerable certainty, opinions of the authors, each of whom is a member of the American Society of Mechanical Engineers (ASME) Code Committee, are advanced as to their plans which will reveal the direction the Code is projected to follow in the future. (author)

  15. 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

  16. WhalePower tubercle blade power performance test report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-15

    Toronto-based WhalePower Corporation has developed turbine blades that are modeled after humpback whale flippers. The blades, which incorporate tubercles along the leading edge of the blade, have been fitted to a Wenvor 25 kW turbine installed in North Cape, Prince Edward Island at a test site for the Wind Energy Institute of Canada (WEICan). A test was conducted to characterize the power performance of the prototype wind turbine. This report described the wind turbine configuration with particular reference to turbine information, power rating, blade information, tower information, control systems and grid connections. The test site was also described along with test equipment and measurement procedures. Information regarding power output as a function of wind speed was included along with power curves, power coefficient and annual energy production. The results for the power curve and annual energy production contain a level of uncertainty. While measurements for this test were collected and analyzed in accordance with International Electrotechnical Commission (IEC) standards for performance measurements of electricity producing wind turbines (IEC 61400-12-1), the comparative performance data between the prototype WhalePower wind turbine blade and the Wenvor standard blade was not gathered to IEC data standards. Deviations from IEC-61400-12-1 procedures were listed. 6 tabs., 16 figs., 3 appendices.

  17. Globalization of ASME Nuclear Codes and Standards

    International Nuclear Information System (INIS)

    Swayne, Rick; Erler, Bryan A.

    2006-01-01

    With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. (authors)

  18. Automated System Tests High-Power MOSFET's

    Science.gov (United States)

    Huston, Steven W.; Wendt, Isabel O.

    1994-01-01

    Computer-controlled system tests metal-oxide/semiconductor field-effect transistors (MOSFET's) at high voltages and currents. Measures seven parameters characterizing performance of MOSFET, with view toward obtaining early indication MOSFET defective. Use of test system prior to installation of power MOSFET in high-power circuit saves time and money.

  19. Practical experience applied to the design of injection and sample manifolds to perform in-place surveillance tests according to ANSI/ASME N-510

    Energy Technology Data Exchange (ETDEWEB)

    Banks, E.M.; Wikoff, W.O.; Shaffer, L.L. [NUCON International, Inc., Columbus, OH (United States)

    1997-08-01

    At the current level of maturity and experience in the nuclear industry, regarding testing of air treatment systems, it is now possible to design and qualify injection and sample manifolds for most applications. While the qualification of sample manifolds is still in its infancy, injection manifolds have reached a mature stage that helps to eliminate the {open_quotes}hit or miss{close_quotes} type of design. During the design phase, manifolds can be adjusted to compensate for poor airflow distribution, laminar flow conditions, and to take advantage of any system attributes. Experience has shown that knowing the system attributes before the design phase begins is an essential element to a successful manifold design. The use of a spreadsheet type program commonly found on most personal computers can afford a greater flexibility and a reduction in time spent in the design phase. The experience gained from several generations of manifold design has culminated in a set of general design guidelines. Use of these guidelines, along with a good understanding of the type of testing (theoretical and practical), can result in a good manifold design requiring little or no field modification. The requirements for manifolds came about because of the use of multiple banks of components and unconventional housing inlet configurations. Multiple banks of adsorbers and pre and post HEPA`s required that each bank be tested to insure that each one does not exceed a specific allowable leakage criterion. 5 refs., 5 figs., 1 tab.

  20. ACIGA's high optical power test facility

    International Nuclear Information System (INIS)

    Ju, L; Aoun, M; Barriga, P

    2004-01-01

    Advanced laser interferometer detectors utilizing more than 100 W of laser power and with ∼10 6 W circulating laser power present many technological problems. The Australian Consortium for Interferometric Gravitational Astronomy (ACIGA) is developing a high power research facility in Gingin, north of Perth, Western Australia, which will test techniques for the next generation interferometers. In particular it will test thermal lensing compensation and control strategies for optical cavities in which optical spring effects and parametric instabilities may present major difficulties

  1. 46 CFR 53.01-3 - Adoption of section IV of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Adoption of section IV of the ASME Boiler and Pressure...) MARINE ENGINEERING HEATING BOILERS General Requirements § 53.01-3 Adoption of section IV of the ASME Boiler and Pressure Vessel Code. (a) Heating boilers shall be designed, constructed, inspected, tested...

  2. Production LHC HTS power lead test results

    CERN Document Server

    Tartaglia, M; Fehér, S; Huang, Y; Orris, D F; Pischalnikov, Y; Rabehl, Roger Jon; Sylvester, C D; Zbasnik, J

    2005-01-01

    The Fermilab Magnet test facility has built and operated a test stand to characterize the performance of HTS power leads. We report here the results of production tests of 20 pairs of 7.5 kA HTS power leads manufactured by industry for installation in feed boxes for the LHC Interaction Region quadrupole strings. Included are discussions of the thermal, electrical, and quench characteristics under "standard" and "extreme" operating conditions, and the stability of performance across thermal cycles.

  3. Production LHC HTS power lead test results

    International Nuclear Information System (INIS)

    Tartaglia, M.A.; Carcagno, R.H.; Feher, S.; Huang, Y.; Orris, D.F.; Pischalnikov, Y.; Rabehl, R.J.; Sylvester, C.; Zbasnik, J.

    2004-01-01

    The Fermilab Magnet test facility has built and operated a test stand to characterize the performance of HTS power leads. We report here the results of production tests of 20 pairs of 7.5 kA HTS power leads manufactured by industry for installation in feed boxes for the LHC Interaction Region quadrupole strings. Included are discussions of the thermal, electrical, and quench characteristics under ''standard'' and ''extreme'' operating conditions, and the stability of performance across thermal cycles

  4. High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

    Science.gov (United States)

    Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

    System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

  5. Review of ASME nuclear codes and standards- subcommittee on repairs, replacements, and modifications

    International Nuclear Information System (INIS)

    Mawson, T.J.

    1990-01-01

    As requested by the ASME board on Nuclear Codes and Standards, the Pressure Vessel Research Committee initiated a project to review Sections III and XI of the ASME Boiler and Pressure Vessel Code for the purposes of improving, clarifying, providing transition, consistency, compatibility, and simplifying code requirements. The project was organized with six subcommittees to address various Code activities: design; tests and examinations; documentation; quality assurance; repair, replacement and modification; and general requirements. This paper discusses how the subcommittee on repair, replacement and modification was organized to review the repair, replacement and modification requirements of the ASME boiler and pressure vessel code, Section III and Section XI for Class 1, 2, and 3 and MC components and their supports, and other documents of the nuclear industry related to the repair, replacement and modification requirements of the ASME code

  6. A power recirculating test rig for ball screw endurance tests

    Directory of Open Access Journals (Sweden)

    Giberti Hermes

    2016-01-01

    Full Text Available A conceptual design of an innovative test rig for endurance tests of ball screws is presented in this paper. The test rig layout is based on the power recirculating principle and it also allows to overtake the main critical issues of the ball screw endurance tests. Among these there are the high power required to make the test, the lengthy duration of the same and the high loads between the screw and the frame that holds it. The article describes the test rig designed scheme, the kinematic expedients to be adopted in order to obtain the required performance and functionality and the sizing procedure to choose the actuation system.

  7. The 1997 NRC IST workshops and the status of questions and issues directed to the ASME O and M committee

    International Nuclear Information System (INIS)

    DiBiasio, A.M.

    1998-05-01

    This paper describes the results of the four NRC Inservice Testing (IST) Workshops which were held in early 1997 pertaining to NRC Inspection Procedure P 73756, Inservice Testing of Pumps and Valves. It also presents the status of the ASME code committees' resolution of certain questions forwarded to the ASME by the NRC. These questions relate to code interpretations, inconsistencies in the code, and industry concerns that are most appropriately resolved through the ASME consensus process. The ASME committees reviewed the questions at their December 1997 and March 1998 code meetings. Of particular interest are those questions for which the ASME code committees did not agree with the NRC response. These questions, as well as those which the committees provided some additional insight or input, are presented in this paper

  8. Testing power-law cross-correlations: Rescaled covariance test

    Czech Academy of Sciences Publication Activity Database

    Krištoufek, Ladislav

    2013-01-01

    Roč. 86, č. 10 (2013), 418-1-418-15 ISSN 1434-6028 R&D Projects: GA ČR GA402/09/0965 Institutional support: RVO:67985556 Keywords : power-law cross-correlations * testing * long-term memory Subject RIV: AH - Economics Impact factor: 1.463, year: 2013 http://library.utia.cas.cz/separaty/2013/E/kristoufek-testing power-law cross-correlations rescaled covariance test.pdf

  9. SDZ ASM 981: an emerging safe and effective treatment for atopic dermatitis.

    Science.gov (United States)

    Luger, T; Van Leent, E J; Graeber, M; Hedgecock, S; Thurston, M; Kandra, A; Berth-Jones, J; Bjerke, J; Christophers, E; Knop, J; Knulst, A C; Morren, M; Morris, A; Reitamo, S; Roed-Petersen, J; Schoepf, E; Thestrup-Pedersen, K; Van Der Valk, P G; Bos, J D

    2001-04-01

    SDZ ASM 981 is a selective inhibitor of the production of pro-inflammatory cytokines from T cells and mast cells in vitro. It is the first ascomycin macrolactam derivative under development for the treatment of inflammatory skin diseases. This study was designed to determine the safety and efficacy of SDZ ASM 981 cream at concentrations of 0.05%, 0.2%, 0.6% and 1.0% in the treatment of patients with atopic dermatitis and to select the concentration to be used in phase III studies. This was a double-blind, randomized, parallel-group, multicentre dose-finding study. A total of 260 patients were randomly assigned to treatment with SDZ ASM 981 cream at concentrations of 0.05%, 0.2%, 0.6%, or 1.0%, matching vehicle cream, or the internal control 0.1% betamethasone-17-valerate cream (BMV). Treatment was given twice daily for up to 3 weeks. A clear dose-response relationship for SDZ ASM 981 was evident, with 0.2%, 0.6% and 1.0% SDZ ASM 981 creams all being significantly more effective than vehicle (P = 0.041, 0.001 and 0.008, respectively) in terms of baseline to end-point changes in the Eczema Area Severity Index (EASI) and pruritus score. The 1.0% cream was the most effective SDZ ASM 981 concentration. BMV was more effective than the SDZ ASM 981 creams tested in this study. It appears that the efficacy plateau was not reached with the SDZ ASM 981 creams within 3 weeks treatment. SDZ ASM 981 was well tolerated. Burning or a feeling of warmth were the only adverse events reported more frequently in the 0.6% and 1.0% SDZ ASM 981 treatment groups than in the vehicle treatment group (42.9%, 48.9% and 34.9%, respectively). Few systemic adverse events were reported during the study (headache was the most frequent systemic event reported by 15 of 252 patients) and none was considered to be related to treatment. The local tolerability profile of the 1.0% cream was similar to that of the lower concentrations. 1.0% SDZ ASM 981 cream, which was shown to be safe, well tolerated and

  10. Optimized PWR power ascension reload testing

    International Nuclear Information System (INIS)

    Emery, S.P.; Long, S.W.; Nazareth, V.F.; Herschthal, M.A.

    1987-01-01

    Reduction in critical path testing time following refueling is actively supported by utilities to increase plant capacity factor and to minimize replacement power costs. Combustion Engineering (C-E) has developed a fast power ascension program (FPAP), which reduces this critical path testing by minimizing holds at intermediate power levels and by automating data acquisition and analysis. A very successful demonstration of the FPAP was performed recently during the cycle 3 startup of Southern California Edison's San Onofre Unit 2 reactor, which resulted in a critical path time savings of ∼ 3 days

  11. HTS power lead testing at the Fermilab magnet test facility

    Energy Technology Data Exchange (ETDEWEB)

    Rabehl, R.; Carcagno, R.; Feher, S.; Huang, Y.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.; /Fermilab

    2005-08-01

    The Fermilab Magnet Test Facility has tested high-temperature superconductor (HTS) power leads for cryogenic feed boxes to be placed at the Large Hadron Collider (LHC) interaction regions and at the new BTeV C0 interaction region of the Fermilab Tevatron. A new test facility was designed and operated, successfully testing 20 pairs of HTS power leads for the LHC and 2 pairs of HTS power leads for the BTeV experiment. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. Results from the facility commissioning are included, as is the performance of a new insulation method to prevent frost accumulation on the warm ends of the power leads.

  12. HTS power lead testing at the Fermilab magnet test facility

    International Nuclear Information System (INIS)

    Rabehl, R.; Carcagno, R.; Feher, S.; Huang, Y.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.

    2005-01-01

    The Fermilab Magnet Test Facility has tested high-temperature superconductor (HTS) power leads for cryogenic feed boxes to be placed at the Large Hadron Collider (LHC) interaction regions and at the new BTeV CO interaction region of the Fermilab Tevatron. A new test facility was designed and operated, successfully testing 20 pairs of HTS power leads for the LHC and 2 pairs of HTS power leads for the BTeV experiment. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. Results from the facility commissioning are included, as is the performance of a new insulation method to prevent frost accumulation on the warm ends of the power leads

  13. SNS EXTRACTION KICKER POWER SUPPLY PROTOTYPE TEST

    International Nuclear Information System (INIS)

    MI, J.L.; SANDBERG, J.; SANDERS, R.; SOUKAS, A.; ZHANG, W.

    2000-01-01

    The SNS (Spallation Neutron Source) accumulator ring Extraction System consists of a Fast kicker and a Lambertson Septum magnet. The proposed design will use 14 kicker magnets powered by an Extraction Kicker Power Supply System. They will eject the high power beam from the SNS accumulator ring into RTBT (Ring to Target Beam Tunnel) through a Lambertson Septum magnet. This paper describes some test results of the SNS Extraction Kicker power supply prototype. The high repetition rate of 60 pulse per second operation is the challenging part of the design. In the prototype testing, a 3 kA damp current of 700ns pulse-width, 200 nS rise time and 60 Hz repetition rate at 32 kV PFN operation voltage has been demonstrated. An Extraction kicker power supply system design diagram is depicted

  14. International Accreditation of ASME Codes and Standards

    International Nuclear Information System (INIS)

    Green, Mervin R.

    1989-01-01

    ASME established a Boiler Code Committee to develop rules for the design, fabrication and inspection of boilers. This year we recognize 75 years of that Code and will publish a history of that 75 years. The first Code and subsequent editions provided for a Code Symbol Stamp or mark which could be affixed by a manufacturer to a newly constructed product to certify that the manufacturer had designed, fabricated and had inspected it in accordance with Code requirements. The purpose of the ASME Mark is to identify those boilers that meet ASME Boiler and Pressure Vessel Code requirements. Through thousands of updates over the years, the Code has been revised to reflect technological advances and changing safety needs. Its scope has been broadened from boilers to include pressure vessels, nuclear components and systems. Proposed revisions to the Code are published for public review and comment four times per year and revisions and interpretations are published annually; it's a living and constantly evolving Code. You and your organizations are a vital part of the feedback system that keeps the Code alive. Because of this dynamic Code, we no longer have columns in newspapers listing boiler explosions. Nevertheless, it has been argued recently that ASME should go further in internationalizing its Code. Specifically, representatives of several countries, have suggested that ASME delegate to them responsibility for Code implementation within their national boundaries. The question is, thus, posed: Has the time come to franchise responsibility for administration of ASME's Code accreditation programs to foreign entities or, perhaps, 'institutes.' And if so, how should this be accomplished?

  15. Tenth ASME wind energy symposium

    International Nuclear Information System (INIS)

    Berg, D.E.; Veers, P.S.

    1991-01-01

    This book contains papers presented at the Fourteenth Annual Energy-Sources Technology Conference and Exhibition. Included are the following papers: Wind Power Farm Site Selection, Turbulence characterization for wind energy development, Effects of insect configuration on wind turbine airfoils, Power fluctuations from horizontal and vertical axis wind turbines, Power regulation by active yaw control for a teetered wind rotor, and economic aspects of wind energy

  16. Approaching application of risk-based inspection to ASME code section XI

    International Nuclear Information System (INIS)

    Hedden, Owen F.

    1995-01-01

    This paper will describe current efforts within the ASME Boiler and Pressure Vessel Committee's Subcommittee on Nuclear Inservice Inspection to introduce risk-based technology to optimize inservice inspection of nuclear power plants. The subcommittee is responsible for the content of ASME Boiler and Pressure Vessel Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. The paper will first provide the historical background for the inspection program currently in Section XI. It will then describe the development of new technology through the ASME Center for Research and Technology Development program. Next, the work now going on in two of the groups under the Section XI committee will be described in detail. Each of these two efforts is directed toward the application of new risk-based inspection technology to nuclear piping systems. Finally, the directions of additional research and applications of the technology will be discussed. (author)

  17. Power plant cable condition monitoring and testing at Georgia Power

    International Nuclear Information System (INIS)

    Champion, T.C.

    1988-01-01

    Georgia Power's Research Center has been heavily involved in the evaluation of electrical insulating materials and cables since its inception more than 17 years ago. For the past ten years that expertise has been applied to cables used in generation plants. This paper discusses the results of two test programs. The first is a quality control inspection on 169 samples of new power generation cables. The second is a material degradation evaluation on four short cable samples removed from a coal fired plant during an equipment upgrade. The new material evaluation was performed to identify the cause of a high failure rate upon initial hi-pot testing of newly installed cables. The material degradation evaluation was performed to evaluate the need for replacement of existing cables during an equipment upgrade. Results of the evaluations have led to development of a detailed proposal for a program to evaluate cable degradation and remaining life for cables used in power generation facilities

  18. High Power Alternator Test Unit (ATU) Electrical System Test

    Science.gov (United States)

    Birchenough, Arthur; Hervol, David

    2007-01-01

    The Alternator Test Unit (ATU) in the Lunar Power System Facility (LPSF) located at the NASA Glenn Research Center (GRC) in Cleveland, OH was used to simulate the operating conditions and evaluate the performance of the ATU and it s interaction with various LPSF components in accordance with the JIMO AC Power System Requirements. The testing was carried out at the breadboard development level. Results of these tests will be used for the development and validation of analytical models for performance and lifetime prediction.

  19. Activated sludge model No. 2d, ASM2d

    DEFF Research Database (Denmark)

    Henze, M.

    1999-01-01

    The Activated Sludge Model No. 2d (ASM2d) presents a model for biological phosphorus removal with simultaneous nitrification-denitrification in activated sludge systems. ASM2d is based on ASM2 and is expanded to include the denitrifying activity of the phosphorus accumulating organisms (PAOs......). This extension of ASM2 allows for improved modeling of the processes, especially with respect to the dynamics of nitrate and phosphate. (C) 1999 IAWQ Published by Elsevier Science Ltd. All rights reserved....

  20. Efectos del entrenamiento físico en piscina climatizada sobre la capacidad aeróbica de un grupo de niños asmáticos. (Effects of the physical training in a warm-water pool on the aerobic power of a group of asthmatic children.

    Directory of Open Access Journals (Sweden)

    Juan David Fernández Villada

    2009-07-01

    Full Text Available ResumenObjetivo: Observar cambios en la capacidad aeróbica de un grupo de niños asmáticos después de un entrenamiento físico en piscina climatizada (EFPC. Método: 22 niños asmáticos, se dividieron en grupo experimental (GE y grupo control (GC de 11 cada uno. El GE se sometió a un EFPC durante 18 semanas de manera controlada, mientras el GC solo hacía sus actividades cotidianas. A ambos grupos se les realizó pruebas de esfuerzo en banda rodante para determinar el máximo consumo de oxígeno en METS y la frecuencia cardiaca máxima (FCM. Se midió en cada sesión de EFPC si los niños presentaban asma inducida por esfuerzo (AIE. Resultados: Al iniciar el estudio, el promedio de METS y de la FCM fueron similares en ambos grupos (Anova METS p = 0.5568; Anova FCM p = 0.5198; METS: GE 11.61 ± 2.33 DS, GC 11.05 ± 2.08 DS; FCM: GE 165.09 ± 13.85 DS, GC 161.46 ± 12.12 DS. Al finalizar, sólo el GE mejoró en los promedios de los METS de la diferencia pre-post según la prueba de observaciones apareadas (METS GE 6.65 ± 3.46 DS; Anova p = 0.0006. La FCM también aumentó, siendo significativa en el GE (FCM: GE 22.45 ± 19.04 DS; Anova FCM p = 0.0823. Solo el 6,06% de las mediciones presentaron AIE. Conclusión: Se encontró aumento significativo de la capacidad aeróbica en el GE, comparado con el GC. Se sugiere que utilizar piscina climatizada para el entrenamiento físico del asmático puede disminuir el AIE.Abstract Objective: To observe changes in the aerobic power in a group of asthmatic children after of a physical training in warm-water pool (EFPC. Method: 22 asthmatic children, who were divided in to an experimental group (GE and a control group (GC of 11 each. The GE was submitted to an EFPC for 18 weeks in a controlled way, while the GC only did their daily activities. Treadmill ergometer testing were conducted to both groups, to determine the maximal oxygen consumption in MET and the maximum heart rate (FCM. Each section of the

  1. Qualification of NDT systems for in-service inspections of nuclear power plant pressure vessels

    International Nuclear Information System (INIS)

    Elfving, K.

    1998-11-01

    The goal of this study is to determine the requirements of the in-service inspection qualification in Europe, their feasibility in practice and to find out possible manufacture defects in test pieces used in practical trials. The literature study consists of qualification requirements set by European regulatory bodies and by the European nuclear power utilities. Also a brief summary of qualification requirements set by ASME Code, Section XI and comparison between ASME and European qualification requirements is included

  2. Consideration of the Construction Code for TBM-body in ASME BPVC

    International Nuclear Information System (INIS)

    Kim, Dongjun; Kim, Yunjae; Kim, Suk Kwon; Park, Sung Dae; Lee, Dong Won

    2016-01-01

    In this paper, ASME code is briefly introduced, and the TBM-body is classified for selecting the ASME section. With the classification of TBM-body, the appropriate section is determined. Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) has been designed to research on the functions of breeding blanket by KO TBM team. The functions has three subjects as 1) Tritium breeding, 2) Heat conversion and extraction, and 3) Neutron and Gamma-ray shielding. For the process of design, it is needed to select the appropriate construction code as the design criteria. ITER Organization (IO) has proposed that RCC-MR Edition 2007 ver. shall be used for TBM-shield. Because the TBM-shield is connected to the vacuum boundary. For the other part of TBM-set, TBM-body, there is no constraint on the selected code, and the manufacturer can appropriately select the construction code to apply design and fabrication parts. KO TBM Team has considered whether it is appropriate to choose any code for TBM-body. One of the things is ASME code. The advantage of ASME choice is suitable to the domestic status. In the domestic nuclear plant, ASME or KEPIC code is used as regulatory requirements. Based on this, it is possible to prepare a domestic fusion plant regulatory. In this paper, the construction code of TBM-body was determined in ASME BPVC. For the determination of code, the structure of ASME BPVC was introduced and the classification for TBM-body was conducted by the ITER criteria. And the operation conditions of TBM-body that contained creep and irradiation effects was considered to determine the construction code

  3. Consideration of the Construction Code for TBM-body in ASME BPVC

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dongjun; Kim, Yunjae [Korea Univ., Seoul (Korea, Republic of); Kim, Suk Kwon; Park, Sung Dae; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, ASME code is briefly introduced, and the TBM-body is classified for selecting the ASME section. With the classification of TBM-body, the appropriate section is determined. Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) has been designed to research on the functions of breeding blanket by KO TBM team. The functions has three subjects as 1) Tritium breeding, 2) Heat conversion and extraction, and 3) Neutron and Gamma-ray shielding. For the process of design, it is needed to select the appropriate construction code as the design criteria. ITER Organization (IO) has proposed that RCC-MR Edition 2007 ver. shall be used for TBM-shield. Because the TBM-shield is connected to the vacuum boundary. For the other part of TBM-set, TBM-body, there is no constraint on the selected code, and the manufacturer can appropriately select the construction code to apply design and fabrication parts. KO TBM Team has considered whether it is appropriate to choose any code for TBM-body. One of the things is ASME code. The advantage of ASME choice is suitable to the domestic status. In the domestic nuclear plant, ASME or KEPIC code is used as regulatory requirements. Based on this, it is possible to prepare a domestic fusion plant regulatory. In this paper, the construction code of TBM-body was determined in ASME BPVC. For the determination of code, the structure of ASME BPVC was introduced and the classification for TBM-body was conducted by the ITER criteria. And the operation conditions of TBM-body that contained creep and irradiation effects was considered to determine the construction code.

  4. ASME Section XI philosophy related to operating nuclear plant fatigue damage protection

    International Nuclear Information System (INIS)

    Gosselin, S.R.

    1995-01-01

    When faced with operating fatigue concerns, nuclear plants traditionally look to the requirements contained in the original construction design code, ASME Section 3, as a basis for component operability. These rules constitute the requirements for nuclear power plant vessel and component construction and, when combined with the Owner's Design Specification, provide reasonable assurance of reliable operation. However, once construction is complete and operation begins, the purpose of any subsequent evaluations shifts from component ''design qualification'' to component ''fitness for service.'' This is a role that has been assumed for ASME Section 11. This paper presents a philosophy, recently endorsed by the ASME Section 11 Executive Committee, intended to guide future Code activities regarding fatigue and its impact on component serviceability

  5. Prediction of surface cracks from thick-walled pressurized vessels with ASME code

    International Nuclear Information System (INIS)

    Thieme, W.

    1983-01-01

    The ASME-Code, Section XI, Appendix A 'Analysis of flow indications' is still non-mandatory for the pressure components of nuclear power plants. It is certainly difficult to take realistic account of the many factors influencing crack propagation while making life predictions. The accuracy of the US guideline is analysed, and its possible applications are roughly outlined. (orig./IHOE) [de

  6. A test of reproductive power in snakes.

    Science.gov (United States)

    Boback, Scott M; Guyer, Craig

    2008-05-01

    Reproductive power is a contentious concept among ecologists, and the model has been criticized on theoretical and empirical grounds. Despite these criticisms, the model has successfully predicted the modal (optimal) size in three large taxonomic groups and the shape of the body size distribution in two of these groups. We tested the reproductive power model on snakes, a group that differs markedly in physiology, foraging ecology, and body shape from the endothermic groups upon which the model was derived. Using detailed field data from the published literature, snake-specific constants associated with reproductive power were determined using allometric relationships of energy invested annually in egg production and population productivity. The resultant model accurately predicted the mode and left side of the size distribution for snakes but failed to predict the right side of that distribution. If the model correctly describes what is possible in snakes, observed size diversity is limited, especially in the largest size classes.

  7. Proceedings: 2001 ASME/EPRI Radwaste Workshop

    International Nuclear Information System (INIS)

    2001-01-01

    Nuclear utilities continually evaluate methods to improve operations and reduce costs associated with radioactive waste management. The continuing deregulation process has increased the emphasis on this activity. The Annual ASME/EPRI Workshop facilitates this effort by communicating technology and management improvements throughout the industry. This workshop, restricted to utility radwaste professionals, also serves to communicate practical in-plant improvements with the opportunity to discuss them in detail

  8. NIST/ASME Steam Properties Database

    Science.gov (United States)

    SRD 10 NIST/ASME Steam Properties Database (PC database for purchase)   Based upon the International Association for the Properties of Water and Steam (IAPWS) 1995 formulation for the thermodynamic properties of water and the most recent IAPWS formulations for transport and other properties, this updated version provides water properties over a wide range of conditions according to the accepted international standards.

  9. Proceedings: 2000 ASME/EPRI Radwaste Workshop

    International Nuclear Information System (INIS)

    2001-01-01

    Nuclear utilities are continually evaluating methods to improve operations and reduce costs associated with radioactive waste management. The continuing deregulation process has added increased emphasis to this activity. The Annual ASME/EPRI Workshop facilitates this effort by communicating technological and managerial improvements throughout the industry. This workshop, restricted to utility radwaste professionals, also serves to communicate practical in-plant improvements with the opportunity to discuss them in detail

  10. Technical basis for the extension of ASME Code Case N-494 for assessment of austenitic piping

    International Nuclear Information System (INIS)

    Bloom, J.M.

    1995-01-01

    In 1990, the ASME Boiler and Pressure Vessel Code for Nuclear Components approved Code Case N-494 as an alternative procedure for evaluating laws in Light Water Reactor alterative procedure for evaluating flaws in Light Water Reactor (LWR) ferritic piping. The approach is an alternative to Appendix H of the ASME Code and alloys the user to remove some unnecessary conservatism in the existing procedure by allowing the use of pipe specific material properties. The Code Case is an implementation of the methodology of the Deformation Plasticity Failure Assessment diagram (DPFAD). The key ingredient in the application of DPFAD is that the material stress-strain curve must be in the format of a simple power law hardening stress-strain curve such as the Ramberg-Osgood (R-O) model. Ferritic materials can be accurately fit by the R-O model and, therefore, it was natural to use the DPFAD methodology for the assessment of LWR ferritic piping. An extension of Code Case N-494 to austenitic piping required a modification of the existing DPFAD methodology. The Code Case N-494 approach was revised using the PWFAD procedure in the same manner as in the development of the original N-494 approach for ferritic materials. A lower bound stress-strain curve was used to generate a PWFAD curve for the geometry of a part-through wall circumferential flaw in a cylinder under tension. Earlier work demonstrated that a cylinder under axial tension with a 50% flaw depth, 90 degrees in circumference, and radius to thickness of 10, produced a lower bound FAD curve. Validation of the new proposed Code Case procedure for austenitic piping was performed using actual pipe test data. Using the lower bound PWFAD curve, pipe test results were conservatively predicted. The resultant development of ht PWFAD curve for austenitic piping led to a revision of Code Case N-494 to include a procedure for assessment of flaws in austenitic piping

  11. Equilíbrio corporal em crianças e adolescentes asmáticos e não asmáticos

    Directory of Open Access Journals (Sweden)

    Marta Cristina Rodrigues da Silva

    2013-06-01

    Full Text Available O objetivo foi analisar e comparar o equilíbrio corporal em crianças e adolescentes asmáticos e não asmáticos. Fizeram parte do grupo de estudos 24 sujeitos com idades de 7 a 14 anos divididos em dois grupos: grupo asmático e grupo controle. Para avaliação do equilíbrio corporal utilizou-se uma plataforma de força. Foram utilizadas as condições, olhos abertos e fechados com três tentativas aleatórias, com duração de 30 segundos cada uma. Os resultados apontaram diferença significativa entre os grupos, no teste de equilíbrio com olhos abertos apresentando maior amplitude de deslocamento na direção ântero-posterior (COPap (p = 0,04, e médio lateral (COPml (p = 0,02 no grupo asmático. Enquanto que no teste com olhos fechados a diferença foi significante apenas na amplitude de deslocamento ântero-posterior (COPap (p = 0,02 e Área de Elipse (p=0,03. Desse modo, a asma com suas limitações e consequências parece influenciar negativamente no equilíbrio corporal de seus portadores quando comparados com crianças sem a patologia e da mesma faixa etária.

  12. Feature extraction for face recognition via Active Shape Model (ASM) and Active Appearance Model (AAM)

    Science.gov (United States)

    Iqtait, M.; Mohamad, F. S.; Mamat, M.

    2018-03-01

    Biometric is a pattern recognition system which is used for automatic recognition of persons based on characteristics and features of an individual. Face recognition with high recognition rate is still a challenging task and usually accomplished in three phases consisting of face detection, feature extraction, and expression classification. Precise and strong location of trait point is a complicated and difficult issue in face recognition. Cootes proposed a Multi Resolution Active Shape Models (ASM) algorithm, which could extract specified shape accurately and efficiently. Furthermore, as the improvement of ASM, Active Appearance Models algorithm (AAM) is proposed to extracts both shape and texture of specified object simultaneously. In this paper we give more details about the two algorithms and give the results of experiments, testing their performance on one dataset of faces. We found that the ASM is faster and gains more accurate trait point location than the AAM, but the AAM gains a better match to the texture.

  13. Risk evaluation for motor operated valves in an Inservice Testing Program at a PWR nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    Li, Y.C.; Chen, K.T.; Su, Y.L.; Ting, K.; Chien, F.T.; Li, G.D.; Huang, S.H.

    2012-01-01

    Safety related valves such as Motor Operated Valves (MOV), Air Operated Valves (AOV) or Check Valves (CV) play an important role in nuclear power plant. Functioning of these valves mainly aim at emergency reactivity control, post-accident residue heat removal, post-accident radioactivity removal and containment isolation when a design basis accident occurred. In order to maintain these valves under operable conditions, an Inservice Testing Program (IST) is defined for routine testing tasks based on the ASME Boiler and Pressure Vessel Code section XI code requirements. Risk based Inservice Testing Programs have been studied and developed extensively in the nuclear energy industry since the 1990s. Risk Based evaluations of IST can bring positive advantages to the licensee such as identifying the vulnerability of the system, reducing unnecessary testing burden, concentrating testing resources on the critical pass oriented valves and saving plant’s personnel dose exposure. This risk evaluation is incorporated with quantitative and qualitative analyses to the Motor Operated Valves under current Inservice Testing Program for PWR nuclear power plant in Taiwan. With the outcome of the risk classifications for the safety related MOVs through numerical or deterministic analyses, a risk based testing frequency relief is suggested to demonstrate the benefits received from the risk based Inservice Testing Program. The goal made of this study, it could be as a reference and cornerstone for the licensee to perform overall scope Risk-Informed Inservice Testing Program (RI-IST) evaluation by referring relevant methodologies established in this study.

  14. ASM Based Synthesis of Handwritten Arabic Text Pages

    Directory of Open Access Journals (Sweden)

    Laslo Dinges

    2015-01-01

    Full Text Available Document analysis tasks, as text recognition, word spotting, or segmentation, are highly dependent on comprehensive and suitable databases for training and validation. However their generation is expensive in sense of labor and time. As a matter of fact, there is a lack of such databases, which complicates research and development. This is especially true for the case of Arabic handwriting recognition, that involves different preprocessing, segmentation, and recognition methods, which have individual demands on samples and ground truth. To bypass this problem, we present an efficient system that automatically turns Arabic Unicode text into synthetic images of handwritten documents and detailed ground truth. Active Shape Models (ASMs based on 28046 online samples were used for character synthesis and statistical properties were extracted from the IESK-arDB database to simulate baselines and word slant or skew. In the synthesis step ASM based representations are composed to words and text pages, smoothed by B-Spline interpolation and rendered considering writing speed and pen characteristics. Finally, we use the synthetic data to validate a segmentation method. An experimental comparison with the IESK-arDB database encourages to train and test document analysis related methods on synthetic samples, whenever no sufficient natural ground truthed data is available.

  15. ASM Based Synthesis of Handwritten Arabic Text Pages.

    Science.gov (United States)

    Dinges, Laslo; Al-Hamadi, Ayoub; Elzobi, Moftah; El-Etriby, Sherif; Ghoneim, Ahmed

    2015-01-01

    Document analysis tasks, as text recognition, word spotting, or segmentation, are highly dependent on comprehensive and suitable databases for training and validation. However their generation is expensive in sense of labor and time. As a matter of fact, there is a lack of such databases, which complicates research and development. This is especially true for the case of Arabic handwriting recognition, that involves different preprocessing, segmentation, and recognition methods, which have individual demands on samples and ground truth. To bypass this problem, we present an efficient system that automatically turns Arabic Unicode text into synthetic images of handwritten documents and detailed ground truth. Active Shape Models (ASMs) based on 28046 online samples were used for character synthesis and statistical properties were extracted from the IESK-arDB database to simulate baselines and word slant or skew. In the synthesis step ASM based representations are composed to words and text pages, smoothed by B-Spline interpolation and rendered considering writing speed and pen characteristics. Finally, we use the synthetic data to validate a segmentation method. An experimental comparison with the IESK-arDB database encourages to train and test document analysis related methods on synthetic samples, whenever no sufficient natural ground truthed data is available.

  16. Power Tillers for Demining: Blast Test

    Directory of Open Access Journals (Sweden)

    Emanuela Elisa Cepolina

    2007-06-01

    Full Text Available Power tillers are very simple and versatile machines with large scale diffusion in developing countries, where they are commonly used both for agriculture and for transportation purposes. A new integrated participatory approach that makes use of and improves local end-users knowledge has been used to design a new robotic system for humanitarian demining applications in Sri Lanka, using power tiller as core module. A demining machine composed by a tractor unit, a ground processing tool and a vegetation cutting tool is here presented together with results obtained from the first blast test on the preliminary version of tractor unit armouring. Different breakable connections between wheels and axle have been designed to cause physical detachment and interrupt the transmission of the shock wave released by the explosion of a mine under one wheel. Effects of explosions on different types of wheels and on the chassis have been recorded and commented.

  17. Power Tillers for Demining: Blast Test

    Directory of Open Access Journals (Sweden)

    Emanuela Elisa Cepolina

    2008-11-01

    Full Text Available Power tillers are very simple and versatile machines with large scale diffusion in developing countries, where they are commonly used both for agriculture and for transportation purposes. A new integrated participatory approach that makes use of and improves local end-users knowledge has been used to design a new robotic system for humanitarian demining applications in Sri Lanka, using power tiller as core module. A demining machine composed by a tractor unit, a ground processing tool and a vegetation cutting tool is here presented together with results obtained from the first blast test on the preliminary version of tractor unit armouring. Different breakable connections between wheels and axle have been designed to cause physical detachment and interrupt the transmission of the shock wave released by the explosion of a mine under one wheel. Effects of explosions on different types of wheels and on the chassis have been recorded and commented.

  18. Temporal Deductive Verification of Basic ASM Models

    OpenAIRE

    Daho, Hocine El-Habib; University of Oran; Benhamamouch, Djillali; University of Oran

    2010-01-01

    Abstract State Machines (ASMs, for short) provide a practical new computational model which has been applied in the area of software engineering for systems design and analysis. However, reasoning about ASM models occurs, not within a formal deductive system, but basically in the classical informal proofs style of mathematics. Several formal verification approaches for proving correctness of ASM models have been investigated. In this paper we consider the use of the TLA+logic for the deductive...

  19. Qualification of NDT systems for in-service inspections of nuclear power plant pressure vessels; Ydinvoimalaitosten paineastioiden maeaeraeaikaistarkastuksissa kaeytettaevien NDT-jaerjestelmien paetevoeinti

    Energy Technology Data Exchange (ETDEWEB)

    Elfving, K

    1998-11-01

    The goal of this study is to determine the requirements of the in-service inspection qualification in Europe, their feasibility in practice and to find out possible manufacture defects in test pieces used in practical trials. The literature study consists of qualification requirements set by European regulatory bodies and by the European nuclear power utilities. Also a brief summary of qualification requirements set by ASME Code, Section XI and comparison between ASME and European qualification requirements is included 24 refs.

  20. Technical Challenge and Demonstration of Advanced Solution Monitoring and Measurement System (ASMS)

    International Nuclear Information System (INIS)

    Takaya, A.; Mukai, Y.; Nakamura, H.; Hosoma, T.; Yoshimoto, K.; Tamura, T.; Iwamoto, T.

    2010-01-01

    JNFL and JAEA have collaboratively started to develop an Advanced Solution Measurement and monitoring System (ASMS) as a part of technical challenge intended for next generation safeguards NDA equipment. After we completed feasibility study by using small detectors, the second stage of ASMS has installed into PCDF tank located in a cell, and then tested and calibrated by Pu nitrate solution experimentally. There was no experience measuring around 50kg Pu inventory directly, so it was very challenging work. The conventional SMMS (Solution Monitoring and Measurement System) that is composed of precision manometers acquires density, level and temperature of solution, so that the sampling and analysis are essential to obtain the nuclear material amount in the tank. The SMMS has two weak points on verification and monitoring of the nuclear material flow and inventory; (1) Direct measurement of the inventory cannot be done, (2) Solution rework and reagent adjustment operation in actual plant will make miss-interpretation on the monitoring evaluation. The purpose of ASMS development is to establish quantitative plutonium mass measurement technique directly by NDA of high concentrated pure plutonium nitrate solution and monitoring capability for solution transfers in a process. The merits of ASMS are considered below; (1) Provide direct Pu measurement and continuous monitoring capability, (2) Eliminate sampling and analysis at IIV, (3) Reduce unmeasured inventory. The target of the measurement uncertainty of ASMS is set less than 6% (1sigma) which is equivalent to meet the detection level of the partial defect at IIV by NDA. Known-alpha coincidence counting technique is applied to the ASMS, which is similar to the NDAs for MOX powder as a principle measurement technique. Especially, three following points are key techniques to establish ASMS. (1) Pre-determination of plutonium isotopic composition because it impacts alpha and rho-zero values to obtain multiplication

  1. ASME Section XI trends in developing nuclear codes and standards

    International Nuclear Information System (INIS)

    Hedden, O.F.

    1995-01-01

    When the author began working on nuclear power many years ago, he knew that perfection was the only acceptable technical standard. Unfortunately, this became an obsession with perfection that has had unfavorable consequences in some of the non-technical areas of work in ASME nuclear power Codes and Standards. However, the economic problems of the nuclear power industry now demand a more pragmatic approach if the industry is to continue. Not only does each item considered for action need to be evaluated to criteria that may in some cases be less than perfection, but one needs to consider whether it contributes tangibly to either safety or to reduction in technical or administrative burden. These should be the governing, criteria. The introduction of risk-based inspection methodologies will certainly be an important element in doing this successfully. One needs to consider these criteria collectively, as one discusses each item at the committee level, and individually, as one votes on each item. In the past, the author has been concerned that the industry was not acting quickly enough in taking advantage of opportunities offered by the Code to increase safety or to reduce cost. While he still has some concern, he thinks communication channels have been greatly improved. Now he is becoming more concerned with both the collective and individual actions that delay beneficial changes. The second part of the author's talk has to do with the relevance of the code committees in the nuclear power industry regulatory process

  2. AsmL Specification of a Ptolemy II Scheduler

    DEFF Research Database (Denmark)

    Lázaro Cuadrado, Daniel; Koch, Peter; Ravn, Anders Peter

    2003-01-01

    Ptolemy II is a tool that combines different computational models for simulation and design of embedded systems. AsmL is a software specification language based on the Abstract State Machine formalism. This paper reports on development of an AsmL model of the Synchronous Dataflow domain scheduler...

  3. Risk informed in-service inspection and testing in Spain

    International Nuclear Information System (INIS)

    Bros, Juan; Marcelles, Ignacio

    2002-01-01

    The Spanish nuclear regulatory authority, the Consejo de Seguridad Nuclear (CSN), requires the use of codes and standards in force in the country of origin of the plant technology. For this reason, the in-service inspection and testing programs applied at Spanish nuclear power plants basically adhere to the requirements of the ASME XI and ASME OM Codes. It is not surprising that when the earliest developments aimed at drawing up risk informed inservice inspection and testing programs were initiated within the framework of ASME, the Spanish industry should follows such developments very closely. In fact, persons within the Spanish nuclear industry joined different ASME committees involved in the development and approval of the various code cases encompassing these developments. Developments specific to the Spanish nuclear power plants were initiated at a time when the aforementioned reference documentation was in a very advanced stage of development/approval. Two clearly differentiated lines of work got under way: On the one hand, and as regards risk informed in-service testing programs, the American standards were used as the sole reference. In the case of risk informed in-service inspection programs, the Spanish nuclear power plant-owning utilities and the Consejo de Seguridad Nuclear decided to draw up a Spanish guideline that, although using the ASME developments as a reference, would have its own specific characteristics. In relation to the above, and referring to the chronology of the events, the activities performed to date in Spain are described

  4. High Power Fiber Laser Test Bed

    Data.gov (United States)

    Federal Laboratory Consortium — This facility, unique within DoD, power-combines numerous cutting-edge fiber-coupled laser diode modules (FCLDM) to integrate pumping of high power rare earth-doped...

  5. Reload Startup Physics Tests for Tianwan Nuclear Power station

    International Nuclear Information System (INIS)

    Yang Xiaoqiang; Li Wenshuang; Li Youyi; Yao Jinguo; Li Zaipeng Jiangsu

    2010-01-01

    This paper briefly describes the test purposes, test items, test schedules and test equipment's for reload startup physics test's on Unit 1 and 2 of Tianwan Nuclear Power station. Then, an overview of the previous thrice tests and evaluations on the tests results are presented. In the end, the paper shows the development and work direction of optimization project for reload startup physics tests on Unit 1 and 2 of Tianwan Nuclear Power station. (Authors)

  6. FY16 ASME High Temperature Code Activities

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, M. J. [Chromtech Inc., Oak Ridge, TN (United States); Jetter, R. I. [R. I Jetter Consulting, Pebble Beach, CA (United States); Sham, T. -L. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-01

    One of the objectives of the ASME high temperature Code activities is to develop and validate both improvements and the basic features of Section III, Division 5, Subsection HB, Subpart B (HBB). The overall scope of this task is to develop a computer program to be used to assess whether or not a specific component under specified loading conditions will satisfy the elevated temperature design requirements for Class A components in Section III, Division 5, Subsection HB, Subpart B (HBB). There are many features and alternative paths of varying complexity in HBB. The initial focus of this task is a basic path through the various options for a single reference material, 316H stainless steel. However, the program will be structured for eventual incorporation all the features and permitted materials of HBB. Since this task has recently been initiated, this report focuses on the description of the initial path forward and an overall description of the approach to computer program development.

  7. Design specifications for ASME B and PV Code Section III nuclear class 1 piping

    International Nuclear Information System (INIS)

    Richardson, J.A.

    1978-01-01

    ASME B and PV Code Section III code regulations for nuclear piping requires that a comprehensive Design Specification be developed for ensuring that the design and installation of the piping meets all code requirements. The intent of this paper is to describe the code requirements, discuss the implementation of these requirements in a typical Class 1 piping design specification, and to report on recent piping failures in operating light water nuclear power plants in the US. (author)

  8. A Study on Test Technology to Diagnose the Power Apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K H; Kang, Y S; Jeon, Y K; Lee, W Y; Kang, D S; Kyu, H S; Sun, J H; Jo, K H [Korea Electrotechnology Research Institute (Korea, Republic of); Jung, J S; Mun, Y T; Lee, K H; Jung, E H; Kim, J H [Korea Water Resources Corporation (Korea, Republic of)

    1997-02-01

    In this study, we have educated KOWACO(Korea Water Resources Corporation) specialists about the insulation diagnostic technology and trained them the insulation diagnostic test and estimation method of power apparatus. The main results of this study are as follows; A. Education of basic high-voltage engineering. B. Research of insulation characteristic and deterioration mechanism in power apparatus C. Discussion on high-voltage test standard specifications. D. Study on insulation deterioration diagnostics in power apparatus. E. Field testing of insulation diagnosis in power apparatus. F. Engineering of insulation diagnostic testing apparatus to diagnose power apparatus. KOWACO specialists are able to diagnose insulation diagnostic test of power apparatus through this study. As they have instruments to diagnose power apparatus, they can do the test and estimation of the power apparatus insulation diagnosis. (author). refs., figs., tabs.

  9. Testing power system controllers by real-time simulation

    DEFF Research Database (Denmark)

    Venne, Philippe; Guillaud, Xavier; Sirois, Frédéric

    2007-01-01

    In this paper, we present a number of state-of-the art methods for testing power system controllers based on the use of a real-time power system simulator. After introducing Hypersim, we list and discuss the different means of connection between the controller under tests and the power system...

  10. Hacking and penetration testing with low power devices

    CERN Document Server

    Polstra, Philip

    2014-01-01

    Hacking and Penetration Testing with Low Power Devices shows you how to perform penetration tests using small, low-powered devices that are easily hidden and may be battery-powered. It shows how to use an army of devices, costing less than you might spend on a laptop, from distances of a mile or more. Hacking and Penetration Testing with Low Power Devices shows how to use devices running a version of The Deck, a full-featured penetration testing and forensics Linux distribution, and can run for days or weeks on batteries due to their low power consumption. Author Philip Polstra shows how to

  11. New facility for testing LHC HTS power leads

    CERN Document Server

    Rabehl, Roger Jon; Fehér, S; Huang, Y; Orris, D; Pischalnikov, Y; Sylvester, C D; Tartaglia, M

    2005-01-01

    A new facility for testing HTS power leads at the Fermilab Magnet Test Facility has been designed and operated. The facility has successfully tested 19 pairs of HTS power leads, which are to be integrated into the Large Hadron Collider Interaction Region cryogenic feed boxes. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. HTS power lead test results from the commissioning phase of the project are also presented.

  12. Correlations between power and test reactor data bases

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Simonen, E.P.

    1989-02-01

    Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

  13. Accelerator System Model (ASM) user manual with physics and engineering model documentation. ASM version 1.0

    International Nuclear Information System (INIS)

    1993-07-01

    The Accelerator System Model (ASM) is a computer program developed to model proton radiofrequency accelerators and to carry out system level trade studies. The ASM FORTRAN subroutines are incorporated into an intuitive graphical user interface which provides for the open-quotes constructionclose quotes of the accelerator in a window on the computer screen. The interface is based on the Shell for Particle Accelerator Related Codes (SPARC) software technology written for the Macintosh operating system in the C programming language. This User Manual describes the operation and use of the ASM application within the SPARC interface. The Appendix provides a detailed description of the physics and engineering models used in ASM. ASM Version 1.0 is joint project of G. H. Gillespie Associates, Inc. and the Accelerator Technology (AT) Division of the Los Alamos National Laboratory. Neither the ASM Version 1.0 software nor this ASM Documentation may be reproduced without the expressed written consent of both the Los Alamos National Laboratory and G. H. Gillespie Associates, Inc

  14. Accelerator System Model (ASM) user manual with physics and engineering model documentation. ASM version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-07-01

    The Accelerator System Model (ASM) is a computer program developed to model proton radiofrequency accelerators and to carry out system level trade studies. The ASM FORTRAN subroutines are incorporated into an intuitive graphical user interface which provides for the {open_quotes}construction{close_quotes} of the accelerator in a window on the computer screen. The interface is based on the Shell for Particle Accelerator Related Codes (SPARC) software technology written for the Macintosh operating system in the C programming language. This User Manual describes the operation and use of the ASM application within the SPARC interface. The Appendix provides a detailed description of the physics and engineering models used in ASM. ASM Version 1.0 is joint project of G. H. Gillespie Associates, Inc. and the Accelerator Technology (AT) Division of the Los Alamos National Laboratory. Neither the ASM Version 1.0 software nor this ASM Documentation may be reproduced without the expressed written consent of both the Los Alamos National Laboratory and G. H. Gillespie Associates, Inc.

  15. Test report light duty utility arm power distribution system (PDS)

    International Nuclear Information System (INIS)

    Clark, D.A.

    1996-01-01

    The Light Duty Utility Arm (LDUA) Power Distribution System has completed vendor and post-delivery acceptance testing. The Power Distribution System has been found to be acceptable and is now ready for integration with the overall LDUA system

  16. Estimation of sample size and testing power (part 5).

    Science.gov (United States)

    Hu, Liang-ping; Bao, Xiao-lei; Guan, Xue; Zhou, Shi-guo

    2012-02-01

    Estimation of sample size and testing power is an important component of research design. This article introduced methods for sample size and testing power estimation of difference test for quantitative and qualitative data with the single-group design, the paired design or the crossover design. To be specific, this article introduced formulas for sample size and testing power estimation of difference test for quantitative and qualitative data with the above three designs, the realization based on the formulas and the POWER procedure of SAS software and elaborated it with examples, which will benefit researchers for implementing the repetition principle.

  17. Integrity evaluation for stud female threads on pressure vessel according to ASME code using FEM

    International Nuclear Information System (INIS)

    Kim, Moon Young; Chung, Nam Yong

    2003-01-01

    The extension of design life among power plants is increasingly becoming a world-wide trend. Kori no.1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts for man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety for helical-coil method which is used according to code case-N-496-1. From analysis results, we found that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10% by manufacture integrity criteria and 2/3 yield strength criteria on ASME code. It was also confirmed that the helical-coil repair method would be safe

  18. Initial tests of thermoacoustic space power engine

    International Nuclear Information System (INIS)

    Backhaus, S.N.

    2002-01-01

    Future NASA deep-space missions will require radioisotope-powered electric generators that are just as reliable as current RTGs, but more efficient and of higher specific power (Wikg). Thennoacoustic engines at the -1-kW scale have converted high-temperature heat into acoustic, or PV, power without moving parts at 30% efficiency. Consisting of only tubes and a few heat exchangers, thennoacoustic engines are low mass and promise to be highly reliable. Coupling a thennoacoustic engine to a low mass, highly reliable and efficient linear alternator will create a heat-driven electric generator suitable for deep-space applications. Conversion efficiency data will be presented on a demonstration thennoacoustic engine designed for the 1 00-Watt power range.

  19. Gingin High Optical Power Test Facility

    International Nuclear Information System (INIS)

    Zhao, C; Blair, D G; Barrigo, P

    2006-01-01

    The Australian Consortium for Gravitational Wave Astronomy (ACIGA) in collaboration with LIGO is developing a high optical power research facility at the AIGO site, Gingin, Western Australia. Research at the facility will provide solutions to the problems that advanced gravitational wave detectors will encounter with extremely high optical power. The problems include thermal lensing and parametric instabilities. This article will present the status of the facility and the plan for the future experiments

  20. Bootstrap Power of Time Series Goodness of fit tests

    Directory of Open Access Journals (Sweden)

    Sohail Chand

    2013-10-01

    Full Text Available In this article, we looked at power of various versions of Box and Pierce statistic and Cramer von Mises test. An extensive simulation study has been conducted to compare the power of these tests. Algorithms have been provided for the power calculations and comparison has also been made between the semi parametric bootstrap methods used for time series. Results show that Box-Pierce statistic and its various versions have good power against linear time series models but poor power against non linear models while situation reverses for Cramer von Mises test. Moreover, we found that dynamic bootstrap method is better than xed design bootstrap method.

  1. Auriculoterapia en pacientes asmáticos

    Directory of Open Access Journals (Sweden)

    Adolfo González Salvador

    1997-04-01

    Full Text Available Se realiza un estudio para evaluar la eficacia de la auriculopuntura en 30 asmáticos del área de salud de Aguada de Pasajeros, durante los meses de noviembre de 1992 a abril de 1993. El tratamiento se aplicó durante un mes, con seguimiento durante los 5 meses posteriores. Se observó una disminución en la frecuencia, intensidad y duración de las crisis de asma; la mayoría de los pacientes tuvo una evolución satisfactoria y no se presentaron complicaciones. Se concluye que la auriculoterapia es un método útil en pacientes con asma bronquial debido a su eficacia e inocuidadA study was conducted to evaluate the efficacy of auriculopuncture in 30 asthmatic patients from the health area of Aguada de Pasajeros between November, 1992, and April, 1993. The treatment was applied for a month, with a follow-up during the next 5 months. It was observed a reduction in the frequency, intensity and duration of the asthma crises. Most of the patients had a satisfactory evolution and there were no complications. It is concluded that auriculotherapy is a useful method for patients with bronchial asthma due to its effectiveness and innocuousness.

  2. Ultrasonic Testing

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong Jun; Kuk, Jeong Han

    2002-02-15

    This book introduces ultrasonic testing, which tells of outline of ultrasonic testing, principle of ultrasonic testing, prosperities of ultrasonic waves, radiographic test and ultrasonic test, basic theory on ultrasonic testing, mode conversion, transmission and diffraction, ultrasonic flaw detection and probe, standard test piece and reference test piece, like KS(JIS) ASME and ASTM, classification and properties of ultrasonic testing, straight beam method, angle beam method, ASME SEC.V.Art.5 ASTMA 388 and KS B 0817 Korean industrial standard.

  3. Lewis Research Center space station electric power system test facilities

    Science.gov (United States)

    Birchenough, Arthur G.; Martin, Donald F.

    1988-01-01

    NASA Lewis Research Center facilities were developed to support testing of the Space Station Electric Power System. The capabilities and plans for these facilities are described. The three facilities which are required in the Phase C/D testing, the Power Systems Facility, the Space Power Facility, and the EPS Simulation Lab, are described in detail. The responsibilities of NASA Lewis and outside groups in conducting tests are also discussed.

  4. Program for analyzing power boost tests

    International Nuclear Information System (INIS)

    Wills, C.A.

    1982-03-01

    A rapid increase of power in a reactor produces a failure in the fuel. Experiments to study the conditions in the NRU reactor after such failures have been planned and carried out. Given the concentrations of specified isotopes at a number of times over the length of an experiment as produced for example, from the program SARGS and the power history of the reactor, this program calculates the release rates, escape rate coefficients, and fractional releases for the isotopes. These values may be optionally printed and plotted. Decay schemes for a limited number of mass numbers are implemented. The program is written in FORTRAN and runs on the CDC 6600 - CYBER 170 system

  5. Energy Systems Test Area (ESTA) Electrical Power Systems Test Operations: User Test Planning Guide

    Science.gov (United States)

    Salinas, Michael J.

    2012-01-01

    Test process, milestones and inputs are unknowns to first-time users of the ESTA Electrical Power Systems Test Laboratory. The User Test Planning Guide aids in establishing expectations for both NASA and non-NASA facility customers. The potential audience for this guide includes both internal and commercial spaceflight hardware/software developers. It is intended to assist their test engineering personnel in test planning and execution. Material covered includes a roadmap of the test process, roles and responsibilities of facility and user, major milestones, facility capabilities, and inputs required by the facility. Samples of deliverables, test article interfaces, and inputs necessary to define test scope, cost, and schedule are included as an appendix to the guide.

  6. Mechanical properties of JT-60 tokamak machine in power tests

    International Nuclear Information System (INIS)

    Takatsu, Hideyuki; Ohkubo, Minoru; Yamamoto, Masahiro; Ohta, Mitsuru

    1986-01-01

    JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. The coil current was raised step by step, monitoring the mechanical, thermal, electrical and vacuum data. Power tests were concluded with successful results. All of the coil systems were raised up to full power operation in combination and system performance was verified including the structural integrity of tokamak machine. Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic forces were supported as expected in the design. A few limitations to machine operation was made clear quantitatively. And it was found that existing detectors were insufficient to monitor machine integrity and two kinds of detector were proposed to be installed. (author)

  7. Power Testing in Basketball: Current Practice and Future Recommendations.

    Science.gov (United States)

    Wen, Neal; Dalbo, Vincent J; Burgos, Bill; Pyne, David B; Scanlan, Aaron T

    2018-02-01

    Numerous foundational movements performed during basketball are predicated on underlying power-related attributes, including speed, change-of-direction, and jumping. Accordingly, fitness testing batteries for basketball have incorporated an assortment of linear speed tests, change-of-direction tests, and jump tests. However, due to the wide variety of testing options it is difficult for basketball practitioners to select appropriate testing protocols for the assessment of power-related attributes. As a result, there is a need to review the relevant literature to identify game-specific, power-related attributes important in basketball and the most appropriate tests available to assess power-related attributes for basketball practitioners. Therefore, the aims of this review were to: (1) identify essential power-related attributes important in basketball; (2) discuss the suitability of common and novel power-related tests; and (3) provide recommendations for future research and best practice approaches for basketball coaching staff. In this review, we propose a series of novel tests that are more targeted and specific to basketball movements including: (1) 5-m and 10-m linear sprints, (2) Modified Agility T-Test, (3) Change-of-Direction Deficit, (4) lateral bound, (5) Sargent jump, (6) one-step jump, and (7) isometric midthigh pull test. Improved testing of power-related attributes should enable basketball practitioners to develop targeted training plans for enhancing player performance.

  8. Low appendicular skeletal muscle mass (ASM) with limited mobility and poor health outcomes in middle-aged African Americans.

    Science.gov (United States)

    Malmstrom, Theodore K; Miller, Douglas K; Herning, Margaret M; Morley, John E

    2013-09-01

    Recent efforts to provide a consensus definition propose that sarcopenia be considered a clinical syndrome associated with the loss of both skeletal muscle mass and muscle function that occurs with aging. Validation of sarcopenia definitions that include both low muscle mass and poor muscle function is needed. In the population-based African American Health (AAH) study (N = 998 at baseline/wave 1), muscle mass and mobility were evaluated in a clinical testing center in a subsample of N = 319 persons (ages 52-68) at wave 4 (2004). Muscle mass was measured using dual energy x-ray absorptiometry and mobility by a 6-min walk test and 4-m gait walk test. Height corrected appendicular skeletal mass (ASM; 9.0 ± 1.5 in n = 124 males, 8.3 ± 2.2 in n = 195 females) was computed as total lean muscle mass in arms and legs (kilograms) divided by the square of height (meters). Cross-sectional and longitudinal (6-year) associations of low ASM (bottom 25 % AAH sample; ASM with limited mobility (4-m gait walk ≤1 m/s or 6-min walk ASM with limited mobility was associated with IADL difficulties (p = .008) and frailty (p = .040) but not with ADL difficulties or falls in cross-sectional analyses; and with ADL difficulties (p = .022), IADL difficulties (p = .006), frailty (p = .039), and mortality (p = .003) but not with falls in longitudinal analyses adjusted for age and gender. Low ASM alone was marginally associated with mortality (p = .085) but not with other outcomes in cross-sectional or longitudinal analyses. Low ASM with limited mobility is associated with poor health outcomes among late middle-aged African Americans.

  9. An example of a component replacement when applying ASME N509 and ASME N510 to older ventilation systems

    International Nuclear Information System (INIS)

    Arndt, T.E.

    1994-06-01

    This paper presents an example of a component replacement (electric heater) when installed in an older ventilation system that was constructed before the issuance of ASME N509 and N510. Many of the existing ventilation systems at the Hanford Site were designed, fabricated, and installed before the issuance of ASME N509 and N510. Requiring the application of these codes to existing ventilation systems presents challenges to the engineer when design changes are needed. Although it may seem that the application of ASME N509 or N510 may be a hindrance at times, this does not need to occur. Proper preparation at the start of project or design modifications can minimize frustration to the engineer when it is judged that portions of ASME N509 and N510 do not apply in a particular application

  10. An example of a component replacement when applying ASME N509 and ASME N510 to older ventilation systems

    Energy Technology Data Exchange (ETDEWEB)

    Arndt, T.E. [Westinghouse Hanford Company, Richland, WA (United States)

    1995-02-01

    This paper presents an example of a component replacement (electric heater) when installed in an older ventilation system that was constructed before the issuance of ASME N509{sup 1} and N510{sup 2}. Many of the existing ventilation systems at the Hanford Site were designed, fabricated, and installed before the issuance of ASME N509{sup 1} and N510{sup 2}. Requiring the application of these codes to existing ventilation systems presents challenges to the engineer when design changes are needed. Although it may seem that the application of ASME N509{sup 1} or N510{sup 2} may be a hindrance at times, this does not need to occur. Proper preparation at the start of project or design modifications can minimize frustration to the engineer when it is judged that portions of ASME N509{sup 1} and N510{sup 2} do not apply in a particular application.

  11. Overview of Power Quality and Integrated Testing at JSC

    Science.gov (United States)

    Davies, Francis

    2018-01-01

    This presentation describes the basic philosophy behind integrated testing and partially integrated testing. It lists some well known errors in space systems that were or could have been caught during integrated testing. Two examples of integrated testing at the Johnson Space Center (JSC) are mentioned, and then an overview of two test facilities that do power testing (partially integrated testing) at JSC are presented, with information on the capabilities of each. Finally a list of three projects that has problems caught during power quality or Electromagnetic Interference (EMI) testing is presented.

  12. ASME AG-1 Section FC Qualified HEPA Filters; a Particle Loading Comparison - 13435

    International Nuclear Information System (INIS)

    Stillo, Andrew; Ricketts, Craig I.

    2013-01-01

    High Efficiency Particulate Air (HEPA) Filters used to protect personnel, the public and the environment from airborne radioactive materials are designed, manufactured and qualified in accordance with ASME AG-1 Code section FC (HEPA Filters) [1]. The qualification process requires that filters manufactured in accordance with this ASME AG-1 code section must meet several performance requirements. These requirements include performance specifications for resistance to airflow, aerosol penetration, resistance to rough handling, resistance to pressure (includes high humidity and water droplet exposure), resistance to heated air, spot flame resistance and a visual/dimensional inspection. None of these requirements evaluate the particle loading capacity of a HEPA filter design. Concerns, over the particle loading capacity, of the different designs included within the ASME AG-1 section FC code[1], have been voiced in the recent past. Additionally, the ability of a filter to maintain its integrity, if subjected to severe operating conditions such as elevated relative humidity, fog conditions or elevated temperature, after loading in use over long service intervals is also a major concern. Although currently qualified HEPA filter media are likely to have similar loading characteristics when evaluated independently, filter pleat geometry can have a significant impact on the in-situ particle loading capacity of filter packs. Aerosol particle characteristics, such as size and composition, may also have a significant impact on filter loading capacity. Test results comparing filter loading capacities for three different aerosol particles and three different filter pack configurations are reviewed. The information presented represents an empirical performance comparison among the filter designs tested. The results may serve as a basis for further discussion toward the possible development of a particle loading test to be included in the qualification requirements of ASME AG-1

  13. ASME AG-1 Section FC Qualified HEPA Filters; a Particle Loading Comparison - 13435

    Energy Technology Data Exchange (ETDEWEB)

    Stillo, Andrew [Camfil Farr, 1 North Corporate Drive, Riverdale, NJ 07457 (United States); Ricketts, Craig I. [New Mexico State University, Department of Engineering Technology and Surveying Engineering, P.O. Box 30001 MSC 3566, Las Cruces, NM 88003-8001 (United States)

    2013-07-01

    High Efficiency Particulate Air (HEPA) Filters used to protect personnel, the public and the environment from airborne radioactive materials are designed, manufactured and qualified in accordance with ASME AG-1 Code section FC (HEPA Filters) [1]. The qualification process requires that filters manufactured in accordance with this ASME AG-1 code section must meet several performance requirements. These requirements include performance specifications for resistance to airflow, aerosol penetration, resistance to rough handling, resistance to pressure (includes high humidity and water droplet exposure), resistance to heated air, spot flame resistance and a visual/dimensional inspection. None of these requirements evaluate the particle loading capacity of a HEPA filter design. Concerns, over the particle loading capacity, of the different designs included within the ASME AG-1 section FC code[1], have been voiced in the recent past. Additionally, the ability of a filter to maintain its integrity, if subjected to severe operating conditions such as elevated relative humidity, fog conditions or elevated temperature, after loading in use over long service intervals is also a major concern. Although currently qualified HEPA filter media are likely to have similar loading characteristics when evaluated independently, filter pleat geometry can have a significant impact on the in-situ particle loading capacity of filter packs. Aerosol particle characteristics, such as size and composition, may also have a significant impact on filter loading capacity. Test results comparing filter loading capacities for three different aerosol particles and three different filter pack configurations are reviewed. The information presented represents an empirical performance comparison among the filter designs tested. The results may serve as a basis for further discussion toward the possible development of a particle loading test to be included in the qualification requirements of ASME AG-1

  14. Advanced power cycling test for power module with on-line on-state VCE measurement

    DEFF Research Database (Denmark)

    Choi, Ui-min; Trintis, Ionut; Blaabjerg, Frede

    2015-01-01

    module. The proposed concept can perform various stress conditions which is valid in a real mission profile and it is using a real power converter application with small loss. The concept of the proposed test setup is first presented. Then, the on-line on-state collector-emitter voltage VCE measurement......Recent research has made an effort to improve the reliability of power electronic systems to comply with more stringent constraints on cost, safety, predicted lifetime and availability in many applications. For this, studies about failure mechanisms of power electronic components and lifetime...... estimation of power semiconductor devices and capacitors have been done. Accelerated power cycling test is one of the common tests to assess the power device module and develop the lifetime model considering the physics of failure. In this paper, a new advanced power cycling test setup is proposed for power...

  15. Creep-fatigue damage evaluation for SS-316LN (ORNL PLATES): - RCC-MR vs. ASME SEC III - NH

    International Nuclear Information System (INIS)

    Sati, Bhuwan Chandra; Jalaldeen, S.; Velusamy, K.; Selvaraj, P.

    2016-01-01

    Investigations of high temperature tests done on ORNL plate with deformation control loading, under creep-fatigue damage have been presented. The test results with methodology of RCC-MR and ASME-NH life prediction under creep-fatigue loading have been assessed. The stress relaxation effect in calculating the life using RCC-MR under creep-fatigue damage is found to be significant in presence of secondary stress. RCC-MR: 2007 is more realistic number of cycles (predicts 51 number of cycles) as compared to ASME-NH (predicts 312 number of cycles) which is demonstrated by the experimental work (observed 86 numbers of cycles). Between RCC-MR and experimental work, design code seems to be more conservative for life prediction due to creep-fatigue damage. For fatigue damage, the approaches are same and the difference comes from material properties and the starting stress for applying Neuber's rule. ASME approach has the limitation of stress range magnitude. ASME approach predicts lower elastic plus plastic strain for the cases having S* above the linear stress limit. For creep strain and creep damage evaluation, ASME and RCC-MR have different approaches for calculating the stress at the beginning and during the hold period. The RCC-MR takes account of cyclic hardening or softening effects (hardening in the present case of 316 LN) by means of the cyclic stress-strain curve and the benefit of symmetrization effects which are significant for this material. The ASME code neglects these effects and instead relies on an approach based on the isochronous stress-strain curves. (author)

  16. Validity and Reliability of a Medicine Ball Explosive Power Test.

    Science.gov (United States)

    Stockbrugger, Barry A.; Haennel, Robert G.

    2001-01-01

    Evaluated the validity and reliability of a medicine ball throw test to evaluate explosive power. Data on competitive sand volleyball players who performed a medicine ball throw and a standard countermovement jump indicated that the medicine ball throw test was a valid and reliable way to assess explosive power for an analogous total-body movement…

  17. ACIGA's high optical power test facility

    Energy Technology Data Exchange (ETDEWEB)

    Ju, L [School of Physics, University of Western Australia, Perth (Australia); Aoun, M [Computer and Information Science, Edith Cowan University, Perth (Australia); Barriga, P [School of Physics, University of Western Australia, Perth (Australia)] [and others

    2004-03-07

    Advanced laser interferometer detectors utilizing more than 100 W of laser power and with {approx}10{sup 6} W circulating laser power present many technological problems. The Australian Consortium for Interferometric Gravitational Astronomy (ACIGA) is developing a high power research facility in Gingin, north of Perth, Western Australia, which will test techniques for the next generation interferometers. In particular it will test thermal lensing compensation and control strategies for optical cavities in which optical spring effects and parametric instabilities may present major difficulties.

  18. HELIUM-POWERED BALLISTIC TEST APPARATUS

    Data.gov (United States)

    Federal Laboratory Consortium — This research device is used for the initial assessment of new materials and is primarily used for fragmentation testing of yarns, fabric, or laminate samples. The...

  19. Optimization of the powering tests of the LHC superconducting circuits

    CERN Document Server

    Bellesia, B; Denz, R; Fernandez-Robles, C; Pojer, M; Saban, R; Schmidt, R; Solfaroli Camillocci, M; Thiesen, H; Vergara Fernández, A

    2010-01-01

    The Large Hadron Collider has (LHC) 1572 superconducting circuits which are distributed along the eight 3.5 km LHC sectors [1]. Time and resources during the commissioning of the LHC technical systems were mostly consumed by the powering tests of each circuit. The tests consisted in carrying out several powering cycles at different current levels for each superconducting circuit. The Hardware Commissioning Coordination was in charge of planning, following up and piloting the execution of the test program. The first powering test campaign was carried out in summer 2007 for sector 7-8 with an expected duration of 12 weeks. The experience gained during these tests was used by the commissioning team for minimising the duration of the following powering campaigns to comply with the stringent LHC project deadlines. Improvements concerned several areas: strategy, procedures, control tools, automatization, and resource allocation led to an average daily test rate increase from 25 to 200 tests per day. This paper desc...

  20. Testing Rastall's theory using matter power spectrum

    International Nuclear Information System (INIS)

    Batista, C.E.M.; Fabris, J.C.; Daouda, M.H.

    2010-01-01

    Rastall's theory is a modification of the General Relativity theory leading to a different expression for the conservation law in the matter sector compared with the usual one. It has been argued recently that such a theory may have applications to the dark energy problem, since a pressureless fluid may lead to an accelerated universe. In the present work we confront Rastall's theory with the power spectrum data. The results indicate a configuration that essentially reduces Rastall's theory to General Relativity, unless the non-usual conservation law refers to a scalar field, situation where other configurations are eventually possible.

  1. Understanding the Long-Term Spectral Variability of Cygnus X-1 from BATSE and ASM Observations

    Science.gov (United States)

    Zdziarski, Andrzej A.; Poutanen, Juri; Paciesas, William S.; Wen, Linqing; Six, N. Frank (Technical Monitor)

    2002-01-01

    We present a spectral analysis of observations of Cygnus X-1 by the RXTE/ASM (1.5-12 keV) and CGRO/BATSE (20-300 keV), including about 1200 days of simultaneous data. We find a number of correlations between intensities and hardnesses in different energy bands from 1.5 keV to 300 keV. In the hard (low) spectral state, there is a negative correlation between the ASM 1.5-12 keV flux and the hardness at any energy. In the soft (high) spectral state, the ASM flux is positively correlated with the ASM hardness (as previously reported) but uncorrelated with the BATSE hardness. In both spectral states, the BATSE hardness correlates with the flux above 100 keV, while it shows no correlation with the flux in the 20-100 keV range. At the same time, there is clear correlation between the BATSE fluxes below and above 100 keV. In the hard state, most of the variability can be explained by softening the overall spectrum with a pivot at approximately 50 keV. The observations show that there has to be another, independent variability pattern of lower amplitude where the spectral shape does not change when the luminosity changes. In the soft state, the variability is mostly caused by a variable hard (Comptonized) spectral component of a constant shape superimposed on a constant soft blackbody component. These variability patterns are in agreement with the dependence of the rms variability on the photon energy in the two states. We interpret the observed correlations in terms of theoretical Comptonization models. In the hard state, the variability appears to be driven mostly by changing flux in seed photons Comptonized in a hot thermal plasma cloud with an approximately constant power supply. In the soft state, the variability is consistent with flares of hybrid, thermal/nonthermal, plasma with variable power above a stable cold disk. Also, based on broadband pointed observations simultaneous with those of the ASM and BATSE, we find the intrinsic bolometric luminosity increases by a

  2. Power Cycling Test Method for Reliability Assessment of Power Device Modules in Respect to Temperature Stress

    DEFF Research Database (Denmark)

    Choi, Ui-Min; Blaabjerg, Frede; Jørgensen, Søren

    2018-01-01

    Power cycling test is one of the important tasks to investigate the reliability performance of power device modules in respect to temperature stress. From this, it is able to predict the lifetime of a component in power converters. In this paper, representative power cycling test circuits......, measurement circuits of wear-out failure indicators as well as measurement strategies for different power cycling test circuits are discussed in order to provide the current state of knowledge of this topic by organizing and evaluating current literature. In the first section of this paper, the structure...... of a conventional power device module and its related wear-out failure mechanisms with degradation indicators are discussed. Then, representative power cycling test circuits are introduced. Furthermore, on-state collector-emitter voltage (VCE ON) and forward voltage (VF) measurement circuits for wear-out condition...

  3. High power rf component testing for the NLC

    International Nuclear Information System (INIS)

    Vlieks, A.E.; Fowkes, W.R.; Loewen, R.J.; Tantawi, S.G.

    1998-09-01

    In the Next Linear Collider (NLC), the high power rf components must be capable of handling peak rf power levels in excess of 600 MW. In the current view of the NLC, even the rectangular waveguide components must transmit at least 300 MW rf power. At this power level, peak rf fields can greatly exceed 100 MV/m. The authors present recent results of high power tests performed at the Accelerator Structure Test Area (ASTA) at SLAC. These tests are designed to investigate the rf breakdown limits of several new components potentially useful for the NLC. In particular, the authors tested a new TE 01 --TE 10 circular to rectangular wrap-around mode converter, a modified (internal fin) Magic Tee hybrid, and an upgraded flower petal mode converter

  4. Initial proto II pulsed power tests

    International Nuclear Information System (INIS)

    Johnson, D.L.

    1976-01-01

    The Proto II electron beam accelerator is being developed by Sandia Laboratories to study engineering and physics aspects of electron beam pellet fusion. Currently the Marx generator-water capacitor portion of Proto II is undergoing high voltage testing and timing measurements. Eight 112 kJ Marx generators form the primary energy storage system. Each Marx generator pulse charges two parallel 7.5 nF water capacitors to 3 MV. The water capacitors act as intermediate energy storage elements and will transfer their energy to the water insulated pulse-forming lines in 250 ns by means of eight SF 6 gas insulated, trigatron switches. Test data and design considerations of the trigger systems, Marx generators, water capacitors, and trigatron switches are presented

  5. On Locally Most Powerful Sequential Rank Tests

    Czech Academy of Sciences Publication Activity Database

    Kalina, Jan

    2017-01-01

    Roč. 36, č. 1 (2017), s. 111-125 ISSN 0747-4946 R&D Projects: GA ČR GA17-07384S Grant - others:Nadační fond na podporu vědy(CZ) Neuron Institutional support: RVO:67985807 Keywords : nonparametric test s * sequential ranks * stopping variable Subject RIV: BA - General Mathematics OBOR OECD: Pure mathematics Impact factor: 0.339, year: 2016

  6. Penetration of ASM 981 in canine skin: a comparative study.

    Science.gov (United States)

    Gutzwiller, Meret E Ricklin; Reist, Martin; Persohn, Elke; Peel, John E; Roosje, Petra J

    2006-01-01

    ASM 981 has been developed for topical treatment of inflammatory skin diseases. It specifically inhibits the production and release of pro-inflammatory cytokines. We measured the skin penetration of ASM 981 in canine skin and compared penetration in living and frozen skin. To make penetration of ASM 981 visible in dog skin, tritium labelled ASM 981 was applied to a living dog and to defrosted skin of the same dog. Using qualitative autoradiography the radioactive molecules were detected in the lumen of the hair follicles until the infundibulum, around the superficial parts of the hair follicles and into a depth of the dermis of 200 to 500 microm. Activity could not be found in deeper parts of the hair follicles, the dermis or in the sebaceous glands. Penetration of ASM 981 is low in canine skin and is only equally spread in the upper third of the dermis 24 hours after application. Penetration in frozen skin takes even longer than in living canine skin but shows the same distribution.

  7. Test of a High Power Target Design

    CERN Multimedia

    2002-01-01

    %IS343 :\\\\ \\\\ A high power tantalum disc-foil target (RIST) has been developed for the proposed radioactive beam facility, SIRIUS, at the Rutherford Appleton Laboratory. The yield and release characteristics of the RIST target design have been measured at ISOLDE. The results indicate that the yields are at least as good as the best ISOLDE roll-foil targets and that the release curves are significantly faster in most cases. Both targets use 20 -25 $\\mu$m thick foils, but in a different internal geometry.\\\\ \\\\Investigations have continued at ISOLDE with targets having different foil thickness and internal geometries in an attempt to understand the release mechanisms and in particular to maximise the yield of short lived isotopes. A theoretical model has been developed which fits the release curves and gives physical values of the diffusion constants.\\\\ \\\\The latest target is constructed from 2 $\\mu$m thick tantalum foils (mass only 10 mg) and shows very short release times. The yield of $^{11}$Li (half-life of ...

  8. Test and maintenance of the emergency power supply in the nuclear power plant Biblis

    International Nuclear Information System (INIS)

    Kotthoff, K.; Huren, H.

    1986-01-01

    Besides design and construction test and maintenance play an important role for the availability of the emergency power supply. As an example test and maintenance provided for the emergency power supply in a German 4-loop PWR will be described. In general one has to differentiate between test and maintenance performed during power operation of the plant and those carried out during the refuelling outage. For both periods of operation detailed information will be given including type, extent and frequency of test and maintenance work. The results of test and maintenance up to now will be discussed. (authors)

  9. Test and maintenance of the emergency power supply in the nuclear power plant Biblis

    Energy Technology Data Exchange (ETDEWEB)

    Kotthoff, K. [Gesellschaft fuer Reaktorsicherheit - GRS mbH, Schwertnergasse 1, D-5000 Koeln 1, Cologne (Germany); Huren, H. [Rheinisch-Westfaelisches Elektrizitaetswerk AG, Betriebsverwaltung Biblis, Biblis (Germany)

    1986-02-15

    Besides design and construction test and maintenance play an important role for the availability of the emergency power supply. As an example test and maintenance provided for the emergency power supply in a German 4-loop PWR will be described. In general one has to differentiate between test and maintenance performed during power operation of the plant and those carried out during the refuelling outage. For both periods of operation detailed information will be given including type, extent and frequency of test and maintenance work. The results of test and maintenance up to now will be discussed. (authors)

  10. On Locally Most Powerful Sequential Rank Tests

    Czech Academy of Sciences Publication Activity Database

    Kalina, Jan

    2017-01-01

    Roč. 36, č. 1 (2017), s. 111-125 ISSN 0747-4946 R&D Projects: GA ČR GA17-07384S Grant - others:Nadační fond na podporu vědy(CZ) Neuron Institutional support: RVO:67985556 Keywords : nonparametric test s * sequential ranks * stopping variable Subject RIV: BA - General Mathematics OBOR OECD: Pure mathematics Impact factor: 0.339, year: 2016 http://library.utia.cas.cz/separaty/2017/SI/kalina-0474065.pdf

  11. High-precision performance testing of the LHC power converters

    CERN Document Server

    Bastos, M; Dreesen, P; Fernqvist, G; Fournier, O; Hudson, G

    2007-01-01

    The magnet power converters for LHC were procured in three parts, power part, current transducers and control electronics, to enable a maximum of industrial participation in the manufacturing and still guarantee the very high precision (a few parts in 10-6) required by LHC. One consequence of this approach was several stages of system tests: factory reception tests, CERN reception tests, integration tests , short-circuit tests and commissioning on the final load in the LHC tunnel. The majority of the power converters for LHC have now been delivered, integrated into complete converter and high-precision performance testing is well advanced. This paper presents the techniques used for high-precision testing and the results obtained.

  12. Improving Size and Power in Unit Root Testing

    DEFF Research Database (Denmark)

    Haldrup, Niels; Jansson, Michael

    A frequent criticism of unit root tests concerns the poor power and size properties that many of such testsexhibit. However, the past decade or so intensive research has been conducted to alleviate these problems and great advances have been made. The present paper provides a selective survey...... of recent contributions to improve upon both size and power of unit root tests and in so doing the approach of using rigorous statistical optimality criteria in the development of such tests is stressed. In addition to presenting tests where improved size can be achieved by modifying the standard Dickey......-Fuller class of tests, the paper presents theory of optimal testing and the construction of power envelopes for unit root tests underdifferent conditions allowing for serial correlation, deterministic components, assumptions regarding the initial condition, non-Gaussian errors, and the use of covariates....

  13. Data acquisition system for medium power neutral beam test facility

    International Nuclear Information System (INIS)

    Stewart, C.R. Jr.; Francis, J.E. Jr.; Hammons, C.E.; Dagenhart, W.K.

    1978-06-01

    The Medium Power Neutral Beam Test Facility at Oak Ridge National Laboratory was constructed in order to develop, test, and condition powerful neutral beam lines for the Princeton Large Torus experiment at Princeton Plasma Physics Laboratory. The data acquisition system for the test stand monitors source performance, beam characteristics, and power deposition profiles to determine if the beam line is operating up to its design specifications. The speed of the computer system is utilized to provide near-real-time analysis of experimental data. Analysis of the data is presented as numerical tabulation and graphic display

  14. Photovoltaic Test and Demonstration Project. [for solar cell power systems

    Science.gov (United States)

    Forestieri, A. F.; Brandhorst, H. W., Jr.; Deyo, J. N.

    1976-01-01

    The Photovoltaic Test and Demonstration Project was initiated by NASA in June, 1975, to develop economically feasible photovoltaic power systems suitable for a variety of terrestrial applications. Objectives include the determination of operating characteristic and lifetimes of a variety of solar cell systems and components and development of methodology and techniques for accurate measurements of solar cell and array performance and diagnostic measurements for solar power systems. Initial work will be concerned with residential applications, with testing of the first prototype system scheduled for June, 1976. An outdoor 10 kW array for testing solar power systems is under construction.

  15. The Great Plains Wind Power Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John [Texas Tech Univ., Lubbock, TX (United States)

    2014-01-30

    This multi-year, multi-faceted project was focused on the continued development of a nationally-recognized facility for the testing, characterization, and improvement of grid-connected wind turbines, integrated wind-water desalination systems, and related educational and outreach topics. The project involved numerous faculty and graduate students from various engineering departments, as well as others from the departments of Geosciences (in particular the Atmospheric Science Group) and Economics. It was organized through the National Wind Institute (NWI), which serves as an intellectual hub for interdisciplinary and transdisciplinary research, commercialization and education related to wind science, wind energy, wind engineering and wind hazard mitigation at Texas Tech University (TTU). Largely executed by an academic based team, the project resulted in approximately 38 peer-reviewed publications, 99 conference presentations, the development/expansion of several experimental facilities, and two provisional patents.

  16. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-01

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  17. Index to place of publication of ASME Papers, 1978--1988

    International Nuclear Information System (INIS)

    Youngen, G.K.

    1990-06-01

    This index is a list of American Society of Mechanical Engineers (ASME) Papers that are reprinted in the ASME Transactions series of journals. ASME Papers are often cited only by their paper number, making it difficult to determine if the article has ever appeared in print in the journal literature. This index will be useful for tracking down those papers published as journal articles by the ASME. It will also serve as a guide for retention for subscribers to the ASME Papers and Transaction Series. Paper numbers that appear in the journals may be weeded from the collection of ASME Papers

  18. Report on the Current Technical Issues on ASME Nuclear Code and Standard

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2008-11-01

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  19. Analysis of the Current Technical Issues on ASME Code and Standard for Nuclear Mechanical Design(2009)

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, B. S.; Yoo, S. H.

    2009-11-15

    This report describes the analysis on the current revision movement related to the mechanical design issues of the U.S ASME nuclear code and standard. ASME nuclear mechanical design in this report is composed of the nuclear material, primary system, secondary system and high temperature reactor. This report includes the countermeasures based on the ASME Code meeting for current issues of each major field. KAMC(ASME Mirror Committee) of this project is willing to reflect a standpoint of the domestic nuclear industry on ASME nuclear mechanical design and play a technical bridge role for the domestic nuclear industry in ASME Codes application

  20. Free-piston Stirling component test power converter

    Science.gov (United States)

    Dochat, George; Dhar, Manmohan

    1991-01-01

    The National Aeronautics and Space Administration (NASA) has been evaluating free-piston Stirling power converters (FPSPCs) for use on a wide variety of space missions. They provide high reliability, long life, and efficient operation and can be coupled with all potential heat sources, various heat input and heat rejection systems, and various power management and distribution systems. FPSPCs can compete favorably with alternative power conversion systems over a range of hundreds of watts to megawatts. Mechanical Technology Incorporated (MTI) is developing FPSPC technology under contract to NASA Lewis Research Center and will demonstrate this technology in two full-scale power converters operating at space temperature conditions. The testing of the first of these, the component test power converter (CTPC), was initiated in Spring 1991 to evaluate mechanical operation at space operating temperatures. The CTPC design, hardware fabrication, and initial test results are reviewed.

  1. Estimates of the burst reliability of thin-walled cylinders designed to meet the ASME Code allowables

    International Nuclear Information System (INIS)

    Stancampiano, P.A.; Zemanick, P.P.

    1976-01-01

    Pressure containment components in nuclear power plants are designed by the conventional deterministic safety factor approach to meet the requirements of the ASME Pressure Vessel Code, Section III. The inevitable variabilities and uncertainties associated with the design, manufacture, installation, and service processes suggest a probabilistic design approach may also be pertinent. Accordingly, the burst reliabilities of two thin-walled 304 SS cylindrical vessels such as might be employed in liquid metal plants are estimated. A large vessel fabricated from rolled plate per ASME SA-240 and a smaller pipe sized vessel also fabricated from rolled plate per ASME SA-358 are considered. The vessels are sized to just meet the allowable ASME Code primary membrance stresses at 800 0 F (427 0 C). The bursting probability that the operating pressure is greater than the burst strength of the cylinders is calculated using stress-strength interference theory by direct Monte Carlo simulation on a high speed digital computer. A sensitivity study is employed to identify those design parameters which have the greatest effect on the reliability. The effects of preservice quality assurance defect inspections on the reliability are also evaluated parametrically

  2. Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability

    International Nuclear Information System (INIS)

    Chopra, O. K.; Shack, W. J.

    2003-01-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ((var e psilon)-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue (var e psilon)-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue (var e psilon)-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented

  3. Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Shack, W. J.; Energy Technology

    2003-10-03

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue {var_epsilon}-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented.

  4. LHC Report: superconducting circuit powering tests

    CERN Multimedia

    Mirko Pojer

    2015-01-01

    After the long maintenance and consolidation campaign carried out during LS1, the machine is getting ready to start operation with beam at 6.5 TeV… the physics community can’t wait! Prior to this, all hardware and software systems have to be tested to assess their correct and safe operation.   Most of the cold circuits (those with high current/stored energy) possess a sophisticated magnet protection system that is crucial to detect a transition of the coil from the superconducting to the normal state (a quench) and safely extract the energy stored in the circuits (about 1 GJ per dipole circuit at nominal current). LHC operation relies on 1232 superconducting dipoles with a field of up to 8.33 T operating in superfluid helium at 1.9 K, along with more than 500 superconducting quadrupoles operating at 4.2 or 1.9 K. Besides, many other superconducting and normal resistive magnets are used to guarantee the possibility of correcting all beam parameters, for a total of mo...

  5. Semiparametric Power Envelopes for Tests of the Unit Root Hypothesis

    DEFF Research Database (Denmark)

    Jansson, Michael

    This paper derives asymptotic power envelopes for tests of the unit root hypothesis in a zero-mean AR(1) model. The power envelopes are derived using the limits of experiments approach and are semiparametric in the sense that the underlying error distribution is treated as an unknown...

  6. Application of MATLAB in testing digital power supply controllers

    International Nuclear Information System (INIS)

    Ke Xinhua; Chinese Academy of Sciences, Beijing; Lu Songlin; Shen Tianjian

    2008-01-01

    Based on introducing MATLAB SISOTOOL in MATLAB control box and the magnet power supply system in detail, this paper presented the application of MATLAB SISOTOOL in Testing Digital Power Supply controllers. This control tool should be popularized because of its characteristics of convenience and easy-to-use. (authors)

  7. Micro-Combined Heat and Power Device Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — NIST has developed a test facility for micro-combined heat and power (micro-CHP) devices to measure their performance over a range of different operating strategies...

  8. ac power control in the Core Flow Test Loop

    International Nuclear Information System (INIS)

    McDonald, D.W.

    1980-01-01

    This work represents a status report on a development effort to design an ac power controller for the Core Flow Test Loop. The Core Flow Test Loop will be an engineering test facility which will simulate the thermal environment of a gas-cooled fast-breeder reactor. The problems and limitations of using sinusoidal ac power to simulate the power generated within a nuclear reactor are addressed. The transformer-thyristor configuration chosen for the Core Flow Test Loop power supply is presented. The initial considerations, design, and analysis of a closed-loop controller prototype are detailed. The design is then analyzed for improved performance possibilities and failure modes are investigated at length. A summary of the work completed to date and a proposed outline for continued development completes the report

  9. Towards a foundation for holistic power system validation and testing

    DEFF Research Database (Denmark)

    Blank, M.; Lehnhoff, S.; Heussen, Kai

    2016-01-01

    , and intelligent solutions for system operation have transformed the power system into a smart grid. To support the development process of smart grid solutions on system level they have to be tested in a holistic manner covering the multi-domain aspect of a such complex systems. This paper introduces the concept...... of holistic power system testing and discuss first steps towards a corresponding methodology that is being developed in the European ERIGrid research infrastructure project....

  10. Demonstration test for reliability of valves for atomic power plants

    International Nuclear Information System (INIS)

    Hosaka, Shiro

    1978-01-01

    The demonstration test on the reliability of valves for atomic power plants being carried out by the Nuclear Engineering Test Center is reported. This test series is conducted as six-year project from FY 1976 to FY 1981 at the Isogo Test Center. The demonstration test consists of (1) environmental test, (2) reaction force test, (3) vibration test, (4) stress measurement test, (5) operational characteristic test, (6) flow resistance coefficient measuring test, (7) leakage test and (8) safety valve and relief valve test. These contents are explained about the special requirements for nuclear use, for example, the enviornmental condition after the design base accident of PWRs and BWRs, the environmental test sequence for isolation valves of containment vessels under the emergency condition, the seismic test condition for valves of nuclear use, the various stress measurements under thermal transient conditions, the leak test after 500 cycles between the normal operating conditions for PWRs and BWRs and the start up conditions and so on. As for the testing facilities, the whole flow diagram is shown, in which the environmental test section, the vibration test section, the steam test section, the hot water test section, the safety valve test section and main components are included. The specifications of each test section and main components are presented. (Nakai, Y.)

  11. Evaluation of the linear power of HANARO test fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Choong Sung; Seo, C. G.; Lee, B. C.; Kim, H. R

    2001-02-01

    The HANARO fuel was developed by AECL and it is configured in a bundle of rods containing uranium silicide. AECL has conducted a variety of tests using specimen in order to achieve its qualification and licensing and the highest linear power was evaluated to be 112.8kW/m. In design stage of HANARO, the best estimated maximum linear power at hot spot was found to occur in the transition core from the initial to the equilibrium and its value was 108kW/m, which exceeds 112.8kW/m if the physics uncertainty of the HANARO nuclear design model is taken into account. Consequently, the licensing body issued the conditional permit to operate HANARO and the fuel integrity at the linear power higher than 112.8kW/m was requested to be confirmed through irradiation tests by realizing its repeatability. Hereby, KAERI designed uninstrumented and instrumented test fuel bundles and conducted their burnup tests. In parallel with the tests, the nuclear design model has been revised and updated to enable us to pursue the pin-by-pin power history. This report describes the best estimated power history of the test fuel bundles using the revised model. In conclusion, HANARO fuel keeps its integrity at power condition greater than 120kW/m.

  12. ASM observations of X-ray flares from 4U 0115+63 and ASM 1354-64.

    Science.gov (United States)

    Tsunemi, H.; Kitamoto, S.

    The authors report two X-ray flares detected with the All Sky Monitor (ASM) on board the GINGA satellite. One is from the recurrent X-ray pulsar 4U 0115+63 and the other is from the probable recurrent X-ray nova named ASM 1354-64. The maximum intensity for 4U 0115+63 was 180 mCrab and its duration was at least 22 days. Its spectrum was hard and resembled those of X-ray pulsars. The maximum intensity of ASM 1354-64 was 300 mCrab. It faded down below the detection limit at the end of August 1987. Its spectrum was soft and was similar to those of black hole candidates.

  13. Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

    2002-08-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

  14. Mechanical design and fabrication of power feed cavity test setup

    International Nuclear Information System (INIS)

    Ghodke, S.R.; Dhavle, A.S.; Sharma, Vijay; Sarkar, Shreya; Kumar, Mahendra; Nayak, Susanta; Barnwal, Rajesh; Jayaprakash, D.; Mondal, J.; Nimje, V.T.; Mittal, K.C.; Gantayet, L.M.

    2013-01-01

    Power feed cavity set up consists of nine number of accelerating cavity and eight numbers of coupling cavity for testing of power feed cavity with coupling flange for 2856 MHz S band standing wave coupled cavity linac. When we are assembling the cavity and applying the pressure, its resonance frequency changes with applied pressure/load. After some critical pressure/load frequency change becomes negligible or zero. This set up will be used to find out assembly performance of power feed cavity and its coupler. Top four cavity or eight half cells as well as bottom four cavity or eight half cells will be brazed separately. Power feed cavity will be sandwiched between this two brazed cavity assemblies. This paper discuss about linear motion bush, linear motion rod, load cell, hydraulic actuator, power pack, stepper motor PLC control, jig boring, alignment, tolerances and assembly procedure for this test setup. (author)

  15. TRACKING TESTS FOR THE SNS FAST INJECTION BUMP POWER SUPPLY

    International Nuclear Information System (INIS)

    ENG, W.; CUTLER, R.; DEWAN, S.

    2004-01-01

    The tracking requirement of the SNS Fast Injection Bump power supplies is described. In addition to the usual tracking between the load current and the input reference of a power supply, these power supplies must also track between pairs of units under slightly different loads. This paper describes the use of a current-null test to measure tracking performances. For the actual tests, a single dummy magnet load was used to measure the tracking between the first two production units at the manufacturer's facility. Using the Yokogawa WE7000 waveform. PC-based measurement instrument, input and output waveforms are digitized and stored in data files. A program written for this application is then used to extract data from these files to construct, analyze the waveforms and characterize the power supply performance. Results of the measurements of two SNS Fast Injection Bump power supplies will be presented in this paper

  16. Evaluation of power control concepts using the PMAD systems test bed. [Power Management and Distribution

    Science.gov (United States)

    Beach, R. F.; Kimnach, G. L.; Jett, T. A.; Trash, L. M.

    1989-01-01

    The Lewis Research Center's Power Management and Distribution (PMAD) System testbed and its use in the evaluation of control concepts applicable to the NASA Space Station Freedom electric power system (EPS) are described. The facility was constructed to allow testing of control hardware and software in an environment functionally similar to the space station electric power system. Control hardware and software have been developed to allow operation of the testbed power system in a manner similar to a supervisory control and data acquisition (SCADA) system employed by utility power systems for control. The system hardware and software are described.

  17. ASME code considerations for the compact heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Nestell, James [MPR Associates Inc., Alexandria, VA (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-31

    The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a working fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and

  18. European wind turbine testing procedure developments. Task 2: Power quality

    DEFF Research Database (Denmark)

    Sørensen, Poul Ejnar; Friis Pedersen, Troels; Gerdes, G.

    2001-01-01

    The present report describes the work done in the power quality subtask of the European Wind Turbine Testing Procedure Developments project funded by the EU SMT program. The objective of the power quality subtask has been to make recommendations andprovide background for new standards...... for measurement and testing of wind turbine power quality. The focus in the work has been to support the ongoing standardisation work in IEC with a new standard IEC61400-21 for measurement and assessment of powerquality characteristics of grid connected wind turbines. The work has also been based on the power...... quality measuremnet procedure in the Measnet cooperation of European test stations for wind turbines. The first working item of the project has been toverify the state of the art of the measurement procedures by analyses and comparisons of the measurements and data processing software of the participating...

  19. 14 CFR 330.31 - What data must air carriers submit concerning ASMs or RTMs?

    Science.gov (United States)

    2010-01-01

    ... combination passenger/cargo carrier, you must have submitted your August 2001 total completed ASM report to... correct an error that you document to the Department, you must not alter the ASM or RTM reports you...

  20. Brayton Power Conversion Unit Tested: Provides a Path to Future High-Power Electric Propulsion Missions

    Science.gov (United States)

    Mason, Lee S.

    2003-01-01

    Closed-Brayton-cycle conversion technology has been identified as an excellent candidate for nuclear electric propulsion (NEP) power conversion systems. Advantages include high efficiency, long life, and high power density for power levels from about 10 kWe to 1 MWe, and beyond. An additional benefit for Brayton is the potential for the alternator to deliver very high voltage as required by the electric thrusters, minimizing the mass and power losses associated with the power management and distribution (PMAD). To accelerate Brayton technology development for NEP, the NASA Glenn Research Center is developing a low-power NEP power systems testbed that utilizes an existing 2- kWe Brayton power conversion unit (PCU) from previous solar dynamic technology efforts. The PCU includes a turboalternator, a recuperator, and a gas cooler connected by gas ducts. The rotating assembly is supported by gas foil bearings and consists of a turbine, a compressor, a thrust rotor, and an alternator on a single shaft. The alternator produces alternating-current power that is rectified to 120-V direct-current power by the PMAD unit. The NEP power systems testbed will be utilized to conduct future investigations of operational control methods, high-voltage PMAD, electric thruster interactions, and advanced heat rejection techniques. The PCU was tested in Glenn s Vacuum Facility 6. The Brayton PCU was modified from its original solar dynamic configuration by the removal of the heat receiver and retrofitting of the electrical resistance gas heater to simulate the thermal input of a steady-state nuclear source. Then, the Brayton PCU was installed in the 3-m test port of Vacuum Facility 6, as shown. A series of tests were performed between June and August of 2002 that resulted in a total PCU operational time of about 24 hr. An initial test sequence on June 17 determined that the reconfigured unit was fully operational. Ensuing tests provided the operational data needed to characterize PCU

  1. Purchasing Power Parity : Evidence from a New Test

    NARCIS (Netherlands)

    Klaassen, F.J.G.M.

    1999-01-01

    Most economists intuitively consider purchasing power parity (PPP) to be true. Nevertheless, quite surprisingly, the empirical literature is not very supportive for PPP. In this paper, however, we find evidence in favor of PPP using a new test. The test is embedded in a Markov regime-switching model

  2. Best Practices: Power Quality and Integrated Testing at JSC

    Science.gov (United States)

    Davis, Lydia

    2018-01-01

    This presentation discusses Best Practices for Power Quality and Integrated Testing at JSC in regards to electrical systems. These high-level charts include mostly generic information; however, a specific issue is discussed involving flight hardware that could have been discovered prior to flight with an integrated test.

  3. Regulatory Endorsement Activities for ASME Nuclear Codes and Standards

    International Nuclear Information System (INIS)

    West, Raymond A.

    2006-01-01

    The ASME Board on Nuclear Codes and Standards (BNCS) has formed a Task Group on Regulatory Endorsement (TG-RE) that is currently in discussions with the United States Nuclear Regulatory Commission (NRC) to look at suggestions and recommendations that can be used to help with the endorsement of new and revised ASME Nuclear Codes and Standards (NC and S). With the coming of new reactors in the USA in the very near future we need to look at both the regulations and all the ASME NC and S to determine where we need to make changes to support these new plants. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors. This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort. (author)

  4. Advanced Test Accelerator (ATA) pulse power technology development

    International Nuclear Information System (INIS)

    Reginato, L.L.; Branum, D.; Cook, E.

    1981-01-01

    The Advanced Test Accelerator (ATA) is a pulsed linear induction accelerator with the following design parameters: 50 MeV, 10 kA, 70 ns, and 1 kHz in a ten-pulse burst. Acceleration is accomplished by means of 190 ferrite-loaded cells, each capable of maintaining a 250 kV voltage pulse for 70 ns across a 1-inch gap. The unique characteristic of this machine is its 1 kHz burst mode capability at very high currents. This paper dscribes the pulse power development program which used the Experimental Test Accelerator (ETA) technology as a starting base. Considerable changes have been made both electrically and mechanically in the pulse power components with special consideration being given to the design to achieve higher reliability. A prototype module which incorporates all the pulse power components has been built and tested for millions of shots. Prototype components and test results are described

  5. AGS tune jump power supply design and test

    International Nuclear Information System (INIS)

    Mi, J.; Glenn, J.W.; Huang, H.; Marneris, I.; Rosas, P.; Sandberg, J.; Tan, Y.; Zhang, W.

    2011-01-01

    A horizontal tune jump system has been installed to overcome the horizontal intrinsic spin resonances, which requires jumping the horizontal tune 0.04 units 82 times, 41 up and 41 down. Two quadruple magnets have been installed in AGS ring to perform this. The pulsed magnet current ranges from about 140A near injection to about 1400A later. The current pulse rise and fall time are around 100uS and flat tops time is around 4mS. These quadruples have separated supplies. This tune jump pulse power supply employees all semiconductor parts as well as the main switches. During dummy load and magnet testing, the test results showed that the power supply could meet the specification. This article will describe some details of power supply simulation, design and testing. Some test waveforms and pictures are presented in this paper.

  6. Avaliação da inflamação de vias aéreas em asmáticos após o teste de broncoprovocação com metacolina

    Directory of Open Access Journals (Sweden)

    MATOS FABRÍCIO LINO DE

    2001-01-01

    Full Text Available Introdução: O teste de broncoprovocação com metacolina é comumente empregado em pneumologia para medir a reatividade brônquica com finalidade de diagnóstico ou acompanhamento da asma. Objetivo: Pesquisar efeitos tardios da inalação de metacolina na inflamação brônquica avaliada pelo escarro induzido. Casuística e métodos: Foram selecionados dez pacientes com asma leve ou moderada, não tabagistas, com medicação e quadro clínico estáveis. Às 12:00h, os pacientes receberam inalação, aleatoriamente designada, de metacolina (broncoprovocação ou soro fisiológico. Às 18:00h, foi realizada a indução de escarro. Em outro dia, com intervalo de uma semana, os pacientes completaram o protocolo, recebendo a outra inalação (metacolina ou soro fisiológico e nova indução de escarro. Resultados: Após metacolina, obtiveram-se 8,6 ± 9g de escarro, 8,6 ± 6 milhões de células, sendo 78 ± 10% viáveis e 6,8 ± 7% eosinófilos. Esses dados não foram significativamente diferentes dos resultados obtidos após soro fisiológico: escarro = 7,6 ± 6g, células = 12,4 ± 12 milhões, 82 ± 10% viáveis e 6,6 ± 9% eosinófilos. A queda de pico de fluxo observada durante a indução de escarro também não diferiu: 21,4 ± 12% após metacolina e 18,4 ± 15% após soro fisiológico. A queda de pico de fluxo durante a indução correlacionou-se com a quantidade de escarro (p = 0,018 e percentagem de eosinófilos (p = 0,003. Outras correlações entre parâmetros funcionais e do escarro não foram significantes. Conclusão: O teste de broncoprovocação com metacolina realizado seis horas antes da indução de escarro não altera significativamente a quantidade e nem a constituição celular do escarro.

  7. High Power RF Test Facility at the SNS

    CERN Document Server

    Kang, Yoon W; Campisi, Isidoro E; Champion, Mark; Crofford, Mark; Davis, Kirk; Drury, Michael A; Fuja, Ray E; Gurd, Pamela; Kasemir, Kay-Uwe; McCarthy, Michael P; Powers, Tom; Shajedul Hasan, S M; Stirbet, Mircea; Stout, Daniel; Tang, Johnny Y; Vassioutchenko, Alexandre V; Wezensky, Mark

    2005-01-01

    RF Test Facility has been completed in the SNS project at ORNL to support test and conditioning operation of RF subsystems and components. The system consists of two transmitters for two klystrons powered by a common high voltage pulsed converter modulator that can provide power to two independent RF systems. The waveguides are configured with WR2100 and WR1150 sizes for presently used frequencies: 402.5 MHz and 805 MHz. Both 402.5 MHz and 805 MHz systems have circulator protected klystrons that can be powered by the modulator capable of delivering 11 MW peak and 1 MW average power. The facility has been equipped with computer control for various RF processing and complete dual frequency operation. More than forty 805 MHz fundamental power couplers for the SNS superconducting linac (SCL) cavitites have been RF conditioned in this facility. The facility provides more than 1000 ft2 floor area for various test setups. The facility also has a shielded cave area that can support high power tests of normal conducti...

  8. Experiences with drug testing at a nuclear power plant

    International Nuclear Information System (INIS)

    Ray, H.B.

    1987-01-01

    After more than 2 yr of operation of a drug testing program at the San Onofre nuclear power plant site, the Southern California Edison Co. has had a number of experiences and lessons considered valuable. The drug testing program at San Onofre, implemented in September of 1984, continues in essentially the same form today. Prior to describing the program, the paper reviews several underlying issues that believed to be simultaneously satisfied by the program: trustworthiness, fitness and safety, public trust, and privacy and search. The overall drug testing program, periodic drug monitoring program, and unannounced drug testing program are described. In addition to the obvious features of a good drug testing program, which are described in the EEI guide, it is essential to consider such issues as the stated program rationale, employee relations, and disciplinary action measures when contemplating or engaging in drug testing at nuclear power plants

  9. PUREX SAMCONS uninterruptible power supply (UPS) acceptance test report

    International Nuclear Information System (INIS)

    Blackaby, W.B.

    1997-01-01

    This Acceptance Test Report for the PUREX Surveillance and Monitoring and Control System (SAMCONS) Uninterruptible Power Supply (UPS) Acceptance Test Procedure validates the operation of the UPS, all alarming and display functions and the ability of the UPS to supply power to the SAMCONS as designed. The proper installation of the PUREX SAMCONS Trailer UPS components and wiring will be systematically evaluated by performance of this procedure. Proper operation of the SAMCONS computer UPS will be verified by performance of a timed functional load test, and verification of associated alarms and trouble indications. This test procedure will be performed in the SAMCONS Trailer and will include verification of receipt of alarms at the SAMCONS computer stations. This test may be performed at any time after the completion of HNF-SD-CP-ATP-083, PUREX Surveillance and Monitoring and Control System (SAMCONS) Acceptance Test Procedure, when computer display and alarm functions have been proven to operate correctly

  10. Results from Evaluation of Proposed ASME AG-1 Section FI Metal Media Filters - 13063

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, John A.; Giffin, Paxton K.; Parsons, Michael S.; Waggoner, Charles A. [Institute for Clean Energy Technology, Mississippi State University, 205 Research Blvd Starkville, MS 39759 (United States)

    2013-07-01

    High efficiency particulate air (HEPA) filtration technology is commonly used in Department of Energy (DOE) facilities that require control of radioactive particulate matter (PM) emissions due to treatment or management of radioactive materials. Although HEPA technology typically makes use of glass fiber media, metal and ceramic media filters are also capable of filtering efficiencies beyond the required 99.97%. Sintered metal fiber filters are good candidates for use in DOE facilities due to their resistance to corrosive environments and resilience at high temperature and elevated levels of relative humidity. Their strength can protect them from high differential pressure or pressure spikes and allow for back pulse cleaning, extending filter lifetime. Use of these filters has the potential to reduce the cost of filtration in DOE facilities due to life cycle cost savings. ASME AG-1 section FI has not been approved due to a lack of protocols and performance criteria for qualifying section FI filters. The Institute for Clean Energy Technology (ICET) with the aid of the FI project team has developed a Section FI test stand and test plan capable of assisting in the qualification ASME AG-1 section FI filters. Testing done at ICET using the FI test stand evaluates resistance to rated air flow, test aerosol penetration and resistance to heated air of the section FI filters. Data collected during this testing consists of temperature, relative humidity, differential pressure, flow rate, upstream particle concentration, and downstream particle concentration. (authors)

  11. Advanced Accelerated Power Cycling Test for Reliability Investigation of Power Device Modules

    DEFF Research Database (Denmark)

    Choi, Uimin; Jørgensen, Søren; Blaabjerg, Frede

    2016-01-01

    This paper presents an apparatus and methodology for an advanced accelerated power cycling test of insulated-gate bipolar transistor (IGBT) modules. In this test, the accelerated power cycling test can be performed under more realistic electrical operating conditions with online wear-out monitoring...... of tested power IGBT module. The various realistic electrical operating conditions close to real three-phase converter applications can be achieved by the simple control method. Further, by the proposed concept of applying the temperature stress, it is possible to apply various magnitudes of temperature...... swing in a short cycle period and to change the temperature cycle period easily. Thanks to a short temperature cycle period, test results can be obtained in a reasonable test time. A detailed explanation of apparatus such as configuration and control methods for the different functions of accelerated...

  12. Performance test of uninterruptible power system of PIEF

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Chae; Kim, Eun Ka; Chun, Yong Bum; Park, Dea Gyu; Chu, Yong Sun; Bae, Sang Min; Koo, Dae Seo

    1998-02-01

    Because of the special features of post-irradiation examination (PIE) facility to handle very high radioactive materials like spent nuclear fuels, the electric system of the facility was designed and constructed according to a very strict requirement which is applied to nuclear power plant. A safety grade of Class 1E was adopted in the power utility system of PIEF to guarantee stable power supply to the facility without any expected interruption. In order cope with a emergency condition like a power interruption of KEPCO, a emergency power supplying system consisting of a diesel generator (3-phase, 6600/440, 1,000 kW) and uninterruptibel power supply (UPS) system was installed in PIEF. UPS power is connected to the radiation monitoring system and several other main safety devices to assure of normal operations of them for not less than 30 minutes. According to the recommendations and regulations in nuclear law, a monthly and yearly regular inspection for the UPS and emergency power supplying system are performed. In this report, a brief description to establish self-inspection technology and procedures for the above mentioned electric power supplying system at PIEF, including a principle of operation, inspection scheme, trouble shooting, and performance test techniques were made. (author). 8 refs., 3 tabs., 4 figs.

  13. Free-piston Stirling component test power converter test results and potential Stirling applications

    Science.gov (United States)

    Dochat, G. R.

    1992-01-01

    As the principal contractor to NASA-Lewis Research Center, Mechanical Technology Incorporated is under contract to develop free-piston Stirling power converters in the context of the competitive multiyear Space Stirling Technology Program. The first generation Stirling power converter, the component test power converter (CTPC) initiated cold end testing in 1991, with hot testing scheduled for summer of 1992. This paper reviews the test progress of the CTPC and discusses the potential of Stirling technology for various potential missions at given point designs of 250 watts, 2500 watts, and 25,000 watts.

  14. Power-Cooling-Mismatch Test Series Test PCM-7. Experiment operating specifications

    International Nuclear Information System (INIS)

    Sparks, D.T.; Smith, R.H.; Stanley, C.J.

    1979-02-01

    The experiment operating specifications for the Power-Cooling-Mismatch (PCM) Test PCM-7 to be conducted in the Power Burst Facility are described. The PCM Test Series was designed on the basis of a parametric evaluation of fuel behavior response with cladding temperature, rod internal pressure, time in film boiling, and test rod power being the variable parameters. The test matrix, defined in the PCM Experiment Requirements Document (ERD), encompasses a wide range of situations extending from pre-CHF (critical heat flux) PCMs to long duration operation in stable film boiling leading to rod failure

  15. Standards for measurements and testing of wind turbine power quality

    Energy Technology Data Exchange (ETDEWEB)

    Soerensen, P [Risoe National Lab., Roskilde (Denmark); Gerdes, G; Klosse, R; Santjer, F [DEWI, Wilhelmshaven (Germany); Robertson, N; Davy, W [NEL, Glasgow (United Kingdom); Koulouvari, M; Morfiadakis, E [CRES, Pikermi (Greece); Larsson, Aa [Chalmers Univ. of Technology, Goeteborg (Sweden)

    1999-03-01

    The present paper describes the work done in power quality sub-task of the project `European Wind Turbine Testing Procedure Developments` funded by the EU SMT program. The objective of the power quality sub-task has been to make analyses and new recommendation(s) for the standardisation of measurement and verification of wind turbine power quality. The work has been organised in three major activities. The first activity has been to propose measurement procedures and to verify existing and new measurement procedures. This activity has also involved a comparison of the measurements and data processing of the participating partners. The second activity has been to investigate the influence of terrain, grid properties and wind farm summation on the power quality of wind turbines with constant rotor speed. The third activity has been to investigate the influence of terrain, grid properties and wind farm summation on the power quality of wind turbines with variable rotor speed. (au)

  16. Practical test of the LINAC4 RF power system

    CERN Document Server

    Schwerg, N

    2011-01-01

    The high RF power for the Linac4 accelerating structures will be generated by thirteen 1.3 MW klystrons, previously used for the CERN LEP accelerator, and six new klystrons of 2.8 MW all operating at a frequency of 352.2 MHz. The power distribution scheme features a folded magic tee feeding the power from one 2.8 MW klystron to two LEP circulators. We present first results from the Linac4 test place, validating the approach and the used components as well as reporting on the klystron re-tuning activities.

  17. Power systems for the RHIC first sextant test

    International Nuclear Information System (INIS)

    Schultheiss, C.; Bruno, D.; Feng, P.K.

    1997-01-01

    The first sextant test of the RHIC project is an opportunity to evaluate the many systems that must work together for the accelerator to operate. For the main dipole string, the actual main quadrupole power supply with its DSP regulator and output circuit compartment will be used. Temporary supplies will be used for the main quadrupole string, quadrupole offset, and quadrupole shunt supplies. This will let the authors both measure the performance of the main supply as well as determine the interaction among other power elements in the circuit. Correction elements will also be powered. The actual gamma-T power supplies will be used, as well as temporary supplies for the dipole correctors and sextupole supplies. Some of these units are required for beam to be transported, others are to be operated without beam to measure their performance, and how they interact with their superconducting loads. The power supply equipment, and that of other systems, required an infrastucture of AC power and output cable distribution in the RHIC tunnel, outlying service buildings, and interconnecting the tunnel to the service buildings. This note will describe the performance of the RHIC power supply systems during the sextant test, and the experience gained from this exercise

  18. On-site tests on the nuclear power plants

    International Nuclear Information System (INIS)

    Morilhat, P.; Favennec, J.M.; Neau, P.; Preudhomme, E.

    1996-01-01

    On-site tests and experiments are performed by EDF Research and Development Division on the nuclear power plants to assess the behaviour of major components submitted to thermal and vibratory solicitations. On-going studies deal with the qualification of new nuclear power plant standard and with the feedback of plants under operation. The tests, particularly the investigation tests, correspond to large investments and entail an important data volume which must ensure the continuity over a long period of the order of magnitude of the in-service plant life (around 40 years). This paper addresses the on-site experimental activities, describes the means to be used, and gives an example: the qualification of SG of new 1450 MW nuclear power plants. (author)

  19. A High-power Electric Propulsion Test Platform in Space

    Science.gov (United States)

    Petro, Andrew J.; Reed, Brian; Chavers, D. Greg; Sarmiento, Charles; Cenci, Susanna; Lemmons, Neil

    2005-01-01

    This paper will describe the results of the preliminary phase of a NASA design study for a facility to test high-power electric propulsion systems in space. The results of this design study are intended to provide a firm foundation for subsequent detailed design and development activities leading to the deployment of a valuable space facility. The NASA Exploration Systems Mission Directorate is sponsoring this design project. A team from the NASA Johnson Space Center, Glenn Research Center, the Marshall Space Flight Center and the International Space Station Program Office is conducting the project. The test facility is intended for a broad range of users including government, industry and universities. International participation is encouraged. The objectives for human and robotic exploration of space can be accomplished affordably, safely and effectively with high-power electric propulsion systems. But, as thruster power levels rise to the hundreds of kilowatts and up to megawatts, their testing will pose stringent and expensive demands on existing Earth-based vacuum facilities. These considerations and the human access to near-Earth space provided by the International Space Station (ISS) have led to a renewed interest in space testing. The ISS could provide an excellent platform for a space-based test facility with the continuous vacuum conditions of the natural space environment and no chamber walls to modify the open boundary conditions of the propulsion system exhaust. The test platform could take advantage of the continuous vacuum conditions of the natural space environment. Space testing would provide open boundary conditions without walls, micro-gravity and a realistic thermal environment. Testing on the ISS would allow for direct observation of the test unit, exhaust plume and space-plasma interactions. When necessary, intervention by on-board personnel and post-test inspection would be possible. The ISS can provide electrical power, a location for

  20. Assessment of ASME code examinations on regenerative, letdown and residual heat removal heat exchangers

    International Nuclear Information System (INIS)

    Gosselin, Stephen R.; Cumblidge, Stephen E.; Anderson, Michael T.; Simonen, Fredric A.; Tinsley, G A.; Lydell, B.; Doctor, Steven R.

    2005-01-01

    Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principle reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers regenerative heat exchangers and residual (decay) heat removal heat exchangers Design, operating, common preventative maintenance practices, and potential degradation mechanisms are reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp- database and covers PWR plants worldwide for the period 1970-2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers is performed. The results are then reviewed to discuss the examinations relative to plant safety and

  1. Statistical power analyses using G*Power 3.1: tests for correlation and regression analyses.

    Science.gov (United States)

    Faul, Franz; Erdfelder, Edgar; Buchner, Axel; Lang, Albert-Georg

    2009-11-01

    G*Power is a free power analysis program for a variety of statistical tests. We present extensions and improvements of the version introduced by Faul, Erdfelder, Lang, and Buchner (2007) in the domain of correlation and regression analyses. In the new version, we have added procedures to analyze the power of tests based on (1) single-sample tetrachoric correlations, (2) comparisons of dependent correlations, (3) bivariate linear regression, (4) multiple linear regression based on the random predictor model, (5) logistic regression, and (6) Poisson regression. We describe these new features and provide a brief introduction to their scope and handling.

  2. Spectrum of SMPD1 mutations in Asian-Indian patients with acid sphingomyelinase (ASM)-deficient Niemann-Pick disease.

    Science.gov (United States)

    Ranganath, Prajnya; Matta, Divya; Bhavani, Gandham SriLakshmi; Wangnekar, Savita; Jain, Jamal Mohammed Nurul; Verma, Ishwar C; Kabra, Madhulika; Puri, Ratna Dua; Danda, Sumita; Gupta, Neerja; Girisha, Katta M; Sankar, Vaikom H; Patil, Siddaramappa J; Ramadevi, Akella Radha; Bhat, Meenakshi; Gowrishankar, Kalpana; Mandal, Kausik; Aggarwal, Shagun; Tamhankar, Parag Mohan; Tilak, Preetha; Phadke, Shubha R; Dalal, Ashwin

    2016-10-01

    Acid sphingomyelinase (ASM)-deficient Niemann-Pick disease is an autosomal recessive lysosomal storage disorder caused by biallelic mutations in the SMPD1 gene. To date, around 185 mutations have been reported in patients with ASM-deficient NPD world-wide, but the mutation spectrum of this disease in India has not yet been reported. The aim of this study was to ascertain the mutation profile in Indian patients with ASM-deficient NPD. We sequenced SMPD1 in 60 unrelated families affected with ASM-deficient NPD. A total of 45 distinct pathogenic sequence variants were found, of which 14 were known and 31 were novel. The variants included 30 missense, 4 nonsense, and 9 frameshift (7 single base deletions and 2 single base insertions) mutations, 1 indel, and 1 intronic duplication. The pathogenicity of the novel mutations was inferred with the help of the mutation prediction software MutationTaster, SIFT, Polyphen-2, PROVEAN, and HANSA. The effects of the identified sequence variants on the protein structure were studied using the structure modeled with the help of the SWISS-MODEL workspace program. The p. (Arg542*) (c.1624C>T) mutation was the most commonly identified mutation, found in 22% (26 out of 120) of the alleles tested, but haplotype analysis for this mutation did not identify a founder effect for the Indian population. To the best of our knowledge, this is the largest study on mutation analysis of patients with ASM-deficient Niemann-Pick disease reported in literature and also the first study on the SMPD1 gene mutation spectrum in India. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  3. Power Performance Test Report for the SWIFT Wind Turbine

    Energy Technology Data Exchange (ETDEWEB)

    Mendoza, I.; Hur, J.

    2012-12-01

    This report summarizes the results of a power performance test that NREL conducted on the SWIFT wind turbine. This test was conducted in accordance with the International Electrotechnical Commission's (IEC) standard, Wind Turbine Generator Systems Part 12: Power Performance Measurements of Electricity Producing Wind Turbines, IEC 61400-12-1 Ed.1.0, 2005-12. However, because the SWIFT is a small turbine as defined by IEC, NREL also followed Annex H that applies to small wind turbines. In these summary results, wind speed is normalized to sea-level air density.

  4. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  5. The Test-Retest Reliability of New Generation Power Indices of Wingate All-Out Test

    Directory of Open Access Journals (Sweden)

    Ozgur Ozkaya

    2018-04-01

    Full Text Available Although reliability correlations of traditional power indices of the Wingate test have been well documented, no study has analyzed new generation power indices based on milliseconds obtained from a Peak Bike. The purpose of this study was to investigate the retest reliability of new generation power indices. Thirty-two well-trained male athletes who were specialized in basketball, football, tennis, or track and field volunteered to take part in the study (age: 24.3 ± 2.2 years; body mass: 77 ± 8.3 kg; height: 180.3 ± 6.3 cm. Participants performed two Wingate all-out sessions on two separate days. Intra-class correlation coefficient (ICC, standard error measurement (SEM, smallest real differences (SRD and coefficient of variation (CV scores were analyzed based on the test and retest data. Reliability results of traditional power indices calculated based on 5-s means such as peak power, average power, power drop, and fatigue index ratio were similar with the previous findings in literature (ICC ≥ 0.94; CV ≤ 2.8%; SEM ≤ 12.28; SRD% ≤ 7.7%. New generation power indices such as peak power, average power, lowest power, power drop, fatigue index, power decline, maximum speed as rpm, and amount of total energy expenditure demonstrated high reliability (ICC ≥ 0.94; CV ≤ 4.3%; SEM ≤ 10.36; SRD% ≤ 8.8%. Time to peak power, time at maximum speed, and power at maximum speed showed a moderate level of reliability (ICC ≥ 0.73; CV ≤ 8.9%; SEM ≤ 63.01; SRD% ≤ 22.4%. The results of this study indicate that reliability correlations and SRD% of new generation power and fatigue-related indices are similar with traditional 5-s means. However, new time-related indices are very sensitive and moderately reliable.

  6. Power-cooling mismatch test series. Test PCM-2A; test results report

    International Nuclear Information System (INIS)

    Cawood, G.W.; Holman, G.W.; Martinson, Z.R.; Legrand, B.L.

    1976-09-01

    The report describes the results of an in-pile experimental investigation of pre- and postcritical heat flux (CHF) behavior of a single 36-inch-long, pressurized water reactor (PWR) type, UO 2 -fueled, zircaloy-clad fuel rod. The nominal coolant conditions for pressure and temperature were representative of those found in a commercial PWR. Nine separate departures from nucleate boiling (DNB) cycles were performed by either of two different methods: (a) decreasing the coolant flow rate while the fuel rod power was held constant, or (b) increasing the fuel rod power while the coolant flow rate was held constant. DNB was obtained during eight of the nine cycles performed. For the final flow reduction, the mass flux was decreased to 6.1 x 10 5 lb/hr-ft 2 at a constant peak linear heat generation rate of 17.8 kW/ft. The fuel rod was allowed to remain in film boiling for about 210 seconds during this final flow reduction. The fuel rod remained intact during the test. Results of on-line measurements of the fuel rod behavior are presented together with discussion of instrument performance. A comparison of the data with Fuel Rod Analysis Program-Transient 2 (FRAP-T2) computer code calculations is included

  7. A user's perspective on the merits and shortcomings of ASME Section III

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1994-01-01

    There are several aspects of Section III which when compared to process industry codes (ASME VIII, ASME B31.3, API, etc.) have proven to be a significant improvement in engineering practice. There are, however, other aspects of ASME III which have added to costs without clear benefits in safety or reliability. The authors present a user's perspective on some of the relative merits and shortcomings of the nuclear codes (ASME III and XI) compared to the process industry codes (such as ASME VIII, B31.3 and API)

  8. Components and system tests on the RFX toroidal power supply

    International Nuclear Information System (INIS)

    Toigo, V.; Zanotto, L.; Gaio, E.; Perna, M.; Bordignon, P.; Coffetti, A.; Novaro, R.; Bertolotto, P.; Rinaldi, E.; Villa, G.

    2005-01-01

    The paper deals with the component and system tests performed on the new toroidal power supply system of the RFX experiment. The high technological innovation of the system required a deep experimental characterization and validation campaign; special factory tests were performed on prototypes of single components aimed at verifying the most critical design aspects. Consequently an articulated series of tests were performed, based on a step-by-step approach to achieve the desired coordinate operation of the whole system. The test procedures and the most significant results are described in the paper

  9. High power test of low enriched UZrH

    Energy Technology Data Exchange (ETDEWEB)

    West, Gordon [General Atomic Co., San Diego, CA (United States)

    1980-07-01

    TRIGA-LEU fuel is currently undergoing high power tests in the 30 MW Oak Ridge Reactor. These tests are being funded by the Department of Energy through the RERTR program [Reduced Enrichment Research and Test Reactor program administered by Argonne National Laboratory] and began in mid-December, 1979 on a 16-rod shrouded cluster. The fuel rods are 0.51 in. 0D, clad with 0.16 in. Incoloy and the fuel length is 22 in. It is planned to test the UZrH fuel with 45, 30 and 20 wt-% U (nominal 20% enriched), to burnup values of about 50% of the contained U-235 in the 45 wt-% rods and about 40% and 35% burnup in the 30 wt-%, and 20 wt-% U fuel. It will take about 2 years of irradiation to produce the desired burnup in the 45 wt-% U fuel. Currently being tested are six 45 wt-% U and five 30 wt-% U rods. The remaining 5 rods are stainless steel dummies which were necessary to meet an operational requirement of the ORR which limits the power generation in a fuel rod to a value which would not raise the coolant temperature above the saturation level. Maximum calculated fuel rod powers were 40 kW, which would produce a fuel temperature of about 650 deg. C. The measured temperatures are about 400 deg. C and 350 deg. C for the 45 and 30 wt-% U fuel, respectively. Flow and {delta}T measurements show the cluster power generation to be about 250 kW, or about 65% of the design value. Reasons for the lower than expected power are still being evaluated and a proposal has been submitted for rearrangement of the fuel rods within the cluster to raise the powers and temperatures in the TRIGA-LEU fuel rods. (author)

  10. Advanced Grid Support Functionality Testing for Florida Power and Light

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, Austin [National Renewable Energy Lab. (NREL), Golden, CO (United States); Martin, Gregory [National Renewable Energy Lab. (NREL), Golden, CO (United States); Hurtt, James [Florida Power and Light, Juno Beach, FL (United States)

    2017-03-21

    This report describes the results of laboratory testing of advanced photovoltaic (PV) inverter testing undertaken by the National Renewable Energy Laboratory (NREL) on behalf of the Florida Power and Light Company (FPL). FPL recently commissioned a 1.1 MW-AC PV installation on a solar carport at the Daytona International Speedway in Daytona Beach, Florida. In addition to providing a source of clean energy production, the site serves as a live test bed with 36 different PV inverters from eight different manufacturers. Each inverter type has varied support for advanced grid support functions (GSFs) that are becoming increasingly commonplace, and are being required through revised interconnection standards such as UL1741, IEEE1547, and California (CA) Rule 21. FPL is interested in evaluating the trade-offs between different GSFs, their compliance to emerging standards, and their effects on efficiency and reliability. NREL has provided a controlled laboratory environment to undertake such a study. This work covered nine different classes of tests to compare inverter capabilities and performance for four different inverters that were selected by FPL. The test inverters were all three-phase models rated between 24-36 kW, and containing multiple PV input power point trackers. Advanced grid support functions were tested for functional behavior, and included fixed power factor operation, voltage-ride through, frequency ride-through, volt-var control, and frequency-Watt control. Response to abnormal grid conditions with GSFs enabled was studied through anti-islanding, fault, and load rejection overvoltage tests. Finally, efficiency was evaluated across a range of operating conditions that included power factor, output power, and input voltage variations. Test procedures were derived from requirements of a draft revision of UL741, CA Rule 21, and/or previous studies at NREL. This reports summarizes the results of each test case, providing a comparative performance analysis

  11. Test Report : GS Battery, EPC power HES RESCU

    Energy Technology Data Exchange (ETDEWEB)

    Rose, David Martin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Schenkman, Benjamin L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Borneo, Daniel R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2013-10-01

    The Department of Energy Office of Electricity (DOE/OE), Sandia National Laboratories (SNL) and the Base Camp Integration Lab (BCIL) partnered together to incorporate an energy storage system into a microgrid configured Forward Operating Base to reduce the fossil fuel consumption and to ultimately save lives. Energy storage vendors will be sending their systems to SNL Energy Storage Test Pad (ESTP) for functional testing and then to the BCIL for performance evaluation. The technologies that will be tested are electro-chemical energy storage systems comprising of lead acid, lithium-ion or zinc-bromide. GS Battery and EPC Power have developed an energy storage system that utilizes zinc-bromide flow batteries to save fuel on a military microgrid. This report contains the testing results and some limited analysis of performance of the GS Battery, EPC Power HES RESCU.

  12. Environmental qualification test of electrical penetration for nuclear power stations

    International Nuclear Information System (INIS)

    Kooziro, Tetsuya; Nakagawa, Akitoshi; Toyoda, Shigeru; Uno, Shunpei

    1979-01-01

    Environmental qualification test was conducted according to IEEE Std. 323-1974 in order to evaluate the safety and reliability of electrical penetration of PWR type nuclear power station. Electrical penetration is the assemblies of electric cables attached to the containment vessel and penetrate through the vessel. Since it is a part of the vessel, it is deemed to be one of the primary safety equipments that are important for the safety and reliability of nuclear power stations. Environmental tests were conducted continuously as to heat cycle, vibration and LOCA with the full size specimens of bushing type, pigtail type and triaxial cable type and at the same time thermal life and irradiation tests were conducted on the insulation materials used, in order to obtain the comprehensive evaluation of their electrical and mechanical characteristics. As the result, they all satisfied the requirements for the circuits for actual use during and after various environmental qualification tests according to IEEE Std. 323. (author)

  13. PUREX (SAMCONS) uninterruptible power supply (UPS) acceptance test procedure

    International Nuclear Information System (INIS)

    Blackaby, W.B.

    1997-01-01

    This Acceptance Test Procedure for the PUREX Surveillance and Monitoring and Control System (SAMCONS) Uninterruptible Power Supply (UPS) provides for testing and verifying the proper operation of the control panel alarms and trouble functions, the 6roper functioning of the AC inverter, ability of the battery supply to maintain the SAMCONS load for a minimum of two hours , and proper interaction with the SAMCONS Video graphic displays for alarm displays

  14. Comparative study of design of piping supports class 1, 2 and 3 considering german code KTA and ASME III - NF

    International Nuclear Information System (INIS)

    Faloppa, Altair A.; Fainer, Gerson; Mattar Neto, Miguel; Elias, Marcos V.

    2013-01-01

    The objective of this paper is developing a comparative study of the design criteria for class 1, 2, 3 piping supports considering the American Code ASME Section III - NF and the German Code KTA 3205.1 to the Primary Circuit, KTA 3205.2 to the others systems and KTA 3205.3 series-production standards supports of a PWR nuclear power plant. An additional purpose of the paper is a general analysis of the main design concepts of the American Code ASME Boiler and Pressure Vessel Code, Section III, Division 1 and German Nuclear Design Code KTA that was performed in order to aid the comparative study proposed. The relevance of this study is to show the differences between codes ASME and KTA since they were applied in the design of the Nuclear Power Plants Angra 1 and Angra 2, and to the design of Angra 3, which is at the moment under construction. It is also considered their use in the design of nuclear installations such as RMB - Reator MultiProposito Brasileiro and LABGENE - Laboratorio de Geracao Nucleoeletrica. (author)

  15. Using the noninformative families in family-based association tests : A powerful new testing strategy

    NARCIS (Netherlands)

    Lange, C; DeMeo, D; Silverman, EK; Weiss, ST; Laird, NM

    2003-01-01

    For genetic association studies with multiple phenotypes, we propose a new strategy for multiple testing with family-based association tests (FBATs). The strategy increases the power by both using all available family data and reducing the number of hypotheses tested while being robust against

  16. Recent development in the ASME O and M committee codes, standards, and guides

    International Nuclear Information System (INIS)

    Rowley, C.W.

    1999-01-01

    The ASME O and M Committee continues to expand and update its code, standards, and guides as contained in the ASME OM Code and the ASME OM Standards/Guides. This paper will describe recent changes to these two ASME documents, including technical inquiries, code cases, and the major reformat of the ASME OM Code 1998 Edition. Also two new Parts to the ASME OM S/G will be discussed: OM Part 23 and OM Part 24, which are close to being initially published. A third new Part to the ASME OM S/G has been authorized and has recently started to get organized: Part 26, 'Thermal Calibration of RTDs'. In addition this paper will describe the future plans for these two documents as provided in the O and M Committee Strategic Plan. (author)

  17. Energy saver A-sector power test results

    International Nuclear Information System (INIS)

    Martin, P.; Flora, R.; Tool, G.; Wolff, D.

    1982-01-01

    The superconducting magnets and associated cryogenic components in A-sector represent the initial phase of installation of the Fermilab superconducting accelerator, designed to accelerate proton beams to energies of 1 TeV. Installation of the magnets, comprising one-eighth of the ring, was completed in December, 1981. Cooldown and power tests took place in the first half of 1982, concurrent with main ring use for 400 GeV high energy physics. The tests described in this paper involved 151 cryogenic components in the tunnel: 94 dipoles, 24 quadrupoles, 25 spool pieces, 3 feed cans, 4 turn-around boxes and 1 bypass. Refrigeration was supplied by three satellite refrigerators, the Central Helium Liquefier, and two compressor buildings. The magnets were powered by a single power supply

  18. Junction temperature estimation for an advanced active power cycling test

    DEFF Research Database (Denmark)

    Choi, Uimin; Blaabjerg, Frede; Jørgensen, S.

    2015-01-01

    estimation method using on-state VCE for an advanced active power cycling test is proposed. The concept of the advanced power cycling test is explained first. Afterwards the junction temperature estimation method using on-state VCE and current is presented. Further, the method to improve the accuracy...... of the maximum junction temperature estimation is also proposed. Finally, the validity and effectiveness of the proposed method is confirmed by experimental results.......On-state collector-emitter voltage (VCE) is a good indicator to determine the wear-out condition of power device modules. Further, it is a one of the Temperature Sensitive Electrical Parameters (TSEPs) and thus can be used for junction temperature estimation. In this paper, the junction temperature...

  19. Dynamic testing of nuclear power plant structures: an evaluation

    International Nuclear Information System (INIS)

    Weaver, H.J.

    1980-02-01

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

  20. When the test of mediation is more powerful than the test of the total effect.

    Science.gov (United States)

    O'Rourke, Holly P; MacKinnon, David P

    2015-06-01

    Although previous research has studied power in mediation models, the extent to which the inclusion of a mediator will increase power has not been investigated. To address this deficit, in a first study we compared the analytical power values of the mediated effect and the total effect in a single-mediator model, to identify the situations in which the inclusion of one mediator increased statistical power. The results from this first study indicated that including a mediator increased statistical power in small samples with large coefficients and in large samples with small coefficients, and when coefficients were nonzero and equal across models. Next, we identified conditions under which power was greater for the test of the total mediated effect than for the test of the total effect in the parallel two-mediator model. These results indicated that including two mediators increased power in small samples with large coefficients and in large samples with small coefficients, the same pattern of results that had been found in the first study. Finally, we assessed the analytical power for a sequential (three-path) two-mediator model and compared the power to detect the three-path mediated effect to the power to detect both the test of the total effect and the test of the mediated effect for the single-mediator model. The results indicated that the three-path mediated effect had more power than the mediated effect from the single-mediator model and the test of the total effect. Practical implications of these results for researchers are then discussed.

  1. Design, Construction and Testing of an Uninterruptible Power ...

    African Journals Online (AJOL)

    Design, Construction and Testing of an Uninterruptible Power Supply of 300 Watts Capacity. ... is transformed using the transformer to 240v a.c., 50Hz. The system also has in-built voltage reversal protection, automatic change-over and battery ...

  2. Tests Of A Stirling-Engine Power Converter

    Science.gov (United States)

    Dochat, George

    1995-01-01

    Report describes acceptance tests of power converter consisting of pair of opposed free-piston Stirling engines driving linear alternators. Stirling engines offer potential for extremely long life, high reliability, high efficiency at low hot-to-cold temperature ratios, and relatively low heater-head temperatures.

  3. New design algorithm and reliability testing of solar powered near ...

    African Journals Online (AJOL)

    New design algorithm and reliability testing of solar powered near-space flight vehicle for defense and security. ... To overcome this problem, we propose a pseudo-satellite system where telecommunication devices are carried on a perpetually flying solar aircraft cruising at stratospheric altitude. Our aircraft will combine ...

  4. Modeling, Simulation, and Testing of Surf Kites for Power Generation

    NARCIS (Netherlands)

    Williams, P.; Lansdorp, B.; Ruiterkamp, R.; Ockels, W.J.

    2008-01-01

    Non-powered flight vehicles such as kites can provide a means of transmitting wind energy from higher altitudes to the ground via tethers. At Delft University of Technology, construction and testing of such a high altitude wind machine is ongoing. The concept is called the Laddermill. It generates

  5. Results of Laboratory Testing of Advanced Power Strips

    Energy Technology Data Exchange (ETDEWEB)

    Earle, L. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Sparn, B. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2012-08-01

    Presented at the ACEEE Summer Study on Energy Efficiency in Buildings on August 12-17, 2012, this presentation reports on laboratory tests of 20 currently available advanced power strip products, which reduce wasteful electricity use of miscellaneous electric loads in buildings.

  6. Bibliometric Analyses Reveal Patterns of Collaboration between ASMS Members

    Science.gov (United States)

    Palmblad, Magnus; van Eck, Nees Jan

    2018-03-01

    We have explored the collaborative network of the current American Society for Mass Spectrometry (ASMS) membership using bibliometric methods. The analysis shows that 4249 members are connected in a single, large, co-authorship graph, including the majority of the most published authors in the field of mass spectrometry. The map reveals topographical differences between university groups and national laboratories, and that the co-authors with the strongest links have long worked together at the same location. We have collected and summarized information on the geographical distribution of members, showing a high coverage of active researchers in North America and Western Europe. Looking at research fields, we could also identify a number of new or `hot' topics among ASMS members. Interactive versions of the maps are available on-line at https://goo.gl/UBNFMQ (collaborative network) and https://goo.gl/WV25vm (research topics). [Figure not available: see fulltext.

  7. ASM Student Technology and Career Night

    Science.gov (United States)

    Hamilton, Jeff

    2005-01-01

    This viewgraph presentation presents a general overview of Marshall Space Flight Center (MSFC) for students who are perspective MSFC employees. The presentation includes an organizational chart and a summary of MSFC activities, as well as photographs and descriptions of some of the center's test facilities.

  8. ASME Evaluation on Grid Mobile E-Commerce Process

    OpenAIRE

    Dan Chang; Wei Liao

    2012-01-01

    With the development of E-commerce, more scholars have paid attention to research on Mobile E-commerce and mostly focus on the optimization and evaluation of existing process. This paper researches the evaluation of Mobile E-commerce process with a method called ASME. Based on combing and analyzing current mobile business process and utilizing the grid management theory, mobile business process based on grid are constructed. Firstly, the existing process, namely Non-grid Mobile E-commerce, an...

  9. Abelian Sandpile Model (ASM) and Infinite Volume Limit

    Indian Academy of Sciences (India)

    ASM- Properties. Any possible sequence of topplings leads to the same stable configuration [Dhar]. The result of particle addition at and subsequent relaxation is given by an operator. £ бвд £ евд £. , where вд £. ¢. ¦. ¤ззз ¤ вг иг . £. ©. ¢ йа£. (Abelian). 7-b ...

  10. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  11. Graphical Tests for Power Comparison of Competing Designs.

    Science.gov (United States)

    Hofmann, H; Follett, L; Majumder, M; Cook, D

    2012-12-01

    Lineups have been established as tools for visual testing similar to standard statistical inference tests, allowing us to evaluate the validity of graphical findings in an objective manner. In simulation studies lineups have been shown as being efficient: the power of visual tests is comparable to classical tests while being much less stringent in terms of distributional assumptions made. This makes lineups versatile, yet powerful, tools in situations where conditions for regular statistical tests are not or cannot be met. In this paper we introduce lineups as a tool for evaluating the power of competing graphical designs. We highlight some of the theoretical properties and then show results from two studies evaluating competing designs: both studies are designed to go to the limits of our perceptual abilities to highlight differences between designs. We use both accuracy and speed of evaluation as measures of a successful design. The first study compares the choice of coordinate system: polar versus cartesian coordinates. The results show strong support in favor of cartesian coordinates in finding fast and accurate answers to spotting patterns. The second study is aimed at finding shift differences between distributions. Both studies are motivated by data problems that we have recently encountered, and explore using simulated data to evaluate the plot designs under controlled conditions. Amazon Mechanical Turk (MTurk) is used to conduct the studies. The lineups provide an effective mechanism for objectively evaluating plot designs.

  12. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

    Chi, Sung Goo; Kim, Se Chang; Seo, Jong Tae; Eom, Young Meen; Wook, Jeong Dae; Choi, Young Boo

    1995-01-01

    YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor POwer Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/ Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems

  13. The Power of Unit Root Tests Against Nonlinear Local Alternatives

    DEFF Research Database (Denmark)

    Demetrescu, Matei; Kruse, Robinson

    of Econometrics 112, 359-379) in comparison to the linear Dickey-Fuller test. To this end, we consider different adjustment schemes for deterministic terms. We provide asymptotic results which imply that the error variance has a severe impact on the behavior of the tests in the nonlinear case; the reason...... for such behavior is the interplay of nonstationarity and nonlinearity. In particular, we show that nonlinearity of the data generating process can be asymptotically negligible when the error variance is moderate or large (compared to the "amount of nonlinearity"), rendering the linear test more powerful than...

  14. Design Procedure of Graphite Components by ASME HTR Codes

    International Nuclear Information System (INIS)

    Kang, Ji-Ho; Jo, Chang Keun

    2016-01-01

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet

  15. Design Procedure of Graphite Components by ASME HTR Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.

  16. Investigating ASME allowable loads with finite element analyses

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Bezerra, Luciano M.; Miranda, Carlos A. de J.; Cruz, Julio R.B.

    1995-01-01

    The evaluation of nuclear components using finite element analysis (FEA) does not generally fall into the shell type verification adopted by the ASME Code. Consequently, the demonstration that the modes of failure are avoided sometimes is not straightforward. Allowable limits, developed by limit load theory, require the computation of shell membrane and bending stresses. How to calculate these stresses from FEA is not necessarily self-evident. One approach to be considered is to develop recommendations in a case-by-case basis for the most common pressure vessel geometries and loads based on comparison between the results of elastic and plastic FEA. In this paper, FE analyses of common 2D and complex 3D geometries are examined and discussed. It will be clear that in the cases studied, stress separation and categorization are not self-evident and simple tasks to undertake. Certain unclear recommendations of ASME Code can lead the stress analyst to non conservative designs as will be demonstrated in this paper. At the endo of this paper, taking into account comparison between elastic and elastic-plastic FE results from ANSYS some observations, suggestions and conclusions about the degree of conservatism of the ASME recommendations will be addressed. (author)

  17. Rethinking ASME III seismic analysis for piping operability evaluations

    International Nuclear Information System (INIS)

    Adams, T.M.; Stevenson, J.D.

    1994-01-01

    It has been recognized since the mid 1980's that there are very large seismic margins to failure for nuclear piping systems when designed using current industry practice, design criteria, and methods. As a result of this realization there are or have been approximately eighteen initiatives within the ASME , Boiler and Pressure Vessel Code Section III, Division 1, in the form of proposed code cases and proposed code text changes designed to reduce these failure margins to more realistic values. For the most part these initiatives have concentrated on reclassifying seismic inertia stresses in the piping as secondary and increasing the allowable stress limits permitted by Section III of the ASME, Boiler Code. This paper focuses on the application of non-linear spectral analysis methods as a method to reduce the input seismic demand determination and thereby reduce the seismic failure margins. The approach is evaluated using the ASME Boiler Pressure Vessel Code Section III Subgroup on Design benchmark procedure as proposed by the Subgroup's Special Task Group on Integrated Piping Criteria. Using this procedure, criteria are compared to current code criterion and analysis methods, and several other of the currently proposed Boiler and Pressure Vessel, Section III, changes. Finally, the applicability of the non-linear spectral analysis to continued Safe Operation Evaluations is reviewed and discussed

  18. 46 CFR 56.60-1 - Acceptable materials and specifications (replaces 123 and Table 126.1 in ASME B31.1).

    Science.gov (United States)

    2010-10-01

    ... Fittings and Other Piping Components—Magnetic Particle Examination Method. SP-55 Quality Standard for Steel... certification, use is limited to applications within heat exchangers. 12 Hydrostatic testing of these fittings... and Seamless Wrought Steel Pipe. ASME B36.19M 2004 Stainless Steel Pipe. American Society for Testing...

  19. The need to optimize inservice testing and inspection to enhance safety

    International Nuclear Information System (INIS)

    Perry, J.A.

    1996-01-01

    Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view

  20. The need to optimize inservice testing and inspection to enhance safety

    Energy Technology Data Exchange (ETDEWEB)

    Perry, J.A.

    1996-12-01

    Welcome to the Fourth U.S. Nuclear Regulatory Commission and American Society of Mechanical Engineers (USNRC/ASME) Symposium on Valve and Pump Testing in Nuclear Power Plants. This symposium provides a forum to exchange information on technical and regulatory issues associated with the testing of valves and pumps used in nuclear power plants. Progress made since the last symposium will be discussed along with various methods for in service testing of valves and pumps. Active participation by industry representatives, regulators and consultants will entail discussion of a broad array of ideas and points of view regarding how to improve the in service testing of valves and pumps at nuclear power plants. One of the challenges faced is the need to optimize the in service testing and inspection to enhance safety, operability and reliability. The author addresses this challenge from an ASME Nuclear Codes and Standards point of view.

  1. Power functional theory for the dynamic test particle limit

    International Nuclear Information System (INIS)

    Brader, Joseph M; Schmidt, Matthias

    2015-01-01

    For classical Brownian systems both in and out of equilibrium we extend the power functional formalism of Schmidt and Brader (2013 J. Chem. Phys. 138 214101) to mixtures of different types of particles. We apply the framework to develop an exact dynamical test particle theory for the self and distinct parts of the van Hove function, which characterize tagged and collective particle motion. The memory functions that induce non-Markovian dynamics are related to functional derivatives of the excess (over ideal) free power dissipation functional. The method offers an alternative to the recently found nonequilibrium Ornstein–Zernike relation for dynamic pair correlation functions. (paper)

  2. Radiological design criteria for fusion power test facilities

    International Nuclear Information System (INIS)

    Singh, M.S.; Campbell, G.W.

    1982-01-01

    The quest for fusion power and understanding of plasma physics has resulted in planning, design, and construction of several major fusion power test facilities, based largely on magnetic and inertial confinement concepts. We have considered radiological design aspects of the Joint European Torus (JET), Livermore Mirror and Inertial Fusion projects, and Princeton Tokamak. Our analyses on radiological design criteria cover acceptable exposure levels at the site boundary, man-rem doses for plant personnel and population at large, based upon experience gained for the fission reactors, and on considerations of cost-benefit analyses

  3. Small Radioisotope Power System Testing at NASA Glenn Research Center

    Science.gov (United States)

    Dugala, Gina; Bell, Mark; Oriti, Salvatore; Fraeman, Martin; Frankford, David; Duven, Dennis

    2013-01-01

    In April 2009, NASA Glenn Research Center (GRC) formed an integrated product team (IPT) to develop a Small Radioisotope Power System (SRPS) utilizing a single Advanced Stirling Convertor (ASC) with passive balancer. A single ASC produces approximately 80 We making this system advantageous for small distributed lunar science stations. The IPT consists of Sunpower, Inc., to provide the single ASC with a passive balancer, The Johns Hopkins University Applied Physics Laboratory (JHUAPL) to design an engineering model Single Convertor Controller (SCC) for an ASC with a passive balancer, and NASA GRC to provide technical support to these tasks and to develop a simulated lunar lander test stand. The single ASC with a passive balancer, simulated lunar lander test stand, and SCC were delivered to GRC and were tested as a system. The testing sequence at GRC included SCC fault tolerance, integration, electromagnetic interference (EMI), vibration, and extended operation testing. The SCC fault tolerance test characterized the SCCs ability to handle various fault conditions, including high or low bus power consumption, total open load or short circuit, and replacing a failed SCC card while the backup maintains control of the ASC. The integrated test characterized the behavior of the system across a range of operating conditions, including variations in cold-end temperature and piston amplitude, including the emitted vibration to both the sensors on the lunar lander and the lunar surface. The EMI test characterized the AC and DC magnetic and electric fields emitted by the SCC and single ASC. The vibration test confirms the SCCs ability to control the single ASC during launch. The extended operation test allows data to be collected over a period of thousands of hours to obtain long term performance data of the ASC with a passive balancer and the SCC. This paper will discuss the results of each of these tests.

  4. The statistical background to proposed ASME/MPC fracture toughness reference curves

    International Nuclear Information System (INIS)

    Oldfield, W.

    1981-01-01

    The ASME Pressure Vessel Codes define, in Sec. 11, lower bound fracture toughness curves. These curves are used to predict the lower bound fracture toughness on the basis of the RT test procedure. This test is used to remove heat to heat differences, by permitting the lower bound (reference) curve to be moved along the temperature scale according to the measured RT. Numerous objections have been raised to the procedure, and a Subcommittee (the ASME/MPC Working Group on Reference Toughness) is currently revising the codified procedures for fracture toughness prediction. The task has required a substantial amount of statistical work, since the new procedure are to have a statistical basis. Using initiation fracture toughness (J-Integral R curve procedures in the ductile domain) it was shown that when CVN energy data is properly transformed it is highly correlated with valid fracture toughness measurements. A single functional relationship can be used to predict the mean fracture toughness for a sample of steel from a set of CVN energy measurements, and the coefficients of the function tabulated. More importantly, the approximate lower statistical bounds to the initiation fracture toughness behaviour can be similarly predicted, and coefficients for selected bounds have also been tabulated. (orig.)

  5. ASTM and ASME-BPE Standards--Complying with the Needs of the Pharmaceutical Industry.

    Science.gov (United States)

    Huitt, William M

    2011-01-01

    Designing and building a pharmaceutical facility requires the owner, engineer of record, and constructor to be knowledgeable with regard to the industry codes and standards that apply to this effort. Up until 1997 there were no industry standards directed at the needs and requirements of the pharmaceutical industry. Prior to that time it was a patchwork effort at resourcing and adopting nonpharmaceutical-related codes and standards and then modifying them in order to meet the more stringent requirements of the Food and Drug Administration (FDA). In 1997 the American Society of Mechanical Engineers (ASME) published the first Bioprocessing Equipment (BPE) Standard. Through harmonization efforts this relatively new standard has brought together, scrutinized, and refined industry accepted methodologies together with FDA compliance requirements, and has established an American National Standard that provides a comprehensive set of standards that are integral to the pharmaceutical industry. This article describes various American National Standards, including those developed and published by the American Society for Testing and Materials (ASTM), and how they apply to the pharmaceutical industry. It goes on to discuss the harmonization effort that takes place between the various standards developers in an attempt to prevent conflicts and omissions between the many standards. Also included are examples of tables and figures taken from the ASME-BPE Standard. These examples provide the reader with insight to the relevant content of the ASME-BPE Standard. Designing and building a pharmaceutical facility requires the owner, engineer of record, and constructor to be knowledgeable with regard to the industry codes and standards that apply to this effort. Up until 1997 there were no industry standards directed at the needs and requirements of the pharmaceutical industry. Prior to that time it was a patchwork effort at resourcing and adopting nonpharmaceutical-related codes and

  6. Power Systems Development Facility Gasification Test Campaing TC18

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2005-08-31

    In support of technology development to utilize coal for efficient, affordable, and environmentally clean power generation, the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama, routinely demonstrates gasification technologies using various types of coals. The PSDF is an engineering scale demonstration of key features of advanced coal-fired power systems, including a KBR Transport Gasifier, a hot gas particulate control device (PCD), advanced syngas cleanup systems, and high pressure solids handling systems. This report details Test Campaign TC18 of the PSDF gasification process. Test campaign TC18 began on June 23, 2005, and ended on August 22, 2005, with the gasifier train accumulating 1,342 hours of operation using Powder River Basin (PRB) subbituminous coal. Some of the testing conducted included commissioning of a new recycle syngas compressor for gasifier aeration, evaluation of PCD filter elements and failsafes, testing of gas cleanup technologies, and further evaluation of solids handling equipment. At the conclusion of TC18, the PSDF gasification process had been operated for more than 7,750 hours.

  7. Fission Surface Power Technology Demonstration Unit Test Results

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Geng, Steven M.; Sanzi, James L.

    2016-01-01

    The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

  8. A shift from significance test to hypothesis test through power analysis in medical research.

    Science.gov (United States)

    Singh, G

    2006-01-01

    Medical research literature until recently, exhibited substantial dominance of the Fisher's significance test approach of statistical inference concentrating more on probability of type I error over Neyman-Pearson's hypothesis test considering both probability of type I and II error. Fisher's approach dichotomises results into significant or not significant results with a P value. The Neyman-Pearson's approach talks of acceptance or rejection of null hypothesis. Based on the same theory these two approaches deal with same objective and conclude in their own way. The advancement in computing techniques and availability of statistical software have resulted in increasing application of power calculations in medical research and thereby reporting the result of significance tests in the light of power of the test also. Significance test approach, when it incorporates power analysis contains the essence of hypothesis test approach. It may be safely argued that rising application of power analysis in medical research may have initiated a shift from Fisher's significance test to Neyman-Pearson's hypothesis test procedure.

  9. A shift from significance test to hypothesis test through power analysis in medical research

    Directory of Open Access Journals (Sweden)

    Singh Girish

    2006-01-01

    Full Text Available Medical research literature until recently, exhibited substantial dominance of the Fisher′s significance test approach of statistical inference concentrating more on probability of type I error over Neyman-Pearson′s hypothesis test considering both probability of type I and II error. Fisher′s approach dichotomises results into significant or not significant results with a P value. The Neyman-Pearson′s approach talks of acceptance or rejection of null hypothesis. Based on the same theory these two approaches deal with same objective and conclude in their own way. The advancement in computing techniques and availability of statistical software have resulted in increasing application of power calculations in medical research and thereby reporting the result of significance tests in the light of power of the test also. Significance test approach, when it incorporates power analysis contains the essence of hypothesis test approach. It may be safely argued that rising application of power analysis in medical research may have initiated a shift from Fisher′s significance test to Neyman-Pearson′s hypothesis test procedure.

  10. Evaluation of the integrity of reactor vessels designed to ASME Code, Sections I and/or VIII

    International Nuclear Information System (INIS)

    Hoge, K.G.

    1976-01-01

    A documented review of nuclear reactor pressure vessels designed to ASME Code, Sections I and/or VIII is made. The review is primarily concerned with the design specifications and quality assurance programs utilized for the reactor vessel construction and the status of power plant material surveillance programs, pressure-temperature operating limits, and inservice inspection programs. The following ten reactor vessels for light-water power reactors are covered in the report: Indian Point Unit No. 1, Dresden Unit No. 1, Yankee Rowe, Humboldt Bay Unit No. 3, Big Rock Point, San Onofre Unit No. 1, Connecticut Yankee, Oyster Creek, Nine Mile Point Unit No. 1, and La Crosse

  11. A review of Studsvik's international power ramp test projects

    International Nuclear Information System (INIS)

    Mogard, H.; Kjaer-Pedersen, N.

    1985-11-01

    Since 1975 a series of internationally sponsored fuel irradiation research projects have been and are being conducted at Studsvik, Sweden, under the management of Studsvik Energiteknik AB. The sponsoring parties comprise fuel vendors, nuclear power utilities, national research organizations and, in some cases, safety authorities. Geographically the parties represent organizations in Europe, Japan, and the USA. The main research topic of the Studsvik projects is the Pellet Clad Interaction (PCI) induced Stress Corrosion Cracking (SCC) failure occurrence in LWR fuel under power ramping conditions. The research is conducted in the 50 MW R2 test reactor and the associated hot cell laboratory. Prior to the experiments the test fuel is base irradiated, normally in commercially operated light water reactors. Results have been published for the INTER-RAMP, OVER-RAMP, DEMO-RAMP I, DEMO-RAMP II and SUPER-RAMP projects. The release of the TRANS-RAMP I results is imminent. There are two ongoing projects, i.e. SUPER-RAMP EXTENSION and TRANS-RAMP II. The paper presents an overview of the objectives and main results of the various projects. An attempt is made to summarize the more important observations on PCI failure performance in the perspective of design parameters, fuel burnup levels, power histories, power ramp rates, etc. With 14 refs. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-09-01

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

  13. Iterative and non-iterative solutions of engine flows using ASM and k-ε turbulence models

    International Nuclear Information System (INIS)

    Khaleghi, H.; Fallah, E.

    2003-01-01

    Various turbulent models are widely developed in order to make a good prediction of turbulence phenomena in different applications. The standard k-ε model shows a poor prediction for some applications. The Reynolds Stress Model (RSM) is expected to give a better prediction of turbulent characteristics, because a separate differential equation for each Reynolds stress component is solved in this model. In order to save both time and memory in this calculation a new Algebraic Stress Model (ASM) which was developed by Lumly et al in 1995 is used for calculations of flow characteristics in the internal combustion engine chamber. With using turbulent realizability principles, this model becomes a powerful and reliable turbulence model. In this paper the abilities of the model is examined in internal combustion engine flows. The results of ASM and k-ε models are compared with the experimental data. It is shown that the poor predictions of k-ε model are modified by ASM model. Also in this paper non-iterative PISO and iterative SIMPLE solution algorithms are compared. The results show that the PISO solution algorithm is the preferred and more efficient procedure in the calculation of internal combustion engine. (author)

  14. Estimation of sample size and testing power (part 6).

    Science.gov (United States)

    Hu, Liang-ping; Bao, Xiao-lei; Guan, Xue; Zhou, Shi-guo

    2012-03-01

    The design of one factor with k levels (k ≥ 3) refers to the research that only involves one experimental factor with k levels (k ≥ 3), and there is no arrangement for other important non-experimental factors. This paper introduces the estimation of sample size and testing power for quantitative data and qualitative data having a binary response variable with the design of one factor with k levels (k ≥ 3).

  15. Thermal performance test for steam turbine of nuclear power plants

    International Nuclear Information System (INIS)

    Bu Yubing; Xu Zongfu; Wang Shiyong

    2014-01-01

    Through study of steam turbine thermal performance test of CPR1000 nuclear power plant, we solve the enthalpy calculation problems of the steam turbine in wet steam zone using heat balance method which can help to figure out the real overall heat balance diagram for the first time, and we develop a useful software for thermal heat balance calculation. Ling'ao phase II as an example, this paper includes test instrument layout, system isolation, risk control, data acquisition, wetness measurement, heat balance calculation, etc. (authors)

  16. Pressure test behaviour of embalse nuclear power plant containment structure

    International Nuclear Information System (INIS)

    Bruschi, S.; Marinelli, C.

    1984-01-01

    It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

  17. Testing Header Component of Electricity Power Industry Boiler

    International Nuclear Information System (INIS)

    Soedardjo, S.A; Andryansyah, B; Artahari, Dewi; Natsir, Muhammad; Triyadi, Ari; Farokhi

    2000-01-01

    Testing of header component of Suralaya Unit II electricity power by replication method has been carried out. That header component is cross over pipe which interconnection between Primary and Superheater Outlet Header Secondary Superheater Outlet Header with the operation time over 14 years. The main composition of cross over pipe is 2 1/4 Cr 1 Mo or frequently specified as ferritique steel. The replication testing shown that the damage classification on those cross over pipe in A class based on failure classification from Neubauer and Wedel. Simple calculation in favor of cross over pipe remaining lifetime is about 16.5 years moreover

  18. Power Systems Development Facility Gasification Test Run TC11

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2003-04-30

    This report discusses Test Campaign TC11 of the Kellogg Brown & Root, Inc. (KBR) Transport Gasifier train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Gasifier is an advanced circulating fluidized-bed gasifier designed to operate as either a combustor or a gasifier in air- or oxygen-blown mode of operation using a particulate control device (PCD). Test run TC11 began on April 7, 2003, with startup of the main air compressor and the lighting of the gasifier start-up burner. The Transport Gasifier operated until April 18, 2003, when a gasifier upset forced the termination of the test run. Over the course of the entire test run, gasifier temperatures varied between 1,650 and 1,800 F at pressures from 160 to 200 psig during air-blown operations and around 135 psig during enriched-air operations. Due to a restriction in the oxygen-fed lower mixing zone (LMZ), the majority of the test run featured air-blown operations.

  19. Analysis of the ASME methodology for evaluation of erosion-corrosion defects with respect to the differences in the calculation and in materials used at the Dukovany NPP

    International Nuclear Information System (INIS)

    Kadecka, P.

    1995-01-01

    The problem of evaluation of tolerable defects and thinning of pipe walls was analyzed. In fact, a procedure for evaluation of tolerable defects is described in ASME Code Case N 480 based on the ASME ''Rules for Construction of Nuclear Power Plant Components''. The pipe systems of the Dukovany NPP, however, were constructed to different (East European) standards, and therefore caution should be exercised when applying US standards to this plant. The report demonstrates major differences between the ASME Standard and the proposed Czech standard ''A.S.I. Standards Documentation for Strength Calculations of Equipment and Piping of WWER Type Nuclear Power Plants'' developed by the Czech Association of Mechanical Engineers (A.S.I), evaluates the applicability of Code Case N 480 to the Dukovany plant, and proposes a Czech procedure for the evaluation. The basic characteristics of materials cited by ASME II and carbon steels used in the secondary circuit of the Dukovany NPP are also compared. (P.A.). 78 tabs., 2 figs., 4 refs

  20. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  1. Power ramp tests of high burnup BWR segment rods

    International Nuclear Information System (INIS)

    Hayashi, H.; Etoh, Y.; Tsukuda, Y.; Shimada, S.; Sakurai, H.

    2002-01-01

    Lead use assemblies (LUAs) of high burnup 8x8 fuel design for Japanese BWRs were irradiated up to 5 cycles in Fukushima Daini Nuclear Power Station No. 2 Unit. Segment rods were installed in LUAs and used for power ramp tests in Japanese Material Test Reactor (JMTR). Post irradiation examinations (PIEs) of segment rods were carried out at Nippon Nuclear Fuel Development Co., Ltd. before and after ramp tests. Maximum linear heat rates of LUAs were kept above 300 W/cm in the first cycle, above 250 W/cm in the second and third cycles and decreased to 200 W/cm in the fourth cycle and 80 W/cm in the fifth cycle. The integrity of high burnup 8x8 fuel was confirmed up to the bundle burnup of 48 GWd/t after 5 cycles of irradiation. Systematic and high quality data were collected through detailed PIEs. The main results are as follows. The oxide on the outer surface of cladding tubes was uniform and its thickness was less than 20 micro-meter after 5 cycles of irradiation and was almost independent of burnup. Hydrogen contents in cladding tubes were less than 150 ppm after 5 cycles of irradiation, although hydrogen contents increased during the fourth and fifth irradiation cycles. Mechanical properties of cladding tubes were on the extrapolated line of previous data up to 5 cycles of irradiation. Fission gas release rates were in the low level (mainly less than 6%) up to 5 cycles of irradiation due to the design to decrease pellet temperature. Pellet-cladding bonding layers were observed after the third cycle and almost full bonding was observed after the fifth cycle. Pellet volume increased with burnup in proportion to solid swelling rate up to the forth cycle. After the fifth cycle, slightly higher pellet swelling was confirmed. Power ramp tests were carried out and satisfactory performance of Zr-lined cladding tube was confirmed up to 60 GWd/t (segment average burnup). One segment rod irradiated for 3 cycles failed by a single step ramp test at terminal ramp power of 614 W

  2. Power Systems Development Facility Gasification Test Campaign TC25

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2008-12-01

    In support of technology development to utilize coal for efficient, affordable, and environmentally clean power generation, the Power Systems Development Facility (PSDF), located in Wilsonville, Alabama, routinely demonstrates gasification technologies using various types of coals. The PSDF is an engineering scale demonstration of key features of advanced coal-fired power systems, including a KBR Transport Gasifier, a hot gas particulate control device, advanced syngas cleanup systems, and high-pressure solids handling systems. This report summarizes the results of TC25, the second test campaign using a high moisture lignite coal from the Red Hills mine in Mississippi as the feedstock in the modified Transport Gasifier configuration. TC25 was conducted from July 4, 2008, through August 12, 2008. During TC25, the PSDF gasification process operated for 742 hours in air-blown gasification mode. Operation with the Mississippi lignite was significantly improved in TC25 compared to the previous test (TC22) with this fuel due to the addition of a fluid bed coal dryer. The new dryer was installed to dry coals with very high moisture contents for reliable coal feeding. The TC25 test campaign demonstrated steady operation with high carbon conversion and optimized performance of the coal handling and gasifier systems. Operation during TC25 provided the opportunity for further testing of instrumentation enhancements, hot gas filter materials, and advanced syngas cleanup technologies. The PSDF site was also made available for testing of the National Energy Technology Laboratory's fuel cell module and Media Process Technology's hydrogen selective membrane with syngas from the Transport Gasifier.

  3. Development and testing of improved statistical wind power forecasting methods.

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, J.; Bessa, R.J.; Keko, H.; Sumaili, J.; Miranda, V.; Ferreira, C.; Gama, J.; Botterud, A.; Zhou, Z.; Wang, J. (Decision and Information Sciences); (INESC Porto)

    2011-12-06

    Wind power forecasting (WPF) provides important inputs to power system operators and electricity market participants. It is therefore not surprising that WPF has attracted increasing interest within the electric power industry. In this report, we document our research on improving statistical WPF algorithms for point, uncertainty, and ramp forecasting. Below, we provide a brief introduction to the research presented in the following chapters. For a detailed overview of the state-of-the-art in wind power forecasting, we refer to [1]. Our related work on the application of WPF in operational decisions is documented in [2]. Point forecasts of wind power are highly dependent on the training criteria used in the statistical algorithms that are used to convert weather forecasts and observational data to a power forecast. In Chapter 2, we explore the application of information theoretic learning (ITL) as opposed to the classical minimum square error (MSE) criterion for point forecasting. In contrast to the MSE criterion, ITL criteria do not assume a Gaussian distribution of the forecasting errors. We investigate to what extent ITL criteria yield better results. In addition, we analyze time-adaptive training algorithms and how they enable WPF algorithms to cope with non-stationary data and, thus, to adapt to new situations without requiring additional offline training of the model. We test the new point forecasting algorithms on two wind farms located in the U.S. Midwest. Although there have been advancements in deterministic WPF, a single-valued forecast cannot provide information on the dispersion of observations around the predicted value. We argue that it is essential to generate, together with (or as an alternative to) point forecasts, a representation of the wind power uncertainty. Wind power uncertainty representation can take the form of probabilistic forecasts (e.g., probability density function, quantiles), risk indices (e.g., prediction risk index) or scenarios

  4. Radiation Testing of PICA at the Solar Power Tower

    Science.gov (United States)

    White, Susan

    2010-01-01

    Sandia National Laboratory's Solar Power Tower was used to irradiate specimens of Phenolic Impregnated Carbon Ablator (PICA), in order to evaluate whether this thermal protection system material responded differently to potential shock layer radiative heating than to convective heating. Tests were run at 50, 100 and 150 Watts per square centimeter levels of concentrated solar radiation. Experimental results are presented both from spectral measurements on 1- 10 mm thick specimens of PICA, as well as from in-depth temperature measurements on instrumented thicker test specimens. Both spectral measurements and measured in-depth temperature profiles showed that, although it is a porous, low-density material, PICA does not exhibit problematic transparency to the tested high levels of NIR radiation, for all pragmatic cm-to-inch scale thicknesses. PICA acted as a surface absorber to efficiently absorb the incident visible and near infrared incident radiation in the top 2 millimeter layer in the Solar Power Tower tests up to 150 Watts per square centimeter.

  5. Artificial enzyme-powered microfish for water-quality testing.

    Science.gov (United States)

    Orozco, Jahir; García-Gradilla, Victor; D'Agostino, Mattia; Gao, Wei; Cortés, Allan; Wang, Joseph

    2013-01-22

    We present a novel micromotor-based strategy for water-quality testing based on changes in the propulsion behavior of artificial biocatalytic microswimmers in the presence of aquatic pollutants. The new micromotor toxicity testing concept mimics live-fish water testing and relies on the toxin-induced inhibition of the enzyme catalase, responsible for the biocatalytic bubble propulsion of tubular microengines. The locomotion and survival of the artificial microfish are thus impaired by exposure to a broad range of contaminants, that lead to distinct time-dependent irreversible losses in the catalase activity, and hence of the propulsion behavior. Such use of enzyme-powered biocompatible polymeric (PEDOT)/Au-catalase tubular microengine offers highly sensitive direct optical visualization of changes in the swimming behavior in the presence of common contaminants and hence to a direct real-time assessment of the water quality. Quantitative data on the adverse effects of the various toxins upon the swimming behavior of the enzyme-powered artificial swimmer are obtained by estimating common ecotoxicological parameters, including the EC(50) (exposure concentration causing 50% attenuation of the microfish locomotion) and the swimmer survival time (lifetime expectancy). Such novel use of artificial microfish addresses major standardization and reproducibility problems as well as ethical concerns associated with live-fish toxicity assays and hence offers an attractive alternative to the common use of aquatic organisms for water-quality testing.

  6. Testing for purchasing power parity in 21 African countries using several unit root tests

    Science.gov (United States)

    Choji, Niri Martha; Sek, Siok Kun

    2017-04-01

    Purchasing power parity is used as a basis for international income and expenditure comparison through the exchange rate theory. However, empirical studies show disagreement on the validity of PPP. In this paper, we conduct the testing on the validity of PPP using panel data approach. We apply seven different panel unit root tests to test the validity of the purchasing power parity (PPP) hypothesis based on the quarterly data on real effective exchange rate for 21 African countries from the period 1971: Q1-2012: Q4. All the results of the seven tests rejected the hypothesis of stationarity meaning that absolute PPP does not hold in those African Countries. This result confirmed the claim from previous studies that standard panel unit tests fail to support the PPP hypothesis.

  7. Extended Pulse-Powered Humidity-Freeze Cycling for Testing Module-Level Power Electronics

    Energy Technology Data Exchange (ETDEWEB)

    Hacke, Peter L [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Rodriguez, Miguel [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Kempe, Michael D [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Repins, Ingrid L [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-11-28

    An EMI suppression capacitor (polypropylene film type) failed by 'popcorning' due to vapor outgassing in pulse powered humidity-freeze cycles. No shorts or shunts could be detected despite mildly corroded metallization visible in the failed capacitor. Humidity-freeze cycling is optimized to break into moisture barriers. However, further studies will be required on additional module level power electronic (MLPE) devices to optimize the stress testing for condensation to precipitate any weakness to short circuiting and other humidity/bias failure modes.

  8. Why nuclear power failed the market test in the UK

    International Nuclear Information System (INIS)

    Chesshire, J.

    1992-01-01

    The Conservative Party's manifesto for the general election of May 1987 contained two pledges of relevance to the UK electricity supply industry (ESI). These were to privatize the industry; and to continue to support the development of civil nuclear power in the private sector. As anticipated by some independent commentators, in the event these objectives proved incompatible. The costs of nuclear power have long been a vexed issue and UK nuclear costs have been higher than those in many other countries. While most of the UK ESI has now been privatized, nuclear generation remains in the public sector. This article seeks to explore the reasons for this fundamental and politically embarrassing policy reversal, a rarity under three successive Conservative administrations since 1979. It would be incorrect to argue that private ownership and nuclear power are inherently incompatible. Rather the specific - competitive - form of privatization proposed for the UK failed to provide sufficient guarantees for the London capital market. Thus, at least in this specific case, nuclear power failed the market test. The implications of this for the UK nuclear industry have been profound. As a result, the UK case has wider international lessons as the pressures for privatization, liberalization and greater cost transparency bear down upon electric utilities in other countries. (author)

  9. ASM-024, a piperazinium compound, promotes the in vitro relaxation of β2-adrenoreceptor desensitized tracheas.

    Science.gov (United States)

    Israël-Assayag, Evelyne; Beaulieu, Marie-Josée; Cormier, Yvon

    2015-01-01

    Inhaled β2-adrenoreceptor agonists are widely used in asthma and chronic obstructive pulmonary disease (COPD) for bronchoconstriction relief. β2-Adrenoreceptor agonists relax airway smooth muscle cells via cyclic adenosine monophosphate (cAMP) mediated pathways. However, prolonged stimulation induces functional desensitization of the β2-adrenoreceptors (β2-AR), potentially leading to reduced clinical efficacy with chronic or prolonged administration. ASM-024, a small synthetic molecule in clinical stage development, has shown activity at the level of nicotinic receptors and possibly at the muscarinic level and presents anti-inflammatory and bronchodilator properties. Aerosolized ASM-024 reduces airway resistance in mice and promotes in-vitro relaxation of tracheal and bronchial preparations from animal and human tissues. ASM-024 increased in vitro relaxation response to maximally effective concentration of short-acting beta-2 agonists in dog and human bronchi. Although the precise mechanisms by which ASM-024 promotes airway smooth muscle (ASM) relaxation remain unclear, we hypothesized that ASM-024 will attenuate and/or abrogate agonist-induced contraction and remain effective despite β2-AR tachyphylaxis. β2-AR tachyphylaxis was induced with salbutamol, salmeterol and formoterol on guinea pig tracheas. The addition of ASM-024 relaxed concentration-dependently intact or β2-AR desensitized tracheal rings precontracted with methacholine. ASM-024 did not induce any elevation of intracellular cAMP in isolated smooth muscle cells; moreover, blockade of the cAMP pathway with an adenylate cyclase inhibitor had no significant effect on ASM-024-induced guinea pig trachea relaxation. Collectively, these findings show that ASM-024 elicits relaxation of β2-AR desensitized tracheal preparations and suggest that ASM-024 mediates smooth muscle relaxation through a different target and signaling pathway than β2-adrenergic receptor agonists. These findings suggest ASM-024

  10. Computer Program of SIE ASME-NH (Revision 1.0) Code

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2008-01-15

    In this report, the SIE ASME (Structural Integrity Evaluations by ASME-NH) (Revision 1.0), which has a computerized implementation of ASME Pressure Vessels and Piping Code Section III Subsection NH rules, is developed to apply to the next generation reactor design subjecting to the elevated temperature operations over 500 .deg. C and over 30 years design lifetime, and the user's manual for this program is described in detail.

  11. Computer Program of SIE ASME-NH (Revision 1.0) Code

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, J. H.

    2008-01-01

    In this report, the SIE ASME (Structural Integrity Evaluations by ASME-NH) (Revision 1.0), which has a computerized implementation of ASME Pressure Vessels and Piping Code Section III Subsection NH rules, is developed to apply to the next generation reactor design subjecting to the elevated temperature operations over 500 .deg. C and over 30 years design lifetime, and the user's manual for this program is described in detail

  12. RF Power Requirements for PEFP SRF Cavity Test

    International Nuclear Information System (INIS)

    Kim, Han Sung; Seol, Kyung Tae; Kwon, Hyeok Jung; Cho, Yong Sub

    2011-01-01

    For the future extension of the PEFP (Proton Engineering Frontier Project) Proton linac, preliminary study on the SRF (superconducting radio-frequency) cavity is going on including a five-cell prototype cavity development to confirm the design and fabrication procedures and to check the RF and mechanical properties of a low-beta elliptical cavity. The main parameters of the cavity are like followings. - Frequency: 700 MHz - Operating mode: TM010 pi mode - Cavity type: Elliptical - Geometrical beta: 0.42 - Number of cells: 5 - Accelerating gradient: 8 MV/m - Epeak/Eacc: 3.71 - Bpeak/Eacc: 7.47 mT/(MV/m) - R/Q: 102.3 ohm - Epeak: 29.68 MV/m (1.21 Kilp.) - Geometrical factor: 121.68 ohm - Cavity wall thickness: 4.3 mm - Stiffening structure: Double ring - Effective length: 0.45 m For the test of the cavity at low temperature of 4.2 K, many subsystems are required such as a cryogenic system, RF system, vacuum system and radiation shielding. RF power required to generate accelerating field inside cavity depends on the RF coupling parameters of the power coupler and quality factor of the SRF cavity and the quality factor itself is affected by several factors such as operating temperature, external magnetic field level and surface condition. Therefore, these factors should be considered to estimate the required RF power for the SRF cavity test

  13. Technical Review on Fitness-for-Service for Buried Pipe by ASME Code Case N-806

    International Nuclear Information System (INIS)

    Park, Sang Kyu; Lee, Yo Seop; So, Il Su; Lim, Bu Taek

    2012-01-01

    Fitness-for-Service is a useful technology to determine replacement timing, next inspection timing or in-service when nuclear power plant's buried pipes are damaged. If is possible for buried pipes to be aged by material loss, cracks and occlusion as operating time goes by. Therefore Fitness-for-Service technology for buried pipe is useful for plant industry to perform replacement and repair. Fitness-for-Service for buried pipe is studied in terms of existing code and standard for Fitness-for-Service and a current developing code case. Fitness-for-Service for buried pipe was performed according to Code Case N-806 developed by ASME (American Society of Mechanical Engineers)

  14. 3D automatic anatomy segmentation based on iterative graph-cut-ASM.

    Science.gov (United States)

    Chen, Xinjian; Bagci, Ulas

    2011-08-01

    This paper studies the feasibility of developing an automatic anatomy segmentation (AAS) system in clinical radiology and demonstrates its operation on clinical 3D images. The AAS system, the authors are developing consists of two main parts: object recognition and object delineation. As for recognition, a hierarchical 3D scale-based multiobject method is used for the multiobject recognition task, which incorporates intensity weighted ball-scale (b-scale) information into the active shape model (ASM). For object delineation, an iterative graph-cut-ASM (IGCASM) algorithm is proposed, which effectively combines the rich statistical shape information embodied in ASM with the globally optimal delineation capability of the GC method. The presented IGCASM algorithm is a 3D generalization of the 2D GC-ASM method that they proposed previously in Chen et al. [Proc. SPIE, 7259, 72590C1-72590C-8 (2009)]. The proposed methods are tested on two datasets comprised of images obtained from 20 patients (10 male and 10 female) of clinical abdominal CT scans, and 11 foot magnetic resonance imaging (MRI) scans. The test is for four organs (liver, left and right kidneys, and spleen) segmentation, five foot bones (calcaneus, tibia, cuboid, talus, and navicular). The recognition and delineation accuracies were evaluated separately. The recognition accuracy was evaluated in terms of translation, rotation, and scale (size) error. The delineation accuracy was evaluated in terms of true and false positive volume fractions (TPVF, FPVF). The efficiency of the delineation method was also evaluated on an Intel Pentium IV PC with a 3.4 GHZ CPU machine. The recognition accuracies in terms of translation, rotation, and scale error over all organs are about 8 mm, 10 degrees and 0.03, and over all foot bones are about 3.5709 mm, 0.35 degrees and 0.025, respectively. The accuracy of delineation over all organs for all subjects as expressed in TPVF and FPVF is 93.01% and 0.22%, and all foot bones for

  15. 3D automatic anatomy segmentation based on iterative graph-cut-ASM

    International Nuclear Information System (INIS)

    Chen, Xinjian; Bagci, Ulas

    2011-01-01

    Purpose: This paper studies the feasibility of developing an automatic anatomy segmentation (AAS) system in clinical radiology and demonstrates its operation on clinical 3D images. Methods: The AAS system, the authors are developing consists of two main parts: object recognition and object delineation. As for recognition, a hierarchical 3D scale-based multiobject method is used for the multiobject recognition task, which incorporates intensity weighted ball-scale (b-scale) information into the active shape model (ASM). For object delineation, an iterative graph-cut-ASM (IGCASM) algorithm is proposed, which effectively combines the rich statistical shape information embodied in ASM with the globally optimal delineation capability of the GC method. The presented IGCASM algorithm is a 3D generalization of the 2D GC-ASM method that they proposed previously in Chen et al.[Proc. SPIE, 7259, 72590C1-72590C-8 (2009)]. The proposed methods are tested on two datasets comprised of images obtained from 20 patients (10 male and 10 female) of clinical abdominal CT scans, and 11 foot magnetic resonance imaging (MRI) scans. The test is for four organs (liver, left and right kidneys, and spleen) segmentation, five foot bones (calcaneus, tibia, cuboid, talus, and navicular). The recognition and delineation accuracies were evaluated separately. The recognition accuracy was evaluated in terms of translation, rotation, and scale (size) error. The delineation accuracy was evaluated in terms of true and false positive volume fractions (TPVF, FPVF). The efficiency of the delineation method was also evaluated on an Intel Pentium IV PC with a 3.4 GHZ CPU machine. Results: The recognition accuracies in terms of translation, rotation, and scale error over all organs are about 8 mm, 10 deg. and 0.03, and over all foot bones are about 3.5709 mm, 0.35 deg. and 0.025, respectively. The accuracy of delineation over all organs for all subjects as expressed in TPVF and FPVF is 93.01% and 0.22%, and

  16. Power Systems Development Facility Gasification Test Campaign TC24

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2008-03-30

    In support of technology development to utilize coal for efficient, affordable, and environmentally clean power generation, the Power Systems Development Facility (PSDF), located in Wilsonville, Alabama, routinely demonstrates gasification technologies using various types of coals. The PSDF is an engineering scale demonstration of key features of advanced coal-fired power systems, including a KBR Transport Gasifier, a hot gas particulate control device, advanced syngas cleanup systems, and high-pressure solids handling systems. This report summarizes the results of TC24, the first test campaign using a bituminous coal as the feedstock in the modified Transport Gasifier configuration. TC24 was conducted from February 16, 2008, through March 19, 2008. The PSDF gasification process operated for about 230 hours in air-blown gasification mode with about 225 tons of Utah bituminous coal feed. Operational challenges in gasifier operation were related to particle agglomeration, a large percentage of oversize coal particles, low overall gasifier solids collection efficiency, and refractory degradation in the gasifier solids collection unit. The carbon conversion and syngas heating values varied widely, with low values obtained during periods of low gasifier operating temperature. Despite the operating difficulties, several periods of steady state operation were achieved, which provided useful data for future testing. TC24 operation afforded the opportunity for testing of various types of technologies, including dry coal feeding with a developmental feeder, the Pressure Decoupled Advanced Coal (PDAC) feeder; evaluating a new hot gas filter element media configuration; and enhancing syngas cleanup with water-gas shift catalysts. During TC24, the PSDF site was also made available for testing of the National Energy Technology Laboratory's fuel cell module and Media Process Technology's hydrogen selective membrane.

  17. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  18. Guidelines for inservice testing at nuclear power plants

    International Nuclear Information System (INIS)

    Campbell, P.

    1995-04-01

    The staff of the U.S. Nuclear Regulatory Commission (NRC) gives licensees guidelines and recommendations for developing and implementing programs for the inservice testing of pumps and valves at commercial nuclear power plants. The staff discusses the regulations; the components to be included in an inservice testing program; and the preparation and content of cold shutdown justifications, refueling outage justifications, and requests for relief from the American Society of Mechanical Engineers Code requirements. The staff also gives specific guidance on relief acceptable to the NRC and advises licensees in the use of this information at their facilities. The staff discusses the revised standard technical specifications for the inservice testing program requirements and gives guidance on the process a licensee may follow upon finding an instance of noncompliance with the Code

  19. Results of gap conductance tests in the power burst facility

    International Nuclear Information System (INIS)

    Garner, R.W.; Sparks, D.T.

    1977-01-01

    Light water reactor (LWR) fuel rod behavior studies are being conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc. These studies are being performed under contract to the Energy Research and Development Adminstration at the Idaho National Engineering Laboratory (INEL), as part of the Nuclear Regulatory Commission's Water Reactor Safety Research Fuel Behavior Program. Experimental data for verification of analytical models developed to predict light water nuclear fuel rod behavior under normal and postulated accident conditions are being obtained from a variety of in-reactor and out-of-reactor experiments. This paper summarizes the results of tests performed in the Power Burst Facility (PBF) to obtain data from which the thermal response, gap conductance, and stored energy of LWR fuel rods can be determined. Primary objectives of the PBF gap conductance test program are (a) to obtain data on a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rod designs, under a wide range of operating conditions, from which gap conductance values can be determined and (b) to evaluate experimentally the power oscillation method for measuring the gap conductance and thermal response of a fresh or burned LWR fuel rod. Tests have been performed with both irradiated and unirradiated PWR-type fuel and with fresh BWR-type fuel rods. Some PWR rod test results are described, and the thermal response data from BWR rod tests are discussed in greater detail. Comparisons are made of gap conductance values determined by the tests with analytically calculated values using the Fuel Rod Analysis Program-Transient (FRAP-T) computer code. These comparisons provide insight into both the experimental measurements methods and the validity of the gap conductance models

  20. Development of ASME Code Section 11 visual examination requirements

    International Nuclear Information System (INIS)

    Cook, J.F.

    1990-01-01

    Section XI of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) defines three types of nondestructive examinations, visual, surface, and volumetric. Visual examination is important since it is the primary examination method for many safety-related components and systems and is also used as a backup examination for the components and systems which receive surface or volumetric examinations. Recent activity in the Section XI Code organization to improve the rules for visual examinations is reviewed and the technical basis for the new rules, which cover illumination, vision acuity, and performance demonstration, is explained

  1. Long periodicity of Blazar with RXTE ASM, TA and HEGRA

    OpenAIRE

    Osone, S.; Teshima, M.

    2002-01-01

    Long periodicity for Mkn501 during a large flare in 1997 have been reported by TA, HEGRA group. Here, we establish this periodicity with archival data of RXTE All Sky Monitor(ASM), Telescope Array(TA) and HEGRA with a chance probability less than $10^{-5}$. We also find that an origin of 23 day periodicity is related with a change of either a gamma factor of electrons $\\gamma$ or the magnetic field or a beaming factor. And, in order to search for a category which have a long periodicity, we m...

  2. Sizing and preliminary hardware testing of solar powered UAV

    Directory of Open Access Journals (Sweden)

    S. Jashnani

    2013-12-01

    Full Text Available Integrating solar energy into modern aircraft technology has been a topic of interest and has received a lot of attention from researchers over the last two decades. A few among the many potential applications of this technology are the possibility of continuous self sustained flight for purposes such as information relay, surveillance and monitoring. This paper discusses the altitude and payload mass, as independent parameters, and their influence on the size and design of the aircraft. To estimate available solar power, two different models have been presented; one for low altitudes and the other for high altitudes. An engineering ground model was built to simulate the power and propulsion system over 24 h of continuous operation. The paper presents data from tests performed till date and lessons learnt while dealing with the construction of the engineering ground model as well as changes that can be made to improve the design.

  3. Tests of cooling water pumps at Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Travnicek, J.

    1986-01-01

    Tests were performed to examine the operating conditions of the 1600 BQDV cooling pumps of the main coolant circuit of unit 1 of the Dukovany nuclear power plant. For the pumps, the performance was tested in the permissible operating range, points were measured below this range and the guaranteed operating point was verified. Pump efficiency was calculated from the measured values. The discussion of the measurement of parameters has not yet been finished because the obtained values of the amount delivered and thus of the pump efficiency were not up to expectation in all detail. It was also found that for obtaining the guaranteed flow the pump impeller had to be opened to 5deg -5.5deg instead of the declared 3deg. Also tested were pump transients, including the start of the pump, its stop, the operation and failure of one of the two pumps. In these tests, pressures were also measured at the inlet and the outlet of the inner part of the TG 11 turbine condenser. It was shown that the time course and the pressure course of the processes were acceptable. In addition to these tests, pressure losses in the condenser and the cooling water flow through the feed pump electromotor cooler wre tested for the case of a failure of one of the two pumps. (E.S.)

  4. Test results for cables used in nuclear power plants by a new environmental testing method

    Energy Technology Data Exchange (ETDEWEB)

    Handa, Katsue; Fujimura, Shun-ichi; Hayashi, Toshiyasu; Takano, Keiji; Oya, Shingo

    1982-12-01

    In the nuclear power plants using PWRs or BWRs in Japan, environmental tests are provided, in which simulated LOCA conditions are considered so as to conform with Japanese conditions, and many cables which passed these tests are presently employed. Lately, the new environmental testing, in which a credible accident called MSLB (main steam line breakage) is taken into account, is investigated in PWR nuclear power plants, besides LOCA. This paper reports on the results of evaluating some PWR cables for this new environmental testing conditions. The several cables tested were selected out of PH cables (fire-retardant, ethylene propylene rubber insulated, chlorosulfonated polyethylene sheathed cables) as the cables for safety protecting circuits and to be used in containment vessels where the cables are to be exposed to severe environmental test conditions of 2 x 10/sup 8/ Rad ..gamma..-irradiation and simulated LOCA. All these cables have been accepted after the vertical tray burning test provided in the IEEE Standard 383. The new testing was carried out by sequentially applying thermal deterioration, ..gamma..-irradiation, and the exposure to steam (twice 300 s exposures to 190 deg C superheated steam). After completing each step, tensile strength, elongation, insulation resistance and breakdown voltage were measured, respectively. Every cable tested showed satisfactory breakdown voltage after the exposure to steam, thus it was decided to be acceptable. In future, it is required to investigate the influence of the rate of temperature rise on the cable to be tested in MSLB simulation.

  5. Test results for cables used in nuclear power plants by a new environmental testing method

    International Nuclear Information System (INIS)

    Handa, Katsue; Fujimura, Shun-ichi; Hayashi, Toshiyasu; Takano, Keiji; Oya, Shingo

    1982-01-01

    In the nuclear power plants using PWRs or BWRs in Japan, environmental tests are provided, in which simulated LOCA conditions are considered so as to conform with Japanese conditions, and many cables which passed these tests are presently employed. Lately, the new environmental testing, in which a credible accident called MSLB (main steam line breakage) is taken into account, is investigated in PWR nuclear power plants, besides LOCA. This paper reports on the results of evaluating some PWR cables for this new environmental testing conditions. The several cables tested were selected out of PH cables (fire-retardant, ethylene propylene rubber insulated, chlorosulfonated polyethylene sheathed cables) as the cables for safety protecting circuits and to be used in containment vessels where the cables are to be exposed to severe environmental test conditions of 2 x 10 8 Rad γ-irradiation and simulated LOCA. All these cables have been accepted after the vertical tray burning test provided in the IEEE Standard 383. The new testing was carried out by sequentially applying thermal deterioration, γ-irradiation, and the exposure to steam (twice 300 s exposures to 190 deg C superheated steam). After completing each step, tensile strength, elongation, insulation resistance and breakdown voltage were measured, respectively. Every cable tested showed satisfactory breakdown voltage after the exposure to steam, thus it was decided to be acceptable. In future, it is required to investigate the influence of the rate of temperature rise on the cable to be tested in MSLB simulation. (Wakatsuki, Y.)

  6. Testing Situation Awareness Network for the Electrical Power Infrastructure

    Directory of Open Access Journals (Sweden)

    Rafał Leszczyna

    2016-09-01

    Full Text Available The contemporary electrical power infrastructure is exposed to new types of threats. The cause of such threats is related to the large number of new vulnerabilities and architectural weaknesses introduced by the extensive use of Information and communication Technologies (ICT in such complex critical systems. The power grid interconnection with the Internet exposes the grid to new types of attacks, such as Advanced Persistent Threats (APT or Distributed-Denial-ofService (DDoS attacks. When addressing this situation the usual cyber security technologies are prerequisite, but not sufficient. To counter evolved and highly sophisticated threats such as the APT or DDoS, state-of-the-art technologies including Security Incident and Event Management (SIEM systems, extended Intrusion Detection/Prevention Systems (IDS/IPS and Trusted Platform Modules (TPM are required. Developing and deploying extensive ICT infrastructure that supports wide situational awareness and allows precise command and control is also necessary. In this paper the results of testing the Situational Awareness Network (SAN designed for the energy sector are presented. The purpose of the tests was to validate the selection of SAN components and check their operational capability in a complex test environment. During the tests’ execution appropriate interaction between the components was verified.

  7. Power Systems Development Facility Gasification Test Run TC09

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2002-09-30

    This report discusses Test Campaign TC09 of the Kellogg Brown & Root, Inc. (KBR) Transport Gasifier train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Gasifier is an advanced circulating fluidized-bed gasifier designed to operate as either a combustor or a gasifier in air- or oxygen-blown mode of operation using a particulate control device (PCD). The Transport Gasifier was operated as a pressurized gasifier during TC09 in air- and oxygen-blown modes. Test Run TC09 was started on September 3, 2002, and completed on September 26, 2002. Both gasifier and PCD operations were stable during the test run, with a stable baseline pressure drop. The oxygen feed supply system worked well and the transition from air to oxygen was smooth. The gasifier temperature varied between 1,725 and 1,825 F at pressures from 125 to 270 psig. The gasifier operates at lower pressure during oxygen-blown mode due to the supply pressure of the oxygen system. In TC09, 414 hours of solid circulation and over 300 hours of coal feed were attained with almost 80 hours of pure oxygen feed.

  8. Power cycling test and failure analysis of molded Intelligent Power IGBT Module under different temperature swing durations

    DEFF Research Database (Denmark)

    Choi, Uimin; Blaabjerg, Frede; Jørgensen, Søren

    2016-01-01

    on the lifetime of 600 V, 30 A, 3-phase molded Intelligent PowerModules (IPM) and their failuremechanismsare investigated. The study is based on the accelerated power cycling test results of 36 samples under 6 different conditions and tests are performed under realistic electrical conditions by an advanced power...

  9. Power Systems Development Facility Gasification Test Run TC07

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2002-04-05

    This report discusses Test Campaign TC07 of the Kellogg Brown & Root, Inc. (KBR) Transport Reactor train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Reactor is an advanced circulating fluidized-bed reactor designed to operate as either a combustor or a gasifier using a particulate control device (PCD). The Transport Reactor was operated as a pressurized gasifier during TC07. Prior to TC07, the Transport Reactor was modified to allow operations as an oxygen-blown gasifier. Test Run TC07 was started on December 11, 2001, and the sand circulation tests (TC07A) were completed on December 14, 2001. The coal-feed tests (TC07B-D) were started on January 17, 2002 and completed on April 5, 2002. Due to operational difficulties with the reactor, the unit was taken offline several times. The reactor temperature was varied between 1,700 and 1,780 F at pressures from 200 to 240 psig. In TC07, 679 hours of solid circulation and 442 hours of coal feed, 398 hours with PRB coal and 44 hours with coal from the Calumet mine, and 33 hours of coke breeze feed were attained. Reactor operations were problematic due to instrumentation problems in the LMZ resulting in much higher than desired operating temperatures in the reactor. Both reactor and PCD operations were stable and the modifications to the lower part of the gasifier performed well while testing the gasifier with PRB coal feed.

  10. Estimation of sample size and testing power (Part 4).

    Science.gov (United States)

    Hu, Liang-ping; Bao, Xiao-lei; Guan, Xue; Zhou, Shi-guo

    2012-01-01

    Sample size estimation is necessary for any experimental or survey research. An appropriate estimation of sample size based on known information and statistical knowledge is of great significance. This article introduces methods of sample size estimation of difference test for data with the design of one factor with two levels, including sample size estimation formulas and realization based on the formulas and the POWER procedure of SAS software for quantitative data and qualitative data with the design of one factor with two levels. In addition, this article presents examples for analysis, which will play a leading role for researchers to implement the repetition principle during the research design phase.

  11. Testing the predictive power of nuclear mass models

    International Nuclear Information System (INIS)

    Mendoza-Temis, J.; Morales, I.; Barea, J.; Frank, A.; Hirsch, J.G.; Vieyra, J.C. Lopez; Van Isacker, P.; Velazquez, V.

    2008-01-01

    A number of tests are introduced which probe the ability of nuclear mass models to extrapolate. Three models are analyzed in detail: the liquid drop model, the liquid drop model plus empirical shell corrections and the Duflo-Zuker mass formula. If predicted nuclei are close to the fitted ones, average errors in predicted and fitted masses are similar. However, the challenge of predicting nuclear masses in a region stabilized by shell effects (e.g., the lead region) is far more difficult. The Duflo-Zuker mass formula emerges as a powerful predictive tool

  12. Resolving power test of 2-D K+ K+ interferometry

    International Nuclear Information System (INIS)

    Padula, Sandra S.; Roldao, Christiane G.

    1999-01-01

    Adopting a procedure previously proposed to quantitatively study pion interferometry 1 , an equivalent 2-D X 2 analysis was performed to test the resolving power of that method when applied to less favorable conditions, when no significant contribution from long lived resonances is expected, as in kaon interferometry. For that purpose, use is made of the preliminary E859 K + K + interferometry data from Si+Au collisions at 14.6 A GeV/c. Less sensitivity is achieved in the present case, although it is shown that it is still possible to distinguish two distinct decoupling geometries. (author)

  13. The power of reachability testing for timed automata

    DEFF Research Database (Denmark)

    Aceto, Luca; Bouyer, Patricia; Burgueno, A.

    2003-01-01

    The computational engine of the verification tool UPPAAL consists of a collection of efficient algorithms for the analysis of reachability properties of systems. Model-checking of properties other than plain reachability ones may currently be carried out in such a tool as follows. Given a propert...... be reached. Finally, the property language characterizing the power of reachability testing is used to provide a definition of characteristic properties with respect to a timed version of the ready simulation preorder, for nodes of tau-free, deterministic timed automata....

  14. High power testing of a 17 GHz photocathode RF gun

    International Nuclear Information System (INIS)

    Chen, S.C.; Danly, B.G.; Gonichon, J.

    1995-01-01

    The physics and technological issues involved in high gradient particle acceleration at high microwave (RF) frequencies are under study at MIT. The 17 GHz photocathode RF gun has a 1 1/2 cell (π mode) room temperature cooper cavity. High power tests have been conducted at 5-10 MW levels with 100 ns pulses. A maximum surface electric field of 250 MV/m was achieved. This corresponds to an average on-axis gradient of 150 MeV/m. The gradient was also verified by a preliminary electron beam energy measurement. Even high gradients are expected in our next cavity design

  15. PIE of test assembly of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Ran, M.; Yan, J.; Wang, S.

    2000-01-01

    The small dimensional test fuel assembly (3x3-2) for the Qinshan Nuclear Power Plant was irradiated up to 25.7 Gwd/tU in the in-pile loop (15.5 Mpa,320 C) in Heavy Water Research Reactor (HWRR), CIAE, at simulative condition to Qinshan PWR normal and short time overpower operation for verifying the design, technology, and material properties of the fuel assembly. Comprehensive post-irradiation examination (PIE) including dimension measurement, gamma scanning, eddy current test, X ray, radiography, measurement of fission gas release, and quantitative metallography etc. were performed. PIE results show that the diameter of the fuel rods changed, ridges appeared on the cladding, pellets swelled, and the rate of fission gas release was higher than what we expected. The results would be an important basis for further improvement of design, technology and material properties for Qinshan PWR assembly. (author)

  16. Ascomycin macrolactam derivative SDZ ASM 981 inhibits the release of granule-associated mediators and of newly synthesized cytokines in RBL 2H3 mast cells in an immunophilin-dependent manner.

    Science.gov (United States)

    Hultsch, T; Müller, K D; Meingassner, J G; Grassberger, M; Schopf, R E; Knop, J

    1998-09-01

    Mast cells play an important role in the pathological development of many inflammatory and allergic diseases and inhibition of mast cell activation is a potential target for therapeutic intervention. Therefore, the effect of the novel ascomycin macrolactam derivative SDZ ASM 981 on Fc epsilonRI-mediated activation of rat basophilic leukemia (RBL) cells, as a model for mast cell activation, was investigated. First, the ability to inhibit different mast cell immunophilins in vitro was tested. Using recombinant macrophilin-12 (FKBP-12), inhibition of rotamase activity with an IC50 of approximately 6 nM was observed. The rotamase activity of cyclophilin A (18 kDa) was not affected. Secondly, the effect of SDZ ASM 981 on Fc epsilonRI-mediated mast cell activation was investigated in the RBL cell model. SDZ ASM 981 inhibited exocytosis of preformed mediators (e.g. serotonin) with an IC50 of approximately 30 nM. Transcription and release of newly synthesized mediators (e.g. TNF-alpha) was inhibited with an IC50 of approximately 100 nM. The inhibitory effect of SDZ ASM 981 was antagonized by rapamycin. We conclude that SDZ ASM 981 is a potent inhibitor of Fc epsilonRI-mediated activation of mast cells in vitro. The mechanism of action involves formation of (calcineurin) inhibitory complexes with macrophilins. We suggest that this inhibitory action on mast cells might contribute to the antiinflammatory effect of SDZ ASM 981 observed in vivo (e.g. in aptopic dermatitis and psoriasis).

  17. Full scale dynamic testing of Paks nuclear power plant structures

    International Nuclear Information System (INIS)

    Da Rin, E.M.

    1995-01-01

    This report refers to the full-scale dynamic structural testing activities that have been performed in December 1994 at the Paks (H) Nuclear Power Plant, within the framework of: the IAEA Coordinated research Programme 'Benchmark Study for the Seismic Analysis and Testing of WWER-type Nuclear Power Plants, and the nuclear research activities of ENEL-WR/YDN, the Italian National Electricity Board in Rome. The specific objective of the conducted investigation was to obtain valid data on the dynamic behaviour of the plant's major constructions, under normal operating conditions, for enabling an assessment of their actual seismic safety to be made. As described in more detail hereafter, the Paks NPP site has been subjected to low level earthquake like ground shaking, through appropriately devised underground explosions, and the dynamic response of the plant's 1 st reactor unit important structures was appropriately measured and digitally recorded. In-situ free field response was measured concurrently and, moreover, site-specific geophysical and seismological data were simultaneously acquired too. The above-said experimental data is to provide basic information on the geophysical and seismological characteristics of the Paks NPP site, together with useful reference information on the true dynamic characteristics of its main structures and give some indications on the actual dynamic soil-structure interaction effects for the case of low level excitation

  18. Qualification tests for shift personnel in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Fechner, J B [Bundesministerium des Innern, Bonn (Germany, F.R.)

    1978-02-01

    The selection of personnel for training as shift supervisors or reactor operators so far used to be made by a plant operator mainly on the basis of such criteria as examinations, diplomas and other documents verifying the educational background, the type of activity exercised, and professional success. In addition, there are the opininons of trainers and supervisors based on personal observation of future shift personnel on training for specific plants at a training center, at the manufacturer's, the operator's or in activities in the construction and commissioning of the respective nuclear power plant. In the course of this phase, which normally takes several years, supervisors asses not only the professional capabilities of a trainee, but also bis psychic and physical performance and aptitude, e.g., with respect to decision making, leadership qualifications or behavior unter stress. The advisability of introducing psychological aptitude tests was also studied. However, a decision was recently taken to defer such psychological tests for the time being. Yet, nuclear power plant operators are required to submit a statement to their responsible authorities about industrial medical checkups and qualification assessments by supervisors.

  19. Testing for new physics: neutrinos and the primordial power spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Canac, Nicolas; Abazajian, Kevork N. [Department of Physics, University of California at Irvine, Irvine, CA 92697 (United States); Aslanyan, Grigor [Berkeley Center for Cosmological Physics, University of California, Berkeley, CA 94720 (United States); Easther, Richard [Department of Physics, University of Auckland, Private Bag 92019, Auckland (New Zealand); Price, Layne C., E-mail: ncanac@uci.edu, E-mail: aslanyan@berkeley.edu, E-mail: kevork@uci.edu, E-mail: r.easther@auckland.ac.nz, E-mail: laynep@andrew.cmu.edu [McWilliams Center for Cosmology, Department of Physics, Carnegie Mellon University, Pittsburgh, PA 15213 (United States)

    2016-09-01

    We test the sensitivity of neutrino parameter constraints from combinations of CMB and LSS data sets to the assumed form of the primordial power spectrum (PPS) using Bayesian model selection. Significantly, none of the tested combinations, including recent high-precision local measurements of H{sub 0} and cluster abundances, indicate a signal for massive neutrinos or extra relativistic degrees of freedom. For PPS models with a large, but fixed number of degrees of freedom, neutrino parameter constraints do not change significantly if the location of any features in the PPS are allowed to vary, although neutrino constraints are more sensitive to PPS features if they are known a priori to exist at fixed intervals in log k . Although there is no support for a non-standard neutrino sector from constraints on both neutrino mass and relativistic energy density, we see surprisingly strong evidence for features in the PPS when it is constrained with data from Planck 2015, SZ cluster counts, and recent high-precision local measurements of H{sub 0}. Conversely combining Planck with matter power spectrum and BAO measurements yields a much weaker constraint. Given that this result is sensitive to the choice of data this tension between SZ cluster counts, Planck and H{sub 0} measurements is likely an indication of unmodeled systematic bias that mimics PPS features, rather than new physics in the PPS or neutrino sector.

  20. Qualification tests for shift personnel in nuclear power plants

    International Nuclear Information System (INIS)

    Fechner, J.B.

    1978-01-01

    The selection of personnel for training as shift supervisors or reactor operators so far used to be made by a plant operator mainly on the basis of such criteria as examinations, diplomas and other documents verifying the educational background, the type of activity exercised, and professional success. In addition, there are the opininons of trainers and supervisors based on personal observation of future shift personnel on training for specific plants at a training center, at the manufacturer's, the operator's or in activities in the construction and commissioning of the respective nuclear power plant. In the course of this phase, which normally takes several years, supervisors asses not only the professional capabilities of a trainee, but also bis psychic and physical performance and aptitude, e.g., with respect to decision making, leadership qualifications or behavior unter stress. The advisability of introducing psychological aptitude tests was also studied. However, a decision was recently taken to defer such psychological tests for the time being. Yet, nuclear power plant operators are required to submit a statement to their responsible authorities about industrial medical checkups and qualification assessments by supervisors. (orig.) [de

  1. Comparisons of ASME-code fatigue-evaluation methods for nuclear Class 1 piping with Class 2 or 3 piping

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1983-06-01

    The fatigue evaluation procedure used in the ASME Boiler and Pressure Vessel Code, Sect. III, Nuclear Power Plant Components, for Class 1 piping is different from the procedure used for Class 2 or 3 piping. The basis for each procedure is described, and correlations between the two procedures are presented. Conditions under which either procedure or both may be unconservative are noted. Potential changes in the Class 2 or 3 piping procedure to explicitly cover all loadings are discussed. However, the report is intended to be informative, and while the contents of the report may guide future Code changes, specific recommendations are not given herein

  2. Stress test of the nuclear power plants performed in Taiwan

    Energy Technology Data Exchange (ETDEWEB)

    Wu, C.H.; Teng, W.C.; Chang, S.; Chen, Y.B. [Atomic Energy Council, Taipei, Taiwan (China)

    2014-07-01

    In the wake of Japan's Fukushima Daiichi Nuclear Power Plants event, the Atomic Energy Council (AEC) has asked Taiwan's Nuclear Power Plant operator (TPC) to re-examine and re-evaluate the vulnerabilities of its nuclear units, and furthermore, take possible countermeasures against extreme natural disasters, including earthquake, tsunami and rock-and-mud slide. The evaluation process should be based on both within and beyond Design Basis Accidents, by reference to the actions recommended by the world nuclear authorities and groups, namely, IAEA, USNRC, NEI, ENSREG and WANO. Taiwan is a very densely populated region of the world. Furthermore, like Japan, due to its geophysical position, Taiwan is prone to large scale earthquakes, and although historically rare, Taiwan also faces the potential risk of tsunamis. AEC also asked TPC to perform the stress test following the specification given by WENRA (later ENSREG) and conducted in all the EU's nuclear reactors. After completion of the stress test for all the nuclear power plants, AEC was trying to have the reports peer reviewed by international organizations, as EU did. The OECD/NEA accepted AEC's request and formed a review team specific to the review of Taiwan's National Report for the Stress Test. There were 18 follow-up items after the NEA's review. Based on these items, AEC developed five orders to require TPC further enhancing their capabilities to cope with extreme natural hazards. The ENSREG also formed a nine-expert review team for Taiwan's Stress Test in response to AEC's request almost at the same time as the OECD/NEA. The ENSREG review team began their works in June 2013 by desktop review, and ended in early October 2013 by country visit to Taiwan. While the assessment of post-Fukushima evaluation reveals neither immediate nuclear safety concerns nor threats to the public health and safety, AEC requested that TPC focus on strengthening its re-evaluation on design

  3. Stress test of the nuclear power plants performed in Taiwan

    International Nuclear Information System (INIS)

    Wu, C.H.; Teng, W.C.; Chang, S.; Chen, Y.B.

    2014-01-01

    In the wake of Japan's Fukushima Daiichi Nuclear Power Plants event, the Atomic Energy Council (AEC) has asked Taiwan's Nuclear Power Plant operator (TPC) to re-examine and re-evaluate the vulnerabilities of its nuclear units, and furthermore, take possible countermeasures against extreme natural disasters, including earthquake, tsunami and rock-and-mud slide. The evaluation process should be based on both within and beyond Design Basis Accidents, by reference to the actions recommended by the world nuclear authorities and groups, namely, IAEA, USNRC, NEI, ENSREG and WANO. Taiwan is a very densely populated region of the world. Furthermore, like Japan, due to its geophysical position, Taiwan is prone to large scale earthquakes, and although historically rare, Taiwan also faces the potential risk of tsunamis. AEC also asked TPC to perform the stress test following the specification given by WENRA (later ENSREG) and conducted in all the EU's nuclear reactors. After completion of the stress test for all the nuclear power plants, AEC was trying to have the reports peer reviewed by international organizations, as EU did. The OECD/NEA accepted AEC's request and formed a review team specific to the review of Taiwan's National Report for the Stress Test. There were 18 follow-up items after the NEA's review. Based on these items, AEC developed five orders to require TPC further enhancing their capabilities to cope with extreme natural hazards. The ENSREG also formed a nine-expert review team for Taiwan's Stress Test in response to AEC's request almost at the same time as the OECD/NEA. The ENSREG review team began their works in June 2013 by desktop review, and ended in early October 2013 by country visit to Taiwan. While the assessment of post-Fukushima evaluation reveals neither immediate nuclear safety concerns nor threats to the public health and safety, AEC requested that TPC focus on strengthening its re-evaluation on design

  4. High temperature, high pressure gas loop - the Component Flow Test Loop (CFTL)

    International Nuclear Information System (INIS)

    Gat, U.; Sanders, J.P.; Young, H.C.

    1984-01-01

    The high-pressure, high-temperature, gas-circulating Component Flow Test Loop located at Oak Ridge National Laboratory was designed and constructed utilizing Section III of the ASME Boiler and Pressure Vessel Code. The quality assurance program for operating and testing is also based on applicable ASME standards. Power to a total of 5 MW is available to the test section, and an air-cooled heat exchanger rated at 4.4 MW serves as heat sink. The three gas-bearing, completely enclosed gas circulators provide a maximum flow of 0.47 m 3 /s at pressures to 10.7 MPa. The control system allows for fast transients in pressure, power, temperature, and flow; it also supports prolonged unattended steady-state operation. The data acquisition system can access and process 10,000 data points per second. High-temperature gas-cooled reactor components are being tested

  5. Time-dependent fracture of materials at elevated temperature for solar thermal power systems

    International Nuclear Information System (INIS)

    Gupta, G.D.

    1979-01-01

    Various Solar Thermal Power Systems are briefly described. The components of solar power systems in which time-dependent fracture problems become important are identified. Typical materials of interest, temperature ranges, and stress states are developed; and the number of cycles during the design life of these systems are indicated. The ASME Code procedures used by designers to predict the life of these components are briefly described. Some of the major problems associated with the use of these ASME procedures in the design of solar components are indicated. Finally, a number of test and development needs are identified which would enable the designers to predict the life of the solar power system components with a reasonable degree of confidence

  6. Power Systems Development Facility Gasification Test Run TC08

    Energy Technology Data Exchange (ETDEWEB)

    Southern Company Services

    2002-06-30

    This report discusses Test Campaign TC08 of the Kellogg Brown & Root, Inc. (KBR) Transport Gasifier train with a Siemens Westinghouse Power Corporation (Siemens Westinghouse) particle filter system at the Power Systems Development Facility (PSDF) located in Wilsonville, Alabama. The Transport Gasifier is an advanced circulating fluidized-bed gasifier designed to operate as either a combustor or a gasifier in air- or oxygen-blown mode using a particulate control device (PCD). The Transport Gasifier was operated as a pressurized gasifier in air- and oxygen-blown modes during TC08. Test Run TC08 was started on June 9, 2002 and completed on June 29. Both gasifier and PCD operations were stable during the test run with a stable baseline pressure drop. The oxygen feed supply system worked well and the transition from air to oxygen blown was smooth. The gasifier temperature was varied between 1,710 and 1,770 F at pressures from 125 to 240 psig. The gasifier operates at lower pressure during oxygen-blown mode due to the supply pressure of the oxygen system. In TC08, 476 hours of solid circulation and 364 hours of coal feed were attained with 153 hours of pure oxygen feed. The gasifier and PCD operations were stable in both enriched air and 100 percent oxygen blown modes. The oxygen concentration was slowly increased during the first transition to full oxygen-blown operations. Subsequent transitions from air to oxygen blown could be completed in less than 15 minutes. Oxygen-blown operations produced the highest synthesis gas heating value to date, with a projected synthesis gas heating value averaging 175 Btu/scf. Carbon conversions averaged 93 percent, slightly lower than carbon conversions achieved during air-blown gasification.

  7. Test results of an organic Rankine-cycle power module for a small community solar thermal power experiment

    Science.gov (United States)

    Clark, T. B.

    1985-01-01

    The organic Rankine-cycle (ORC) power conversion assembly was tested. Qualification testing of the electrical transport subsystem was also completed. Test objectives were to verify compatibility of all system elements with emphasis on control of the power conversion assembly, to evaluate the performance and efficiency of the components, and to validate operating procedures. After 34 hours of power generation under a wide range of conditions, the net module efficiency exceeded 18% after accounting for all parasitic losses.

  8. ASME Code requirements for multi-canister overpack design and fabrication

    International Nuclear Information System (INIS)

    SMITH, K.E.

    1998-01-01

    The baseline requirements for the design and fabrication of the MCO include the application of the technical requirements of the ASME Code, Section III, Subsection NB for containment and Section III, Subsection NG for criticality control. ASME Code administrative requirements, which have not historically been applied at the Hanford site and which have not been required by the US Nuclear Regulatory Commission (NRC) for licensed spent fuel casks/canisters, were not invoked for the MCO. As a result of recommendations made from an ASME Code consultant in response to DNFSB staff concerns regarding ASME Code application, the SNF Project will be making the following modifications: issue an ASME Code Design Specification and Design Report, certified by a Registered Professional Engineer; Require the MCO fabricator to hold ASME Section III or Section VIII, Division 2 accreditation; and Use ASME Authorized Inspectors for MCO fabrication. Incorporation of these modifications will ensure that the MCO is designed and fabricated in accordance with the ASME Code. Code Stamping has not been a requirement at the Hanford site, nor for NRC licensed spent fuel casks/canisters, but will be considered if determined to be economically justified

  9. Flanged joints with contact outside the bolt circle: ASME Part B design rules

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.; Moore, S.E.

    1976-05-01

    The ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, gives rules which are subdivided into ''Part A'' and ''Part B''. Part A covers flanged joints where contact between flanges occurs through a gasket located inside the bolt holes. Part B covers flanged joints with contact outside the bolt holes. This report (a) summarizes the theory for Part B flanged joints, (b) presents examples which show the significant differences between Part A flanged joints and Part B flanged joints, (c) presents the available test data relevant to the characteristics of Part B flanged joints, (d) gives listings of two computer programs which can be used to evaluate the characteristics of Part B flanged joints, and (e) gives recommendations for Code revisions and other aspects of Part B flanged-joint design

  10. Performance testing of self-powered detector signal converters at Dukovany nuclear power plant - stage 1

    International Nuclear Information System (INIS)

    Erben, O.; Hajek, P.; Zerola, L.; Karsulin, M.

    1990-11-01

    The converters were manufactured at the Institute of Nuclear Research, Rez. Dynamic functions of the converters were tested during the start-up of reactor unit 4, Dukovany nuclear power plant, and their stability during its normal operation. The results and evaluation of the measurements show a good performance of converters. They have a low offset, good stability and the values of current are in a good agreement with the values obtained using other methods. The values of insulation resistance are in a good agreement with the values obtained manually using the method of additional resistance. These converters are planned to be used in the upgraded in-service inspection system in WWER-440 nuclear power plants. (Z.S.) 9 tabs., 22 figs., 1 ref

  11. Partial discharge testing of in-situ power cable accessories

    Energy Technology Data Exchange (ETDEWEB)

    Orban, H. E.

    2002-07-01

    An overview of commercially available diagnostic methods for in-situ power cable accessories is given and relevant field experiences with these diagnostics are described. The discussion includes both PILC and polymeric insulated cables. Two major types of degradation are most frequently involved in cable systems. One is an overall condition caused by chemical aging and /or water treeing. Diagnostics for this type of aging include dissipation factor (loss angle), harmonic analysis, return voltage, isothermal relaxation current, dielectric response, or dc leakage current. The second type of degradation is discrete or incremental; condition assessment utilizes dissipation factor measurements or partial discharge (PD) level measurements. The focus in this paper is on PD diagnostics, especially off-line methods such as the 60 Hz test, the combined AC and VLF diagnostic, and the oscillating wave test system test. Among on-line diagnostics, ultrasonic detection of partial discharge and measurement of partial discharge by installing direct, capacitive or inductive couplers near cable accessories, are described. Overall, partial discharge detection and location in cable accessories is considered inadequate, since interpretation of results is difficult due to the number of variables involved. 28 refs., 1 tab.

  12. Probing the Mysteries of the X-Ray Binary 4U 1210-64 with ASM, MAXI and Suzaku

    Science.gov (United States)

    Coley, Joel B.; Corbet, R.; Mukai, K.; Pottschmidt, K.

    2013-01-01

    Optical and X-ray observations of 4U 1210-64 (1ES 1210-646) suggest that the source is a High Mass X-ray Binary (HMXB) probably powered by the Be mechanism. Data acquired by the RXTE All Sky Monitor (ASM), the ISS Monitor of All-sky X-ray Image (MAXI) and Suzaku provide a detailed temporal and spectral description of this poorly understood source. Long-term data produced by ASM and MAXI indicate that the source shows two distinct high and low states. A 6.7-day orbital period of the system was found in folded light curves produced by both ASM and MAXI. A two day Suzaku observation in Dec. 2010 took place during a transition from the minimum to the maximum of the folded light curve. The two day Suzaku observation reveals large variations in flux indicative of strong orbit to orbit variability. Flares in the Suzaku light curve can reach nearly 1.4 times the mean count rate. From a spectral analysis of the Suzaku data, emission lines in the Fe K alpha region were detected at 6.4 keV, 6.7 keV and 6.97 keV interpreted as FeI, FeXXV and FeXXVI. In addition, emission lines were observed at approximately 1.0 and 2.6 keV, corresponding to NeX and SXVI respectively. Thermal bremsstrahlung or power law models both modified by interstellar and partially covering absorption provide a good fit to the continuum data. This source is intriguing for these reasons: i) No pulse period was observed; ii) 6.7 day orbital period is much less than typical orbital periods seen in Be/X-ray Binaries; iii) The optical companion is a B5V--an unusual spectral class for an HMXB; iv) There are extended high and low X-ray states.

  13. Integrin and GPCR Crosstalk in the Regulation of ASM Contraction Signaling in Asthma.

    Science.gov (United States)

    Teoh, Chun Ming; Tam, John Kit Chung; Tran, Thai

    2012-01-01

    Airway hyperresponsiveness (AHR) is one of the cardinal features of asthma. Contraction of airway smooth muscle (ASM) cells that line the airway wall is thought to influence aspects of AHR, resulting in excessive narrowing or occlusion of the airway. ASM contraction is primarily controlled by agonists that bind G protein-coupled receptor (GPCR), which are expressed on ASM. Integrins also play a role in regulating ASM contraction signaling. As therapies for asthma are based on symptom relief, better understanding of the crosstalk between GPCRs and integrins holds good promise for the design of more effective therapies that target the underlying cellular and molecular mechanism that governs AHR. In this paper, we will review current knowledge about integrins and GPCRs in their regulation of ASM contraction signaling and discuss the emerging concept of crosstalk between the two and the implication of this crosstalk on the development of agents that target AHR.

  14. 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

  15. Testing of mobile surveillance robot at a nuclear power plant

    International Nuclear Information System (INIS)

    White, J.R.; Harvey, H.W.; Farnstrom, K.A.

    1987-01-01

    In-plant testing of a mobile surveillance robot (SURBOT) was performed at the Browns Ferry Nuclear Plant by TVA personnel. The results verified that SURBOT can be used for remote surveillance in 54 separate controlled radiation rooms at the plant. High-quality color video, audio, and other data are collected, digitized by an on-board computer, and transmitted through a cable to the control console for real-time display and videotaping. TVA projects that the use of SURBOT for surveillance during plant operation will produce annual savings of about 100 person-rem radiation exposure and $200,000 in operating costs. Based on the successful results of this program, REMOTEC is now commercializing the SURBOT technology on both wheeled and tracked mobile robots for use in nuclear power plants and other hazardous environments

  16. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    International Nuclear Information System (INIS)

    Horschel, D.S.

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission's program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix

  17. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    Energy Technology Data Exchange (ETDEWEB)

    Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission`s program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix.

  18. Designing and Testing Energy Harvesters Suitable for Renewable Power Sources

    Science.gov (United States)

    Synkiewicz, B.; Guzdek, P.; Piekarski, J.; Zaraska, K.

    2016-01-01

    Energy harvesters convert waste power (heat, light and vibration) directly to electric power . Fast progress in their technology, design and areas of application (e.g. “Internet of Things”) has been observed recently. Their effectiveness is steadily growing which makes their application to powering sensor networks with wireless data transfer reasonable. The main advantage is the independence from wired power sources, which is especially important for monitoring state of environmental parameters. In this paper we describe the design and realization of a gas sensor monitoring CO level (powered by TEG) and two, designed an constructed in ITE, autonomous power supply modules powered by modern photovoltaic cells.

  19. Designing and Testing Energy Harvesters Suitable for Renewable Power Sources

    International Nuclear Information System (INIS)

    Synkiewicz, B.; Guzdek, P.; Piekarski, J.; Zaraska, K.

    2016-01-01

    Energy harvesters convert waste power (heat, light and vibration) directly to electric power . Fast progress in their technology, design and areas of application (e.g. “Internet of Things”) has been observed recently. Their effectiveness is steadily growing which makes their application to powering sensor networks with wireless data transfer reasonable. The main advantage is the independence from wired power sources, which is especially important for monitoring state of environmental parameters. In this paper we describe the design and realization of a gas sensor monitoring CO level (powered by TEG) and two, designed an constructed in ITE, autonomous power supply modules powered by modern photovoltaic cells

  20. Summarisation of construction and commissioning experience for nuclear power integrated test facility

    International Nuclear Information System (INIS)

    Xiao Zejun; Jia Dounan; Jiang Xulun; Chen Bingde

    2003-01-01

    Since the foundation of Nuclear Power Institute of China, it has successively designed various engineering experimental facilities, and constructed nuclear power experimental research base, and accumulated rich construction experiences of nuclear power integrated test facility. The author presents experience on design, construction and commissioning of nuclear power integrated test facility

  1. Comportamiento de variables clínico-epidemiológicas en pacientes asmáticos. Consultorio Médico de Familia 14, Puerto Padre

    Directory of Open Access Journals (Sweden)

    José Ramón Martínez Pérez

    2015-12-01

    Full Text Available Se realizó un estudio descriptivo de corte transversal en 40 pacientes asmáticos de todas las edades, pertenecientes al Consultorio 14 del área de salud de Puerto Padre, provincia de Las Tunas, durante los meses de enero a junio de 2008; con el objetivo de determinar el comportamiento de variables clínico-epidemiológicas en dichos pacientes. Para cumplimentar los objetivos se diseñó una encuesta que recogió las variables de interés. El análisis se hizo a través del sistema Statistix, permitiendo  obtener  la distribución de frecuencia y los porcientos. El test de Chi-Cuadrado se empleó en la comparación de variables cualitativas. Los resultados fueron descritos, interpretados y comparados con otros estudios. Se concluye que los pacientes grado I según Kraeppelin son los mas representados en nuestra población de pacientes asmáticos;  los broncodilatadores son los medicamentos mas utilizados por la población asmática estudiada; el incumplimiento del tratamiento medicamentoso no se relacionó con la severidad de la enfermedad en la población estudiada;  las malas condiciones de la vivienda y  el tabaquismo influyeron de forma negativa en la evolución del paciente asmático.

  2. Nuclear component design ontology building based on ASME codes

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan

    2005-01-01

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

  3. Elastic creep-fatigue evaluation for ASME code

    International Nuclear Information System (INIS)

    Severud, L.K.; Winkel, B.V.

    1987-01-01

    Experience with applying the ASME Code Case N-47 rules for evaluation of creep-fatigue with elastic analysis results has been problematic. The new elastic evaluation methods are intended to bound the stress level and strain range values needed for use in employing the code inelastic analysis creep-fatigue damage counting procedures. To account for elastic followup effects, ad hoc rules for stress classification, shakedown, and ratcheting are employed. Because elastic followup, inelastic strain concentration, and stress-time effects are accounted for, the design fatigue curves in Case N-47 for inelastic analysis are used instead of the more conservative elastic analysis curves. Creep damage assessments are made using an envelope stress-time history that treats multiple load events and repeated cycles during elevated temperature service life. (orig./GL)

  4. Materials, manufacture and testing of pressurized components of high-power steam power plants

    International Nuclear Information System (INIS)

    Blind, D.; Foehl, J.; Issler, L.; Schellhammer, W.; Sturm, D.; Kussmaul, K.; Heinrich, D.; Meyer, H.J.; Prestel, W.

    1981-01-01

    This is the first German review of materials, production and testing of pressure components of high-capacity steam power plants. The authors have been working in this field for years; their special subject has been the availability and reliability of pressure vessels, in particular in nuclear engineering. Fundamentals are presented as well as the findings obtained at the state Materials Testing Institute in Stuttgart. The material is presented in a well-structured classification; the most recent international findings, especially of the USA, are presented. This is possible due to the close cooperation between the Stuttgart institute and a number of US research institutes. The new subject of fracture mechanics is treated in some detail; its fundamentals are discussed from the American point of view while German considerations - in particular of the Reactor Safety Commission - are taken into account in the field of applications. (orig.) [de

  5. Generic 12-Bus Test System for Wind Power Integration Studies

    DEFF Research Database (Denmark)

    Adamczyk, Andrzej Grzegorz; Altin, Müfit; Göksu, Ömer

    2012-01-01

    , inertial response, frequency control, damping of electromechanical oscillations, balanced and unbalanced fault management, etc. Hence, the power system components: conventional power plants with controls, transmission lines, transformers and loads should be represented accurately to achieve realistic power......High wind power penetration levels into power systems requires an appropriate power system model when assessing impact on the overall system stability. The model should capture the wide range of dynamics related to the wind integration studies, such as voltage control, synchronizing power control...... system characteristics. Additionally, the power system model should be simple and computationally manageable in order to simulate multiple scenarios with different control parameters in a reasonable time. In this paper, a generic power system model is presented in order to comprehend the wind integration...

  6. Fuel staging tests at the Kymijaervi power plant

    International Nuclear Information System (INIS)

    Kivelae, M.; Rotter, H.; Virkki, J.

    1990-01-01

    The aim of this study was to measure nitrogen oxide (NO x ) emissions and find the methods to reduce them in plants using coal and natural gas as fuel. The tests involved were made at the Kymijaervi Power Plant, Lahti, Finland. Coal and natural gas was used alone or mixed. With natural gas when using flue gas recirculation, the NO x emission level dropped from 330 mg/m 3 down to 60 mg/m 3 . A negative side effect was that the flue gas temperature increased. At coal combustion and staged combustion, the flue gas recirculation had no significant effect on the NO x emission level. At coal combustion, the staging of combustion air halved the NO x emission but the combustibles increased strongly. With fuel staging, using coal as main fuel and gas as staging fuel, the NO x emission level was decreased from 340 mg/m 3 to 170 mg/m 3 . At the same time the combustibles increased 2 %- units. Also the flue gas temperature increased a little. At the tests, the proportion of natural gas was rather high, one third of the fuel energy input, but it could not be decreased, because the gas flow ratio was already too low to ensure good mixing

  7. Cell-Phone Tower Power System Prototype Testing for Verizon Wireless |

    Science.gov (United States)

    Advanced Manufacturing Research | NREL Cell-Phone Tower Power System Prototype Testing for Verizon Wireless Cell-Phone Tower Power System Prototype Testing for Verizon Wireless For Verizon Wireless , NREL tested a new cell-phone tower power system prototype based on DC interconnection and photovoltaics

  8. Theoretical parameter histories of dynamic tests during power commissioning of Mochovce units power level up to 100 % Nnom

    International Nuclear Information System (INIS)

    Jagrik, J.; Mraz, M.; Rapant, M.; Stefanovic, P.; Kotasek, J.; Gieci, A.; Macko, J.; Mosny, J.

    1998-01-01

    Theoretical histories of selected parameters for dynamic tests carried out in the course of power commissioning of the Mochovce Unit 1 at the power level 100% N nom are shown in the report. The expected histories given were developed based on calculations performed by means of simulator in Nuclear Power Plants Research Institute Trnava, Inc., simulator in EGU Praha and simulator at the Mochovce plant, as well as based on similar tests at both Bohunice and Dukovany plants

  9. Testing the Predictive Power of Coulomb Stress on Aftershock Sequences

    Science.gov (United States)

    Woessner, J.; Lombardi, A.; Werner, M. J.; Marzocchi, W.

    2009-12-01

    Empirical and statistical models of clustered seismicity are usually strongly stochastic and perceived to be uninformative in their forecasts, since only marginal distributions are used, such as the Omori-Utsu and Gutenberg-Richter laws. In contrast, so-called physics-based aftershock models, based on seismic rate changes calculated from Coulomb stress changes and rate-and-state friction, make more specific predictions: anisotropic stress shadows and multiplicative rate changes. We test the predictive power of models based on Coulomb stress changes against statistical models, including the popular Short Term Earthquake Probabilities and Epidemic-Type Aftershock Sequences models: We score and compare retrospective forecasts on the aftershock sequences of the 1992 Landers, USA, the 1997 Colfiorito, Italy, and the 2008 Selfoss, Iceland, earthquakes. To quantify predictability, we use likelihood-based metrics that test the consistency of the forecasts with the data, including modified and existing tests used in prospective forecast experiments within the Collaboratory for the Study of Earthquake Predictability (CSEP). Our results indicate that a statistical model performs best. Moreover, two Coulomb model classes seem unable to compete: Models based on deterministic Coulomb stress changes calculated from a given fault-slip model, and those based on fixed receiver faults. One model of Coulomb stress changes does perform well and sometimes outperforms the statistical models, but its predictive information is diluted, because of uncertainties included in the fault-slip model. Our results suggest that models based on Coulomb stress changes need to incorporate stochastic features that represent model and data uncertainty.

  10. Regulation of dynein-mediated autophagosomes trafficking by ASM in CASMCs.

    Science.gov (United States)

    Xu, Ming; Zhang, Qiufang; Li, Pin-Lan; Nguyen, Thaison; Li, Xiang; Zhang, Yang

    2016-01-01

    Acid sphingomyelinase (ASM; gene symbol Smpd1) has been shown to play a crucial role in autophagy maturation by controlling lysosomal fusion with autophagosomes in coronary arterial smooth muscle cells (CASMCs). However, the underlying molecular mechanism by which ASM controls autophagolysosomal fusion remains unknown. In primary cultured CASMCs, lysosomal Ca2+ induced by 7-ketocholesterol (7-Ket, an atherogenic stimulus and autophagy inducer) was markedly attenuated by ASM deficiency or TRPML1 gene silencing suggesting that ASM signaling is required for TRPML1 channel activity and subsequent lysosomal Ca(2+) release. In these CASMCs, ASM deficiency or TRPML1 gene silencing markedly inhibited 7-Ket-induced dynein activation. In addition, 7-Ket-induced autophagosome trafficking, an event associated with lysosomal Ca(2+) release and dynein activity, was significantly inhibited in ASM-deficient (Smpd1(-/-)) CASMCs compared to that in Smpd1(+/+) CASMCs. Finally, overexpression of TRPML1 proteins restored 7-Ket-induced lysosomal Ca(2+) release and autophagosome trafficking in Smpd1-/- CASMCs. Collectively, these results suggest that ASM plays a critical role in regulating lysosomal TRPML1-Ca(2+) signaling and subsequent dynein-mediated autophagosome trafficking, which leads its role in controlling autophagy maturation in CASMCs under atherogenic stimulation.

  11. Modifications to LLNL Plutonium Packaging Systems (PuPS) to achieve ASME VIII UW-13.2(d) Requirements for the DOE Standard 3013-00 Outer Can Weld

    International Nuclear Information System (INIS)

    Riley, D; Dodson, K

    2001-01-01

    The Lawrence Livermore National Laboratory (LLNL) Plutonium Packaging System (PuPS) prepares packages to meet the DOE Standard 3013 (Reference 1). The PuPS equipment was supplied by the British Nuclear Fuels Limited (BNFL). The DOE Standard 3013 requires that the welding of the Outer Can meets ASME Section VIII Division 1 (Reference 2). ASME Section VIII references to ASME Section IX (Reference 3) for most of the welding requirements, but UW-13.2 (d) of Section VIII requires a certain depth and width of the weld. In this document the UW-13.2(d) requirement is described as the (a+b)/2t s ratio. This ratio has to be greater than or equal to one to meet the requirements of UW-13.2(d). The Outer Can welds had not been meeting this requirement. Three methods are being followed to resolve this issue: (1) Modify the welding parameters to achieve the requirement, (2) Submit a weld case to ASME that changes the UW-13.2(d) requirement for their review and approval, and (3) Change the requirements in the DOE-STD-3013. Each of these methods are being pursued. This report addresses how the first method was addressed for the LLNL PuPS. The experimental work involved adjusting the Outer Can rotational speed and the power applied to the can. These adjustments resulted in being able to achieve the ASME VIII, UW-13.2(d) requirement

  12. Statistical tests for power-law cross-correlated processes

    Science.gov (United States)

    Podobnik, Boris; Jiang, Zhi-Qiang; Zhou, Wei-Xing; Stanley, H. Eugene

    2011-12-01

    For stationary time series, the cross-covariance and the cross-correlation as functions of time lag n serve to quantify the similarity of two time series. The latter measure is also used to assess whether the cross-correlations are statistically significant. For nonstationary time series, the analogous measures are detrended cross-correlations analysis (DCCA) and the recently proposed detrended cross-correlation coefficient, ρDCCA(T,n), where T is the total length of the time series and n the window size. For ρDCCA(T,n), we numerically calculated the Cauchy inequality -1≤ρDCCA(T,n)≤1. Here we derive -1≤ρDCCA(T,n)≤1 for a standard variance-covariance approach and for a detrending approach. For overlapping windows, we find the range of ρDCCA within which the cross-correlations become statistically significant. For overlapping windows we numerically determine—and for nonoverlapping windows we derive—that the standard deviation of ρDCCA(T,n) tends with increasing T to 1/T. Using ρDCCA(T,n) we show that the Chinese financial market's tendency to follow the U.S. market is extremely weak. We also propose an additional statistical test that can be used to quantify the existence of cross-correlations between two power-law correlated time series.

  13. [Ca2+]i oscillations in ASM: relationship with persistent airflow obstruction in asthma.

    Science.gov (United States)

    Sweeney, David; Hollins, Fay; Gomez, Edith; Saunders, Ruth; Challiss, R A John; Brightling, Christopher E

    2014-07-01

    The cause of airway smooth muscle (ASM) hypercontractility in asthma is not fully understood. The relationship of spontaneous intracellular calcium oscillation frequency in ASM to asthma severity was investigated. Oscillations were increased in subjects with impaired lung function abolished by extracellular calcium removal, attenuated by caffeine and unaffected by verapamil or nitrendipine. Whether modulation of increased spontaneous intracellular calcium oscillations in ASM from patients with impaired lung function represents a therapeutic target warrants further investigation. © 2014 The Authors. Respirology published by Wiley Publishing Asia Pty Ltd on behalf of Asian Pacific Society of Respirology.

  14. Highlights of proposed changes to ANSI/ASME N509-80

    International Nuclear Information System (INIS)

    Ornberg, S.C.

    1987-01-01

    The ASME Committee on Nuclear Air and Gas Treatment (CONAGT) are at the time of this writing considering performing maintenance revisions of ANSI N509 and N510 based on the results of a required 5-year review and comments received from users of the standards at workshops and through inquiries. This paper discusses the highlights of the significant revisions to ANSI/ASME N509 and explains the reasons for the changes. It should be emphasized that these revisions are not yet approved by ASME CONAGT, the board of Nuclear Codes and Standards, or ANSI

  15. On the Existence of Uniformly Most Powerful Bayesian Tests With Application to Non-Central Chi-Squared Tests

    OpenAIRE

    Nikooienejad, Amir; Johnson, Valen E.

    2018-01-01

    Uniformly most powerful Bayesian tests (UMPBTs) are an objective class of Bayesian hypothesis tests that can be considered the Bayesian counterpart of classical uniformly most powerful tests. Unfortunately, UMPBTs have only been exposed for application in one parameter exponential family models. The purpose of this article is to describe methodology for deriving UMPBTs for a larger class of tests. Specifically, we introduce sufficient conditions for the existence of UMPBTs and propose a unifi...

  16. Power plant cost estimates put to the test

    International Nuclear Information System (INIS)

    Crowley, J.H.

    1978-01-01

    The growth in standards for nuclear applications and the impact of these codes and standards on the cost of nuclear power plants is described. The preparation of cost estimates and reasons for apparent discrepancies are discussed. Consistent estimates of nuclear power plant costs have been prepared in the USA for over a decade. They show that the difference in capital costs between nuclear and coal fired plants is narrowing and that when total generating costs are calculated nuclear power is substantially cheaper. (UK)

  17. The ASM Curriculum Guidelines for Undergraduate Microbiology: A Case Study of the Advocacy Role of Societies in Reform Efforts.

    Science.gov (United States)

    Horak, Rachel E A; Merkel, Susan; Chang, Amy

    2015-05-01

    A number of national reports, including Vision and Change in Undergraduate Biology Education: A Call to Action, have called for drastic changes in how undergraduate biology is taught. To that end, the American Society for Microbiology (ASM) has developed new Curriculum Guidelines for undergraduate microbiology that outline a comprehensive curriculum for any undergraduate introductory microbiology course or program of study. Designed to foster enduring understanding of core microbiology concepts, the Guidelines work synergistically with backwards course design to focus teaching on student-centered goals and priorities. In order to qualitatively assess how the ASM Curriculum Guidelines are used by educators and learn more about the needs of microbiology educators, the ASM Education Board distributed two surveys to the ASM education community. In this report, we discuss the results of these surveys (353 responses). We found that the ASM Curriculum Guidelines are being implemented in many different types of courses at all undergraduate levels. Educators indicated that the ASM Curriculum Guidelines were very helpful when planning courses and assessments. We discuss some specific ways in which the ASM Curriculum Guidelines have been used in undergraduate classrooms. The survey identified some barriers that microbiology educators faced when trying to adopt the ASM Curriculum Guidelines, including lack of time, lack of financial resources, and lack of supporting resources. Given the self-reported challenges to implementing the ASM Curriculum Guidelines in undergraduate classrooms, we identify here some activities related to the ASM Curriculum Guidelines that the ASM Education Board has initiated to assist educators in the implementation process.

  18. News from the Library: A new key reference work for the engineer: ASME's Boiler and Pressure Vessel Code at the CERN Library

    CERN Multimedia

    CERN Library

    2011-01-01

    The Library is aiming at offering a range of constantly updated reference books, to cover all areas of CERN activity. A recent addition to our collections strengthens our offer in the Engineering field.   The CERN Library now holds a copy of the complete ASME Boiler and Pressure Vessel Code, 2010 edition. This code establishes rules of safety governing the design, fabrication, and inspection of boilers and pressure vessels, and nuclear power plant components during construction. This document is considered worldwide as a reference for mechanical design and is therefore important for the CERN community. The Code published by ASME (American Society of Mechanical Engineers) is kept current by the Boiler and Pressure Committee, a volunteer group of more than 950 engineers worldwide. The Committee meets regularly to consider requests for interpretations, revision, and to develop new rules. The CERN Library receives updates and includes them in the volumes until the next edition, which is expected to ...

  19. 78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Science.gov (United States)

    2013-05-01

    ... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

  20. 77 FR 70484 - Preoperational Testing of Onsite Electric Power Systems To Verify Proper Load Group Assignments...

    Science.gov (United States)

    2012-11-26

    ...-1294, ``Preoperational Testing of On-Site Electric Power Systems to Verify Proper Load Group... entitled ``Preoperational Testing of On- Site Electric Power Systems to Verify Proper Load Group... Electric Power Systems to Verify Proper Load Group Assignments, Electrical Separation, and Redundancy...

  1. Acid sphingomyelinase (Asm) deficiency patients in The Netherlands and Belgium: disease spectrum and natural course in attenuated patients.

    Science.gov (United States)

    Hollak, C E M; de Sonnaville, E S V; Cassiman, D; Linthorst, G E; Groener, J E; Morava, E; Wevers, R A; Mannens, M; Aerts, J M F G; Meersseman, W; Akkerman, E; Niezen-Koning, K E; Mulder, M F; Visser, G; Wijburg, F A; Lefeber, D; Poorthuis, B J H M

    2012-11-01

    Niemann-Pick disease (NPD) is a neurovisceral lysosomal storage disorder caused by acid sphingomyelinase (ASM) deficiency, which can be categorized as either Niemann-Pick disease type A [NPD-A], with progressive neurological disease and death in early childhood, or as Niemann-Pick disease type B [NPD-B], with a more variable spectrum of manifestations. Enzyme replacement therapy (ERT) with recombinant sphingomyelinase is currently studied as potential treatment for NPD-B patients. The objective of this study is to characterize the clinical features of patients with ASM deficiency in the Netherlands and Belgium with focus on the natural disease course of NPD-B patients. Prospective and retrospective data on ASM deficient patients were collected in The Netherlands and part of Belgium. Patients with NPD-B that could be followed prospectively were evaluated every 6-12 months for pulmonary function tests, 6 minute walk test (6 MWT), imaging (bone marrow infiltration measured by QCSI, organ volumes by MRI and CT scan of the lungs) and biochemical markers. Twenty-five patients with ASM deficiency were identified (13 males, 12 females, median age 13years, range 1-59 years). Nine patients had died at the time of the study, including four NPD-A patients at the age of 1,1, 2, 3 and five NPDB patents at the age of 5, 6, 43, 56 and 60 years. There was a high prevalence of homozygosity and compound heterozygosity for the common p.Arg608del mutation in 43% and 19% of NPD-B patients, respectively. In NPD-B patients, thrombocytopenia was present in most, while anemia and leucopenia were less common (33% and 6 % respectively). HDL cholesterol was reduced in most patients. Pulmonary disease was severe in several patients. Follow-up up to 11 years revealed a gradual decrease in platelet count. Detailed investigations in 6 NPD-B patients with follow-up in 4 patients revealed remarkable stable disease parameters up to 6 years, with some decline in pulmonary function and 6 MWT. Bone

  2. Interim status report on the revision of ASME PTC 12.1 -- closed feedwater heaters

    International Nuclear Information System (INIS)

    Stellern, J.L.; Hoobler, J.V.; Milton, J.W.; Welch, T.; Kona, C.; Thompson, H.N.; Tsou, J.L.

    1993-01-01

    The ASME Performance Test Code (PTC) 12.1-1978 for the performance testing of feedwater heaters is being revised extensively and updated. The committee anticipates that the final draft of the proposed Code will be ready for industry review in 1993. This Code revision will greatly enhance the usefulness and cost effectiveness of feedwater heater performance testing. This paper has been prepared to report on the progress of the committee and to disseminate information on the nature of the revision. Included in this paper are some of the notable changes intended for the Code. The most extensive change is the calculation method, which is described in step-by-step detail. An approach is also described for using ultrasonic flow techniques to test individual or split-string feedwater heaters, when flow nozzles are not available. Additionally some educational information on the use and limitations of ultrasonic measurement instrumentation is included. Discussion is also included on the required uncertainty analysis. 3 refs., 2 figs., 2 tabs

  3. NRC needs and their implementation-ASME Section IX code

    International Nuclear Information System (INIS)

    Liaw, B.D.

    1985-01-01

    The guiding principle from the onset of government regulation for the peaceful use of nuclear energy has been to prescribe only the minimum requirements that are needed for safety. In the pioneer regulators' collective mind, the technical details could be left to the regulated industry through its agents like NSSS vendors and A/E's and their surrogate organizations like ASME, ANS, AIF, etc. However, it has evolved through the years, due either to the bureaucratic momentum or the vacuum in industry leadership, into a situation where one sees an ever increasing number of detailed ''requirements'' prescribed by the regulators. Within the scope of activities covered by Section XI, there is no exception: e.g., NUREG-067, -0531 -1061; NUREG-0313 Rev. 0, Rev. 1, and now Rev. 2; IE Bulletins 82-03, 83-02; and Generic Letters 84-11, and 84-07, etc. for one issue of pipe crack alone; and there are more to come. There appears a consensus among all concerned parties including regulators that this is not a desirable situation and that something must be done to reverse this trend. The purpose of this discussion is, therefore, to explore the areas where the Section XI Code can be restructured to meet this need, and to seek ideas from the representatives of the regulated industry on the methods of implementation that are effective, efficient, and acceptable to all concerned parties

  4. Challenge - converting to ANSI/ASME NQA Standard 1

    International Nuclear Information System (INIS)

    Nansen, J.N.

    1983-03-01

    Prior to August, 1980, Westinghouse Hanford Company operated the Hanford Engineering Development Laboratory Quality Program in accordance with the requirements of RDT F 2-2T and RDT F 2-4T and for a short while to ANSI N45.2. Following the issuance of the national consensus standard ANSI/ ASME NQA Standard 1 in August 1979, and its acceptance by DOE, the Laboratory began making changes to meet these requirements in its internal quality program in August 1980. This was followed with the invoking of these requirements on supplier's programs and all procurement activities in March of 1981. This conversion was completed in approximately six months and provided an opportunity for several improvements. The keynote of this standard is flexibility. A major improvement noted was in the flexibility of application. The program can be tailored to meet specific needs. Vendor surveys under NQA-1 have shown positive results in that the organization and structure of the standard is much easier to follow and understand

  5. Review of ASME-NH Design Materials for Creep-Fatigue

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Kim, Jong Bum

    2010-01-01

    To review and recommend the candidate design materials for the Sodium-Cooled Fast Reactor, the material sensitivity evaluations by the comparison of design data between the ASME-NH materials were performed by using the SIE ASME-NH computer program implementing the material database of the ASME-NH. The design material data provided by the ASME-NH code are the elastic modulus and yield Strength, Time-Independent Allowable Stress Intensity value, time-dependent allowable stress intensity value, expected minimum stress-to rupture value, stress rupture Factors for weldment, isochronous stress-strain curves, and design fatigue curves. Among these, the data related with the creep-fatigue evaluation are investigated in this study

  6. 115-year-old society knows how to reach young scientists: ASM Young Ambassador Program.

    Science.gov (United States)

    Karczewska-Golec, Joanna

    2015-12-25

    With around 40,000 members in more than 150 countries, American Society for Microbiology (ASM) faces the challenge of meeting very diverse needs of its increasingly international members base. The newly launched ASM Young Ambassador Program seeks to aid the Society in this effort. Equipped with ASM conceptual support and financing, Young Ambassadors (YAs) design and pursue country-tailored approaches to strengthen the Society's ties with local microbiological communities. In a trans-national setting, the active presence of YAs at important scientific events, such as 16th European Congress on Biotechnology, forges new interactions between ASM and sister societies. The paper presents an overview of the Young Ambassadors-driven initiatives at both global and country levels, and explores the topic of how early-career scientists can contribute to science diplomacy and international relations. Copyright © 2014 Elsevier B.V. All rights reserved.

  7. Mechanisation of ultrasonic testing in nuclear power plants

    International Nuclear Information System (INIS)

    Seifert, W.

    1979-01-01

    Mechanical ultrasonic testing devices should meet the following requirements: Remote-controlled or automatic guidance of the US test systems at the test site according to given test parameters; exact positioning of the test system at the test site; high start-up accuracy and reproducibility; access to test regions that are hardly accessible or inaccessible for manual inspection; reduction of the radiation exposure of the operating personnel, and short assembling and testing time. The manipulators developed according to these requirements permit meandering test courses of the US test system on the pressure vessel surface or circular or semicircular courses around the nozzles or pipes in order to test welds and pipe joints. Every movement of the test system is taken up by a transmitting apparatus. (orig./HP) [de

  8. Real time test bed development for power system operation, control and cyber security

    Science.gov (United States)

    Reddi, Ram Mohan

    The operation and control of the power system in an efficient way is important in order to keep the system secure, reliable and economical. With advancements in smart grid, several new algorithms have been developed for improved operation and control. These algorithms need to be extensively tested and validated in real time before applying to the real electric power grid. This work focuses on the development of a real time test bed for testing and validating power system control algorithms, hardware devices and cyber security vulnerability. The test bed developed utilizes several hardware components including relays, phasor measurement units, phasor data concentrator, programmable logic controllers and several software tools. Current work also integrates historian for power system monitoring and data archiving. Finally, two different power system test cases are simulated to demonstrate the applications of developed test bed. The developed test bed can also be used for power system education.

  9. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

  10. Implementation of an Electric Vehicle Test Bed Controlled by a Virtual Power Plant for Contributing to Regulating Power Reserves

    DEFF Research Database (Denmark)

    Marra, Francesco; Sacchetti, Dario; Pedersen, Anders Bro

    2012-01-01

    and communication interfaces, is able to respond in real-time to smart grid control signals. The EV test bed is equipped with a Lithium-ion battery pack, a Battery Management System (BMS), a charger and a Vehicle-to-Grid (V2G) unit for feeding power back to the grid. The designed solution serves......With the increased focus on Electric Vehicles (EV) research and the potential benefits they bring for smart grid applications, there is a growing need for an evaluation platform connected to the electricity grid. This paper addresses the design of an EV test bed, which using real EV components...... requests from the Danish TSO are used as a proof-of-concept, to demonstrate the EV test bed power response. Test results have proven the capability to respond to frequent power control requests and they reveal the potential EV ability for contributing to regulating power reserves....

  11. Sustaining neutral beam power supply system for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Eckard, R.D.; Wilson, J.H.; Van Ness, H.W.

    1980-01-01

    In late August 1978, a fixed price procurement contract for $25,000,000 was awarded to Aydin Energy Division, Palo Alto, California, for the design, manufacture, installation and acceptance testing of the Lawrence Livermore National Laboratory Mirror Fusion Test Facility (MFTF) Sustaining Neutral Beam Power Supply System (SNBPSS). This system of 24 power supply sets will provide the conditioned power for the 24 neutral beam source modules. Each set will provide the accel potential the arc power, the filament power, and the suppressor power for its associated neutral beam source module. The design and development of the SNBPSS has progressed through the final design phase and is now in production. Testing of the major sub-assembly power supply is proceeding at Aydin and the final acceptance testing of the first two power supplies at LLNL is expected to be completed this year

  12. Inhibition of allergen-induced basophil activation by ASM-024, a nicotinic receptor ligand.

    Science.gov (United States)

    Watson, Brittany M; Oliveria, John Paul; Nusca, Graeme M; Smith, Steven G; Beaudin, Sue; Dua, Benny; Watson, Rick M; Assayag, Evelynne Israël; Cormier, Yvon F; Sehmi, Roma; Gauvreau, Gail M

    2014-01-01

    Nicotinic acetylcholine receptors (nAChRs) were identified on eosinophils and shown to regulate inflammatory responses, but nAChR expression on basophils has not been explored yet. We investigated surface receptor expression of nAChR α4, α7 and α1/α3/α5 subunits on basophils. Furthermore, we examined the effects of ASM-024, a synthetic nicotinic ligand, on in vitro anti-IgE and in vivo allergen-induced basophil activation. Basophils were enriched from the peripheral blood of allergic donors and the expression of nAChR subunits and muscarinic receptors was determined. Purified basophils were stimulated with anti-IgE in the presence of ASM-024 with or without muscarinic or nicotinic antagonists for the measurement of CD203c expression and histamine release. The effect of 9 days of treatment with 50 and 200 mg ASM-024 on basophil CD203c expression was examined in the blood of mild allergic asthmatics before and after allergen inhalation challenge. nAChR α4, α7 and α1/α3/α5 receptor subunit expression was detected on basophils. Stimulation of basophils with anti-IgE increased CD203c expression and histamine release, which was inhibited by ASM-024 (10(-5) to 10(-)(3) M, p ASM-024 was reversed in the presence of muscarinic and nicotinic antagonists. In subjects with mild asthma, ASM-024 inhalation significantly inhibited basophil CD203c expression measured 24 h after allergen challenge (p = 0.03). This study shows that ASM-024 inhibits IgE- and allergen-induced basophil activation through both nicotinic and muscarinic receptors, and suggests that ASM-024 may be an efficacious agent for modulating allergic asthma responses. © 2015 S. Karger AG, Basel.

  13. Acid Sphingomyelinase (ASM) is a Negative Regulator of Regulatory T Cell (Treg) Development.

    Science.gov (United States)

    Zhou, Yuetao; Salker, Madhuri S; Walker, Britta; Münzer, Patrick; Borst, Oliver; Gawaz, Meinrad; Gulbins, Erich; Singh, Yogesh; Lang, Florian

    2016-01-01

    Regulatory T cell (Treg) is required for the maintenance of tolerance to various tissue antigens and to protect the host from autoimmune disorders. However, Treg may, indirectly, support cancer progression and bacterial infections. Therefore, a balance of Treg function is pivotal for adequate immune responses. Acid sphingomyelinase (ASM) is a rate limiting enzyme involved in the production of ceramide by breaking down sphingomyelin. Previous studies in T-cells have suggested that ASM is involved in CD28 signalling, T lymphocyte granule secretion, degranulation, and vesicle shedding similar to the formation of phosphatidylserine-exposing microparticles from glial cells. However, whether ASM affects the development of Treg has not yet been described. Splenocytes, isolated Naive T lymphocytes and cultured T cells were characterized for various immune T cell markers by flow cytometery. Cell proliferation was measured by Carboxyfluorescein succinimidyl ester (CFSE) dye, cell cycle analysis by Propidium Iodide (PI), mRNA transcripts by q-RT PCR and protein expression by Western Blotting respectively. ASM deficient mice have higher number of Treg compared with littermate control mice. In vitro induction of ASM deficient T cells in the presence of TGF-β and IL-2 lead to a significantly higher number of Foxp3+ induced Treg (iTreg) compared with control T-cells. Further, ASM deficient iTreg has less AKT (serine 473) phosphorylation and Rictor levels compared with control iTreg. Ceramide C6 led to significant reduction of iTreg in both ASM deficient and WT mice. The reduction in iTreg leads to induction of IL-1β, IL-6 and IL-17 but not IFN-γ mRNA levels. ASM is a negative regulator of natural and iTreg. © 2016 The Author(s) Published by S. Karger AG, Basel.

  14. Empirical Power Comparison Of Goodness of Fit Tests for Normality In The Presence of Outliers

    International Nuclear Information System (INIS)

    Saculinggan, Mayette; Balase, Emily Amor

    2013-01-01

    Most statistical tests such as t-tests, linear regression analysis and Analysis of Variance (ANOVA) require the normality assumptions. When the normality assumption is violated, interpretation and inferences may not be reliable. Therefore it is important to assess such assumption before using any appropriate statistical test. One of the commonly used procedures in determining whether a random sample of size n comes from a normal population are the goodness-of-fit tests for normality. Several studies have already been conducted on the comparison of the different goodness-of-fit(see, for example [2]) but it is generally limited to the sample size or to the number of GOF tests being compared(see, for example [2] [5] [6] [7] [8]). This paper compares the power of six formal tests of normality: Kolmogorov-Smirnov test (see [3]), Anderson-Darling test, Shapiro-Wilk test, Lilliefors test, Chi-Square test (see [1]) and D'Agostino-Pearson test. Small, moderate and large sample sizes and various contamination levels were used to obtain the power of each test via Monte Carlo simulation. Ten thousand samples of each sample size and contamination level at a fixed type I error rate α were generated from the given alternative distribution. The power of each test was then obtained by comparing the normality test statistics with the respective critical values. Results show that the power of all six tests is low for small sample size(see, for example [2]). But for n = 20, the Shapiro-Wilk test and Anderson – Darling test have achieved high power. For n = 60, Shapiro-Wilk test and Liliefors test are most powerful. For large sample size, Shapiro-Wilk test is most powerful (see, for example [5]). However, the test that achieves the highest power under all conditions for large sample size is D'Agostino-Pearson test (see, for example [9]).

  15. Power-Hardware-In-the-Loop (PHIL) Test of VSC-based HVDC connection for Offshore Wind Power Plants (WPPs)

    DEFF Research Database (Denmark)

    Sharma, Ranjan; Cha, Seung-Tae; Wu, Qiuwei

    2011-01-01

    This paper presents a power-hardware-in-the-loop (PHIL) test for an offshore wind power plant (WPP) interconnected to the onshore grid by a VSC-based HVDC connection. The intention of the PHIL test is to verify the control coordination between the plant side converter of the HVDC connection...... the successful control coordination between the WPP and the plant side VSC converter of the HVDC connection of the WPP....

  16. New Technology for Microfabrication and Testing of a Thermoelectric Device for Generating Mobile Electrical Power

    Science.gov (United States)

    Prasad, Narashimha S.; Taylor, Patrick J.; Trivedi, Sudhir B.; Kutcher, Susan

    2010-01-01

    We report the results of fabrication and testing of a thermoelectric power generation module. The module was fabricated using a new "flip-chip" module assembly technique that is scalable and modular. This technique results in a low value of contact resistivity ( surfaces. Under mild testing, a power of 22 mW/sq cm was obtained from small (electrical power of practical and usable magnitude for remote applications using thermoelectric power generation technologies.

  17. Results of Laboratory Testing of Advanced Power Strips: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Earle, L.; Sparn, B.

    2012-08-01

    This paper describes the results of a laboratory investigation to evaluate the technical performance of advanced power strip (APS) devices when subjected to a range of home entertainment center and home office usage scenarios.

  18. Analysis of Power Transfer Efficiency of Standard Integrated Circuit Immunity Test Methods

    Directory of Open Access Journals (Sweden)

    Hai Au Huynh

    2015-01-01

    Full Text Available Direct power injection (DPI and bulk current injection (BCI methods are defined in IEC 62132-3 and IEC 62132-4 as the electromagnetic immunity test method of integrated circuits (IC. The forward power measured at the RF noise generator when the IC malfunctions is used as the measure of immunity level of the IC. However, the actual power that causes failure in ICs is different from forward power measured at the noise source. Power transfer efficiency is used as a measure of power loss of the noise injection path. In this paper, the power transfer efficiencies of DPI and BCI methods are derived and validated experimentally with immunity test setup of a clock divider IC. Power transfer efficiency varies significantly over the frequency range as a function of the test method used and the IC input impedance. For the frequency range of 15 kHz to 1 GHz, power transfer efficiency of the BCI test was constantly higher than that of the DPI test. In the DPI test, power transfer efficiency is particularly low in the lower test frequency range up to 10 MHz. When performing the IC immunity tests following the standards, these characteristics of the test methods need to be considered.

  19. Domestic nuclear power plants tested in the Earth's orbits

    International Nuclear Information System (INIS)

    Gryaznov, G.M.; Pupko, V.Ya.

    1995-01-01

    The authors review the history of developing the TOPAZ nuclear power units, their design, the fall of the Kosmos-954 satellite with a nuclear unit of thermoelectric type on the territory of Canada. The details of the structural scheme of the TOPAZ nuclear power unit, the main directions of its modification, some aspects of ensuring nuclear and radiation safety are considered as well as the importance of thermoemission units for telecommunication satellites

  20. A Minimal Path Searching Approach for Active Shape Model (ASM)-based Segmentation of the Lung.

    Science.gov (United States)

    Guo, Shengwen; Fei, Baowei

    2009-03-27

    We are developing a minimal path searching method for active shape model (ASM)-based segmentation for detection of lung boundaries on digital radiographs. With the conventional ASM method, the position and shape parameters of the model points are iteratively refined and the target points are updated by the least Mahalanobis distance criterion. We propose an improved searching strategy that extends the searching points in a fan-shape region instead of along the normal direction. A minimal path (MP) deformable model is applied to drive the searching procedure. A statistical shape prior model is incorporated into the segmentation. In order to keep the smoothness of the shape, a smooth constraint is employed to the deformable model. To quantitatively assess the ASM-MP segmentation, we compare the automatic segmentation with manual segmentation for 72 lung digitized radiographs. The distance error between the ASM-MP and manual segmentation is 1.75 ± 0.33 pixels, while the error is 1.99 ± 0.45 pixels for the ASM. Our results demonstrate that our ASM-MP method can accurately segment the lung on digital radiographs.

  1. A minimal path searching approach for active shape model (ASM)-based segmentation of the lung

    Science.gov (United States)

    Guo, Shengwen; Fei, Baowei

    2009-02-01

    We are developing a minimal path searching method for active shape model (ASM)-based segmentation for detection of lung boundaries on digital radiographs. With the conventional ASM method, the position and shape parameters of the model points are iteratively refined and the target points are updated by the least Mahalanobis distance criterion. We propose an improved searching strategy that extends the searching points in a fan-shape region instead of along the normal direction. A minimal path (MP) deformable model is applied to drive the searching procedure. A statistical shape prior model is incorporated into the segmentation. In order to keep the smoothness of the shape, a smooth constraint is employed to the deformable model. To quantitatively assess the ASM-MP segmentation, we compare the automatic segmentation with manual segmentation for 72 lung digitized radiographs. The distance error between the ASM-MP and manual segmentation is 1.75 +/- 0.33 pixels, while the error is 1.99 +/- 0.45 pixels for the ASM. Our results demonstrate that our ASM-MP method can accurately segment the lung on digital radiographs.

  2. Test and evaluation of load converter topologies used in the Space Station Freedom power management and distribution dc test bed

    Science.gov (United States)

    Lebron, Ramon C.; Oliver, Angela C.; Bodi, Robert F.

    1991-01-01

    Power components hardware in support of the Space Station freedom dc Electric Power System were tested. One type of breadboard hardware tested is the dc Load Converter Unit, which constitutes the power interface between the electric power system and the actual load. These units are dc to dc converters that provide the final system regulation before power is delivered to the load. Three load converters were tested: a series resonant converter, a series inductor switch-mode converter, and a switching full-bridge forward converter. The topology, operation principles, and test results are described, in general. A comparative analysis of the three units is given with respect to efficiency, regulation, short circuit behavior (protection), and transient characteristics.

  3. 14 CFR 33.96 - Engine tests in auxiliary power unit (APU) mode.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Engine tests in auxiliary power unit (APU... TRANSPORTATION AIRCRAFT AIRWORTHINESS STANDARDS: AIRCRAFT ENGINES Block Tests; Turbine Aircraft Engines § 33.96 Engine tests in auxiliary power unit (APU) mode. If the engine is designed with a propeller brake which...

  4. A Panel Test of Purchasing Power Parity Under the Null of Stationarity

    OpenAIRE

    Hunter, J; Simpson, M

    2001-01-01

    Purchasing Power Parity (PPP) is tested using a sample of real exchange rate data for twelve European countries. Acknowledging that Augmented Dickey Fuller tests have low power, we apply a Panel test that considers the null of stationarity and corrects for serial dependence using a non-parametric kernel based method.

  5. Safety Evaluation Report, pump and valve inservice testing program, Yankee Rowe Nuclear Power Station (Docket No. 50-29). Revision 1

    International Nuclear Information System (INIS)

    Ransom, C.B.; Rockhold, H.C.

    1985-10-01

    The staff reviewed the Yankee Rowe IST program submittal dated October 2, 1981 and evaluated the proposed tests for compliance with the requirements of the ASME Boiler and Pressure Vessel Code, Section XI, 1977 Edition, through the Summer of 1978 Addenda. Those items not in compliance were discussed in a working session with Yankee Atomic Electric Company personnel, USNRC representatives, and EG and G Idaho reviewers on June 22, 1982. The evaluations presented in this SER of the Yankee Rowe pump and valve inservice testing program and the associated relief requests are those of the NRC staff. These findings apply only to component testing

  6. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  7. Thermal neutron measurement using the instrumented test bundle and assessment of maximum linear power in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. S.; Seo, C. K.; Lee, B. C.; Kim, H. N.; Kang, B. W. [KAERI, Taejon (Korea, Republic of)

    2000-10-01

    The HANARO fuel, U{sub 3}Si-Al, has been developed by AECL and tested in NRU reactor. Due to the lack of the data performed under the high power, the repetitive conduct of the irradiation test was required under the power greater than 108kW/m, which is the estimated maximum linear power in the design stage. Accordingly, the instrumented test bundle with SPND(Self Powered Neutron Detector) was fabricated and its irradiation test was performed in IR2 of HANARO. The measured thermal neutron flux with SPND is compared with calculation results by HANAFMS(HANARO Fuel Management System). The difference in the measured and calculated thermal flux values are below {+-}11% and the accuracy of the linear power predicted by HANAFMS is consequently accompanied. Therefore, it is believed that the maximum linear power above 120kW/m is achieved during the irradiation test of the test bundle.

  8. Development of the present reference fracture toughness curves in the ASME nuclear code

    International Nuclear Information System (INIS)

    Yukawa, S.; Merkle, J.G.

    1984-01-01

    Since the early 1970's, the Sections of the ASME Boiler and Pressure Vessel Code concerned with nuclear power plant components have included fracture mechanics procedures to analyze the effects of postulated or detected flaws. These procedures are contained in Appendix G of Section III and in Appendix A of Section XI of the Code. Specifically, Appendix G procedures are concerned with designing for protection against nonductile failures while Appendix A procedures are for evaluating the disposition of flaws detected during in-service inspection. An important element of the procedures is the inclusion of recommended material fracture toughness values. This paper describes the origin and development of these recommended fracture toughness values. Since these values appear in the Code in a graphical format, the values are often referred to as reference toughness curves. In the context of Code terminology, reference toughness means the allowable values of fracture toughness for the materials of concern that can be used in conjunction with the analytical procedures of Appendices G and A. The paper discusses the basis and rationale underlying the original formulation of these reference toughness curves and the modifications incorporated into them in the course of their adoption into the Code

  9. Evaluation of the control system checkout test at 100% power for Yonggwang Nuclear Power Plant Unit 3

    International Nuclear Information System (INIS)

    Kim, Shin Whan; Lee, Joo Han; Baek, Jong Man; Seo, Jong Tae; Lee, Sang Keun; Kang, In Koo; Ju, Hee Wan; Min, Kyung Soo; Kim, Byung Gon

    1995-01-01

    Control system checkout tests at various powers for Yonggwang Nuclear Power Plant Unit 3(YGN3) were performed to demonstrate the accuracies and proper performances of the control systems of the plant. Tested control systems included the feedwater control system, steam bypass control system, reactor regulation system, control element drive mechanism control system, pressurizer level control system, and pressurizer pressure control system. The measured test data during the control system checkout test at 100% power are evaluated. The test results showed that the control systems of YGN 3 properly control system was simulated by using the LTC code which is the performance analysis code for YGN 3 and 4 design. Comparisons of the predicted results with the measured data confirmed that the feedwater control system controls the steam generator level as designed

  10. Testing for cross-subsidisation in the combined heat and power generation sector

    DEFF Research Database (Denmark)

    Amundsen, Eirik S; Andersen, Per; Jensen, Frank

    2011-01-01

    In this paper we examine cross-subsidisation among combined heat and power producers in Denmark.Information on stand-alone costs for heat generation allows us to empirically compare the Faulhaber tests,tests with an upper bound on stand-alone costs (the Palmer tests) and the fully distributed cos...... test (FDC). All tests indicate a substantial amount of cross-subsidisation from heat generation to power generation. It is shown that the FDC test is closer to that of the Faulhaber tests in its results than the Palmer tests. Thus as the Faulhaber tests are considered in the literature...

  11. GC-ASM: Synergistic Integration of Graph-Cut and Active Shape Model Strategies for Medical Image Segmentation.

    Science.gov (United States)

    Chen, Xinjian; Udupa, Jayaram K; Alavi, Abass; Torigian, Drew A

    2013-05-01

    Image segmentation methods may be classified into two categories: purely image based and model based. Each of these two classes has its own advantages and disadvantages. In this paper, we propose a novel synergistic combination of the image based graph-cut (GC) method with the model based ASM method to arrive at the GC-ASM method for medical image segmentation. A multi-object GC cost function is proposed which effectively integrates the ASM shape information into the GC framework. The proposed method consists of two phases: model building and segmentation. In the model building phase, the ASM model is built and the parameters of the GC are estimated. The segmentation phase consists of two main steps: initialization (recognition) and delineation. For initialization, an automatic method is proposed which estimates the pose (translation, orientation, and scale) of the model, and obtains a rough segmentation result which also provides the shape information for the GC method. For delineation, an iterative GC-ASM algorithm is proposed which performs finer delineation based on the initialization results. The proposed methods are implemented to operate on 2D images and evaluated on clinical chest CT, abdominal CT, and foot MRI data sets. The results show the following: (a) An overall delineation accuracy of TPVF > 96%, FPVF ASM for different objects, modalities, and body regions. (b) GC-ASM improves over ASM in its accuracy and precision to search region. (c) GC-ASM requires far fewer landmarks (about 1/3 of ASM) than ASM. (d) GC-ASM achieves full automation in the segmentation step compared to GC which requires seed specification and improves on the accuracy of GC. (e) One disadvantage of GC-ASM is its increased computational expense owing to the iterative nature of the algorithm.

  12. Thermal hydro-dynamic behavior of the nuclear-powered ship Mutsu in the power-up test

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Tanaka, Yoshimi; Yao, Toshiaki; Inoue, Kimio; Ochiai, Masa-aki

    1991-01-01

    The first Japanese nuclear ship N.S. Mutsu was tested with heavy load changes as ones of the full scale start-up tests. This paper describes the results of power increasing test, test of crash astern propulsion from ahead cruising, and turbine trip test; In the power increasing test, the main steam flow rate increased from 20 % flow to 70 % flow in 13 seconds; In the test of crash astern propulsion from ahead cruising, an astern turbine maneuvering handle was drawn to the position equivalent to about 60 % reactor power in 30 seconds after an ahead turbine stopping; In the turbine trip test, the ahead turbine was tripped manually at about 100 % power operation. From these tests it was verified that the Mutsu reactor is capable to respond smoothly and safely to the heavy load change without reactor scram, without workings of the pressurizer relief valve and main steam dump valve, and without need of any manual adjustment of the reactor control system. (author)

  13. IEEE guide for planning of pre-operational testing programs for class 1E power systems for nuclear-power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The Institute of Electrical and Electronics Engineers (IEEE) guide for pre-operational testing of Class 1E power systems for nuclear-power generating stations is presented. The guidelines apply to power systems both ac and dc supplies but not to the equipment which utilizes the ac and dc power. The pre-operational tests are performed after appropriate construction tests

  14. Estimation of sample size and testing power (Part 3).

    Science.gov (United States)

    Hu, Liang-ping; Bao, Xiao-lei; Guan, Xue; Zhou, Shi-guo

    2011-12-01

    This article introduces the definition and sample size estimation of three special tests (namely, non-inferiority test, equivalence test and superiority test) for qualitative data with the design of one factor with two levels having a binary response variable. Non-inferiority test refers to the research design of which the objective is to verify that the efficacy of the experimental drug is not clinically inferior to that of the positive control drug. Equivalence test refers to the research design of which the objective is to verify that the experimental drug and the control drug have clinically equivalent efficacy. Superiority test refers to the research design of which the objective is to verify that the efficacy of the experimental drug is clinically superior to that of the control drug. By specific examples, this article introduces formulas of sample size estimation for the three special tests, and their SAS realization in detail.

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  17. An investigation of the statistical power of neutrality tests based on comparative and population genetic data

    DEFF Research Database (Denmark)

    Zhai, Weiwei; Nielsen, Rasmus; Slatkin, Montgomery

    2009-01-01

    In this report, we investigate the statistical power of several tests of selective neutrality based on patterns of genetic diversity within and between species. The goal is to compare tests based solely on population genetic data with tests using comparative data or a combination of comparative...... and population genetic data. We show that in the presence of repeated selective sweeps on relatively neutral background, tests based on the d(N)/d(S) ratios in comparative data almost always have more power to detect selection than tests based on population genetic data, even if the overall level of divergence...... selection. The Hudson-Kreitman-Aguadé test is the most powerful test for detecting positive selection among the population genetic tests investigated, whereas McDonald-Kreitman test typically has more power to detect negative selection. We discuss our findings in the light of the discordant results obtained...

  18. Approaching Repetitive Short Circuit Tests on MW-Scale Power Modules by means of an Automatic Testing Setup

    DEFF Research Database (Denmark)

    Reigosa, Paula Diaz; Wang, Huai; Iannuzzo, Francesco

    2016-01-01

    An automatic testing system to perform repetitive short-circuit tests on megawatt-scale IGBT power modules is pre-sented and described in this paper, pointing out the advantages and features of such testing approach. The developed system is based on a non-destructive short-circuit tester, which has...

  19. Design of an Adaptive Power Regulation Mechanism and a Nozzle for a Hydroelectric Power Plant Turbine Test Rig

    Science.gov (United States)

    Mert, Burak; Aytac, Zeynep; Tascioglu, Yigit; Celebioglu, Kutay; Aradag, Selin; ETU Hydro Research Center Team

    2014-11-01

    This study deals with the design of a power regulation mechanism for a Hydroelectric Power Plant (HEPP) model turbine test system which is designed to test Francis type hydroturbines up to 2 MW power with varying head and flow(discharge) values. Unlike the tailor made regulation mechanisms of full-sized, functional HEPPs; the design for the test system must be easily adapted to various turbines that are to be tested. In order to achieve this adaptability, a dynamic simulation model is constructed in MATLAB/Simulink SimMechanics. This model acquires geometric data and hydraulic loading data of the regulation system from Autodesk Inventor CAD models and Computational Fluid Dynamics (CFD) analysis respectively. The dynamic model is explained and case studies of two different HEPPs are performed for validation. CFD aided design of the turbine guide vanes, which is used as input for the dynamic model, is also presented. This research is financially supported by Turkish Ministry of Development.

  20. OOK power model based dynamic error testing for smart electricity meter

    International Nuclear Information System (INIS)

    Wang, Xuewei; Chen, Jingxia; Jia, Xiaolu; Zhu, Meng; Yuan, Ruiming; Jiang, Zhenyu

    2017-01-01

    This paper formulates the dynamic error testing problem for a smart meter, with consideration and investigation of both the testing signal and the dynamic error testing method. To solve the dynamic error testing problems, the paper establishes an on-off-keying (OOK) testing dynamic current model and an OOK testing dynamic load energy (TDLE) model. Then two types of TDLE sequences and three modes of OOK testing dynamic power are proposed. In addition, a novel algorithm, which helps to solve the problem of dynamic electric energy measurement’s traceability, is derived for dynamic errors. Based on the above researches, OOK TDLE sequence generation equipment is developed and a dynamic error testing system is constructed. Using the testing system, five kinds of meters were tested in the three dynamic power modes. The test results show that the dynamic error is closely related to dynamic power mode and the measurement uncertainty is 0.38%. (paper)

  1. OOK power model based dynamic error testing for smart electricity meter

    Science.gov (United States)

    Wang, Xuewei; Chen, Jingxia; Yuan, Ruiming; Jia, Xiaolu; Zhu, Meng; Jiang, Zhenyu

    2017-02-01

    This paper formulates the dynamic error testing problem for a smart meter, with consideration and investigation of both the testing signal and the dynamic error testing method. To solve the dynamic error testing problems, the paper establishes an on-off-keying (OOK) testing dynamic current model and an OOK testing dynamic load energy (TDLE) model. Then two types of TDLE sequences and three modes of OOK testing dynamic power are proposed. In addition, a novel algorithm, which helps to solve the problem of dynamic electric energy measurement’s traceability, is derived for dynamic errors. Based on the above researches, OOK TDLE sequence generation equipment is developed and a dynamic error testing system is constructed. Using the testing system, five kinds of meters were tested in the three dynamic power modes. The test results show that the dynamic error is closely related to dynamic power mode and the measurement uncertainty is 0.38%.

  2. The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Vesterlund, G.; Vaernild, O.

    1980-01-01

    The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

  3. 78 FR 67206 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Science.gov (United States)

    2013-11-08

    ... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Revision to regulatory guide; issuance..., ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants.'' This RG is being revised to provide... Operators Installed Inside the Containment of Nuclear Power Plants,'' dated January 1974. ADDRESSES: Please...

  4. Evaluation of snubber functional test methods: Tier 1

    International Nuclear Information System (INIS)

    Brown, D.P.

    1993-07-01

    The objective of the research is to establish technical bases in support of efforts on the part of the Snubber Utility Group (SNUG) and the Subsection ISTD Working Group of the ASME O ampersand M Code in developing guidelines and methodologies for snubber functional testing to ensure that snubbers are tested in a manner that ensures reliable and meaningful test results. The methodology used in this research includes both a review of available industry information as well as the testing of different snubber models using various test machines. Information is provided pertaining to current industry practices in regard to snubber testing including recommended test procedures, technical description of various test machines, and the number and types of snubbers used in the nuclear power industry. A review of previous test methodology research conducted by the Snubber Utility Group is also included. The effects of variations in controllable test parameters on snubber test results are discussed. Also included are the results of confirmatory tests in which various snubber models were tested using various test machines. Recommendations are provided for standard test methods to be included in Subsection ISTD of the ASME O ampersand M Code [4]. General information and recommendations are provided that may be used by utility personnel in specifying snubber test equipment that is most suited for plant-specific needs as well as information that may be effectively used in the review and interpretation of test results

  5. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  6. Test of job performance aids for power plants. Final report

    International Nuclear Information System (INIS)

    Shriver, E.L.; Zach, S.E.; Foley, J.P. Jr.

    1982-10-01

    The objective of EPRI Research Project 1396-1 was to evaluate the applicability and effectiveness of Job Performance Aids (JPAs) in nuclear power plant situations. For over twenty years, JPAs have been developed in military situations to meet the problems of confusing, incomplete, and inaccurate procedures on maintenance jobs. Kinton, Incorporated of Alexandria, Virginia applied the military experience with JPAs to nuclear power plant situations and identified potential benefits in terms of cost reductions and improved performance. Sample JPAs were developed for Control Room Operations, Maintenance, Plant Operations, Instrumentation and Control, Health Physics, and Quality Assurance tasks (procedures) in selected nuclear plants. JPAs were also developed for a prototype condenser tube leak detection system in the design stage, as well as for generic classes of circuit breaker equipment. Based on the results of the study, the use of JPAs is recommended for plant procedures of medium to high difficulty and for those tasks performed infrequently, even if fairly simple

  7. Life management of power plant based on structural damage testing

    Energy Technology Data Exchange (ETDEWEB)

    Tallermo, H.; Klevtsov, I. [Thermal Engineering Department of Tallinn Technical University, Tallinn (Estonia); Arras, V. [Eesti Energia, Tallinn (Estonia)

    1998-12-31

    Life management system is based on the valid nowadays in Estonian power plants regulation documentation. The system allows to estimate stress distribution in components, find computational assessment of cumulated creep damage, determine when and where it is necessary to cut off the particular number of microsamples or take replicas. Finally, the real metal condition may be assessed on the basis of metallographic specimen research and reasonable 3-R decision - run, repair, replacement - made on further component use. (orig.) 6 refs.

  8. Durability Testing of Additively Manufactured High Power Microwave Structures

    Science.gov (United States)

    2017-10-29

    D. Hoffman, B. Singh and J. Thomas, Handbook of vacuum science and technology . San Diego, CA: Academic Press, 1998. [17] M. Cuneo, "The effect of...the higher He diffusion rate through rubber o-rings, its lower pumping speed within the cryopump and possibly saturation of the charcoal arrays...Third Edition, CRC Press (Boca Raton, FL, USA, 2016). [3] J. C. Whitaker, “Power Vacuum Tubes Handbook ”, Second Edition, CRC Press (New York, NY, USA

  9. Life management of power plant based on structural damage testing

    Energy Technology Data Exchange (ETDEWEB)

    Tallermo, H; Klevtsov, I [Thermal Engineering Department of Tallinn Technical University, Tallinn (Estonia); Arras, V [Eesti Energia, Tallinn (Estonia)

    1999-12-31

    Life management system is based on the valid nowadays in Estonian power plants regulation documentation. The system allows to estimate stress distribution in components, find computational assessment of cumulated creep damage, determine when and where it is necessary to cut off the particular number of microsamples or take replicas. Finally, the real metal condition may be assessed on the basis of metallographic specimen research and reasonable 3-R decision - run, repair, replacement - made on further component use. (orig.) 6 refs.

  10. Seismic analysis and testing of nuclear power plants

    International Nuclear Information System (INIS)

    1979-01-01

    The following subjects are discussed in this guide: General Recommendations for seismic classification, loading combinations and allowable limits; seismic analysis methods; implications for seismic design; seismic testing and qualification; seismic instrumentation; modelling techniques; material property characterization; seismic response of soil deposits and earth structures; liquefaction and ground failure; slope stability; sloshing effects in water pools; qualification testing by means of the transport vehicle

  11. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  12. Sustaining neutral beam power supply system for the Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Eckard, R.D.; Van Ness, H.W.

    1979-01-01

    A fixed-price procurement contract for $24.9 million was awarded to Aydin Energy Division, Palo Alto, CA, for the design, manufacture, installation, and acceptance testing of the sustaining neutral beam power supply system (SNBPSS). This system is scheduled for completion in early 1981 and will provide the conditioned power for the 24 neutral beam source modules. Each of the 24 power supply sets will provide the accel potential of 80 kV at 88 A, the arc power, the filament power, and the suppressor power for its associated neutral beam source module

  13. Comparisons of ANS, ASME, AWS, and NFPA standards cited in the NRC standard review plan, NUREG-0800, and related documents

    International Nuclear Information System (INIS)

    Ankrum, A.R.; Bohlander, K.L.; Gilbert, E.R.; Spiesman, J.B.

    1995-11-01

    This report provides the results of comparisons of the cited and latest versions of ANS, ASME, AWS and NFPA standards cited in the NRC Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (NUREG 0800) and related documents. The comparisons were performed by Battelle Pacific Northwest Laboratories in support of the NRC's Standard Review Plan Update and Development Program. Significant changes to the standards, from the cited version to the latest version, are described and discussed in a tabular format for each standard. Recommendations for updating each citation in the Standard Review Plan are presented. Technical considerations and suggested changes are included for related regulatory documents (i.e., Regulatory Guides and the Code of Federal Regulations) citing the standard. The results and recommendations presented in this document have not been subjected to NRC staff review

  14. Photovoltaic power system tests on an 8-kilowatt single-phase line-commutated inverter

    Science.gov (United States)

    Stover, J. B.

    1978-01-01

    Efficiency and power factor were measured as functions of solar array voltage and current. The effects of input shunt capacitance and series inductance were determined. Tests were conducted from 15 to 75 percent of the 8 kW rated inverter input power. Measured efficiencies ranged from 76 percent to 88 percent at about 50 percent of rated inverter input power. Power factor ranged from 36 percent to 72 percent.

  15. Creep property testing of energy power plant component material

    International Nuclear Information System (INIS)

    Nitiswati, Sri; Histori; Triyadi, Ari; Haryanto, Mudi

    1999-01-01

    Creep testing of SA213 T12 boiler piping material from fossil plant, Suralaya has been done. The aim of the testing is to know the creep behaviour of SA213 T12 boiler piping material which has been used more than 10 yeas, what is the material still followed ideal creep curve (there are primary stage, secondary stage, and tertiary stage). This possibility could happened because the material which has been used more than 10 years usually will be through ageing process because corrosion. The testing was conducted in 520 0C, with variety load between 4% until 50% maximum allowable load based on strength of the material in 520 0C

  16. 40 CFR 86.1772-99 - Road load power, test weight, and inertia weight class determination.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 19 2010-07-01 2010-07-01 false Road load power, test weight, and inertia weight class determination. 86.1772-99 Section 86.1772-99 Protection of Environment ENVIRONMENTAL... for Light-Duty Vehicles and Light-Duty Trucks § 86.1772-99 Road load power, test weight, and inertia...

  17. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    Energy Technology Data Exchange (ETDEWEB)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

    2016-10-01

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

  18. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  19. Performance Test Results for the Laser-Powered Microthruster

    International Nuclear Information System (INIS)

    Phipps, Claude R.; Luke, James R.; Helgeson, Wesley; Johnson, Richard

    2006-01-01

    Microthrusters are useful for orienting and repositioning small craft above the atmosphere. We report technical results obtained during a successful 5-year program to develop a commercially-viable laser-powered microthruster. Its main advantage is the ability to generate a broad thrust range under programmable electronic control with minimal electrical power. The device applies millisecond-duration diode-laser pulses to a fuel tape to produce an ablation jet. By employing laser-initiated energetic polymers in our ablation fuel tapes, we obtained momentum coupling coefficients as large as 3mN/W of incident laser power, giving a continuous thrust range from 50μN to 10mN. With our standard 30m x 8mm fuel tape, fueled thruster mass is 0.5kg and 50N-s lifetime impulse is achieved. With an order-of-magnitude greater fuel mass, the thruster could accomplish re-entry or substantial orbit-raising of a 10-kg microsatellite. In its usual configuration, specific impulse is 200 seconds, and ablation efficiency, the ratio of exhaust kinetic energy to incident laser optical energy is 180%. We compare performance of several laser-initiated micropropellants which we studied, including polyvinyl nitrate (PVN), glycidyl azide polymer (GAP), and nitrocellulose (NC). All were doped with a laser-absorbing component, either carbon nanopearls with 10nm mean diameter or dyes tuned to the 920-nm laser wavelength but transparent at visible wavelengths. Our demonstrated momentum coupling coefficient is sufficient to levitate a 0.15-kg object with a 500-W laser beam having appropriate characteristics

  20. Three-dimensional testing of power plant components

    International Nuclear Information System (INIS)

    Martin, A.

    1989-01-01

    Industrial photogrammetry is a dimensional checking procedure whose main advantages are the fast acquisition of the basic data (image), contactless inspection, and independent data processing. As a result of these basic characteristics, photogrammetry is particularly well suited to the maintenance of nuclear power plants. Since 1983, Framatome has employed photogrammetry in a number of cases for 3D dimensional checks and inspections of systems for repair purposes. To this day, e.g., the tube plates of steam generators have been inspected, the dimensional stability of the support rings in steam generators have been checked, and the alignment pins of fuel elements have been examined in this way. (orig.) [de

  1. Tests of cosmic ray radiography for power industry applications

    Science.gov (United States)

    Durham, J. M.; Guardincerri, E.; Morris, C. L.; Bacon, J.; Fabritius, J.; Fellows, S.; Poulson, D.; Plaud-Ramos, K.; Renshaw, J.

    2015-06-01

    In this report, we assess muon multiple scattering tomography as a non-destructive inspection technique in several typical areas of interest to the nuclear power industry, including monitoring concrete degradation, gate valve conditions, and pipe wall thickness. This work is motivated by the need for imaging methods that do not require the licensing, training, and safety controls of x-rays, and by the need to be able to penetrate considerable overburden to examine internal details of components that are otherwise inaccessible, with minimum impact on industrial operations. In some scenarios, we find that muon tomography may be an attractive alternative to more typical measurements.

  2. Tests of cosmic ray radiography for power industry applications

    Directory of Open Access Journals (Sweden)

    J. M. Durham

    2015-06-01

    Full Text Available In this report, we assess muon multiple scattering tomography as a non-destructive inspection technique in several typical areas of interest to the nuclear power industry, including monitoring concrete degradation, gate valve conditions, and pipe wall thickness. This work is motivated by the need for imaging methods that do not require the licensing, training, and safety controls of x-rays, and by the need to be able to penetrate considerable overburden to examine internal details of components that are otherwise inaccessible, with minimum impact on industrial operations. In some scenarios, we find that muon tomography may be an attractive alternative to more typical measurements.

  3. Tests of cosmic ray radiography for power industry applications

    Energy Technology Data Exchange (ETDEWEB)

    Durham, J. M., E-mail: durham@lanl.gov; Guardincerri, E.; Morris, C. L.; Bacon, J.; Fabritius, J.; Fellows, S.; Poulson, D.; Plaud-Ramos, K. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Renshaw, J. [Electric Power Research Institute, Charlotte, NC 28262-8550 (United States)

    2015-06-15

    In this report, we assess muon multiple scattering tomography as a non-destructive inspection technique in several typical areas of interest to the nuclear power industry, including monitoring concrete degradation, gate valve conditions, and pipe wall thickness. This work is motivated by the need for imaging methods that do not require the licensing, training, and safety controls of x-rays, and by the need to be able to penetrate considerable overburden to examine internal details of components that are otherwise inaccessible, with minimum impact on industrial operations. In some scenarios, we find that muon tomography may be an attractive alternative to more typical measurements.

  4. Power-hardware-in-the-loop test of VSC-HVDC connection for off-shore wind power plants

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Ranjan [Siemens Wind Power A/S, Brande (Denmark); Technical Univ. of Denmark (Denmark). Center for Electric Technology; Cha, Seung T.; Wu, Qiuwei; Rasmussen, Tonny W.; Oestergaard, Jacob [Technical Univ. of Denmark (Denmark). Center for Electric Technology; Jensen, Kim H. [Siemens Wind Power A/S, Brande (Denmark)

    2011-07-01

    This paper present a power-hardware-in-the-loop (PHIL) test of an off-shore wind power plant (WPP) interconnected to the on-shore grid via a VSC-HVDC connection. The intention of the PHIL test is to verify the hardware interaction and the control co-ordination between the plant side VSC of the HVDC system and the wind turbines within the WPP in order to ensure smooth operation of the WPP under both normal and fault operating condition. The PHIL test platform is comprised of a real time digital simulator (RTDS), a Spitzenberger Spies three phase 7,5 kW power amplifier, a purpose built VSC and a DC chopper. The WPP is simulated in the RTDS as a single full-scale wind turbine. The simulated WPP interacts with the WPP side VSC through the power amplifier. The interface between the RTDS and the power amplifier is done via an analogue GTAO I/O card of the RTDS and the input channel of the amplifier. The amplifier scales up the voltages at the point of connection of the WPP in the RTDS to the voltage level for the WPP side VSC. The WPP side VSC converter is equipped with a DC chopper. The test results show the successful control coordination between the WPP and the plant side VSC converter of the HVDC connection of the WPP. (orig.)

  5. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  6. The approach to analysis of significance of flaws in ASME section III and section XI

    International Nuclear Information System (INIS)

    Cowan, A.

    1979-01-01

    ASME III Appendix G and ASME XI Appendix A describe linear elastic fracture mechanics methods to assess the significance of defects in thick-walled pressure vessels for nuclear reactor systems. The assessment of fracture toughness, Ksub(Ic), is based upon recommendations made by a Task Group of the USA Pressure Vessel Research Committee and is dependent upon correlations with drop weight and Charpy V-notch data to give a lower bound of fracture toughness Ksub(IR). The methods used in the ASME Appendices are outlined noting that, whereas ASME III Appendix G defines a procedure for obtaining allowable pressure vessel loadings for normal service in the presence of a defect, ASME XI Appendix A defines methods for assessing the significance of defects (found by volumetric inspection) under normal and emergency and faulted conditions. The methods of analysis are discussed with respect to material properties, flaw characterisation, stress analysis and recommended safety factors; a short discussion is given on the applicability of the data and methods to other materials and non-nuclear structures. (author)

  7. Reliability test and failure analysis of high power LED packages

    International Nuclear Information System (INIS)

    Chen Zhaohui; Zhang Qin; Wang Kai; Luo Xiaobing; Liu Sheng

    2011-01-01

    A new type application specific light emitting diode (LED) package (ASLP) with freeform polycarbonate lens for street lighting is developed, whose manufacturing processes are compatible with a typical LED packaging process. The reliability test methods and failure criterions from different vendors are reviewed and compared. It is found that test methods and failure criterions are quite different. The rapid reliability assessment standards are urgently needed for the LED industry. 85 0 C/85 RH with 700 mA is used to test our LED modules with three other vendors for 1000 h, showing no visible degradation in optical performance for our modules, with two other vendors showing significant degradation. Some failure analysis methods such as C-SAM, Nano X-ray CT and optical microscope are used for LED packages. Some failure mechanisms such as delaminations and cracks are detected in the LED packages after the accelerated reliability testing. The finite element simulation method is helpful for the failure analysis and design of the reliability of the LED packaging. One example is used to show one currently used module in industry is vulnerable and may not easily pass the harsh thermal cycle testing. (semiconductor devices)

  8. A Powerful Test for Comparing Multiple Regression Functions.

    Science.gov (United States)

    Maity, Arnab

    2012-09-01

    In this article, we address the important problem of comparison of two or more population regression functions. Recently, Pardo-Fernández, Van Keilegom and González-Manteiga (2007) developed test statistics for simple nonparametric regression models: Y(ij) = θ(j)(Z(ij)) + σ(j)(Z(ij))∊(ij), based on empirical distributions of the errors in each population j = 1, … , J. In this paper, we propose a test for equality of the θ(j)(·) based on the concept of generalized likelihood ratio type statistics. We also generalize our test for other nonparametric regression setups, e.g, nonparametric logistic regression, where the loglikelihood for population j is any general smooth function [Formula: see text]. We describe a resampling procedure to obtain the critical values of the test. In addition, we present a simulation study to evaluate the performance of the proposed test and compare our results to those in Pardo-Fernández et al. (2007).

  9. Applying reliability centered maintenance analysis principles to inservice testing

    International Nuclear Information System (INIS)

    Flude, J.W.

    1994-01-01

    Federal regulations require nuclear power plants to use inservice test (IST) programs to ensure the operability of safety-related equipment. IST programs are based on American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements. Many of these plants also use Reliability Centered Maintenance (RCM) to optimize system maintenance. ASME Code requirements are hard to change. The process for requesting authority to use an alternate strategy is long and expensive. The difficulties of obtaining this authority make the use of RCM method on safety-related systems not cost effective. An ASME research task force on Risk Based Inservice Testing is investigating changing the Code. The change will allow plants to apply RCM methods to the problem of maintenance strategy selection for safety-related systems. The research task force is working closely with the Codes and Standards sections to develop a process related to the RCM process. Some day plants will be able to use this process to develop more efficient and safer maintenance strategies

  10. TESTING FOR PURCHASING POWER PARITY FOR SELECTED CISCOUNTRIES

    Directory of Open Access Journals (Sweden)

    Arş. Gör. Esra Ballı

    2016-07-01

    Full Text Available This study examines the validity of PPP for Commonwealth of Independent States(CIS namely Armenia, Azerbaijan, Belarus, Kazakhstan, Kyrgyzstan, Russia andUkraine for the period 1995M1-2015M11. In this study we perform stationary teston twobases. In the first place, we initiate conventional unit root test such asADF. Secondly, we utilize unit root tests allowing for structural break. Given thespan and characteristics of the period which involves a significant break such as1998 Russian economic crisis, we conclude that PPP holds for Armenia, Belarus,Kazakhstan and Kyrgyzstan, while it does not hold for Azerbaijan, Russia andUkraine.

  11. Modified Anderson-Darling tests with selective power improvement

    OpenAIRE

    Cabaña Nigro, Alejandra

    2001-01-01

    Transformed empirical processes (TEPs) have been used by the authors in a previous paper to construct consistent and selectively efficient goodness-of-fit tests of the Kolmogorov-Smirnov type. A straightforward application of the same ideas to the construction of tests of the Cramér-von-Mises type with the same properties leads to cumbersome computations. This short note of the incovenients encountered, and introduces a new family of quadratic statistics of the Cramér-von Mises type, in order...

  12. Testing of power frequency magnetic and electric fieldmeters

    International Nuclear Information System (INIS)

    Brueggemeyer, H.

    1994-01-01

    A lot of epidemiological studies and experiments with cells and animals were made in the last decade to find out, whether electromagnetic fields, with fieldstrength as they occur in our normal environment, can influence biological systems. To make the results traceable, it is necessary first to look at the quality of the used fieldmeters. For this reason 16 field meters used in Germany were tested. The test was made by using calibration setups, the field near transmission lines and the field around common appliances. (orig.) [de

  13. Statistical Tests Black swans or dragon-kings? A simple test for deviations from the power law★

    Science.gov (United States)

    Janczura, J.; Weron, R.

    2012-05-01

    We develop a simple test for deviations from power law tails. Actually, from the tails of any distribution. We use this test - which is based on the asymptotic properties of the empirical distribution function - to answer the question whether great natural disasters, financial crashes or electricity price spikes should be classified as dragon-kings or `only' as black swans.

  14. JT-60 power tests from mechanical and thermal viewpoints of tokamak machine

    International Nuclear Information System (INIS)

    Takatsu, H.; Yamamoto, M.; Ohkubo, M.

    1986-01-01

    JT-60 power tests were carried out, to demonstrate, in advance of actual plasma operation, satisfactory performance of the tokamak machine, power suppliers and control system in combination. The tests began with low power ones of individual coil systems, progressed to full power ones and concluded successfully. The present paper describes the principal results of JT-60 power tests from mechanical and thermal viewpoints of tokamak machine. All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and mechanical integrity of tokamak machine. Measured strain and displacement showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. Vibration of the vacuum vessel was found to be large up to 48 m/s/sup 2/ and caused excessive vibration of the lateral port gate-valves. A few limitations to machine operation were also made clear quantatively

  15. A nuclear power plant certification test plan and checklist

    International Nuclear Information System (INIS)

    Halverson, S.M.

    1989-01-01

    Regulations within the nuclear industry are requiring that all reference plant simulators be certified prior to or during 1991. A certification test plan is essential to ensure that this goal is met. A description of each step in the certification process is provided in this paper, along with a checklist to help ensure completion of each item

  16. Scalable Power-Component Models for Concept Testing

    Science.gov (United States)

    2011-08-17

    motor speed can be either positive or negative dependent upon the propelling or regenerative braking scenario. The simulation provides three...the machine during generation or regenerative braking . To use the model, the user modifies the motor model criteria parameters by double-clicking... SYSTEMS ENGINEERING AND TECHNOLOGY SYMPOSIUM MODELING & SIMULATION, TESTING AND VALIDATION (MSTV) MINI-SYMPOSIUM AUGUST 9-11 DEARBORN, MICHIGAN

  17. Testing in power plant construction as well as in the petrochemical and chemical industry

    International Nuclear Information System (INIS)

    Riess, N.; Schittko, H.

    1978-01-01

    In general, the upgrading of requirements for the most different fields of engineering is also characterized by a corresponding effort in testing. In this context especially nondestructive tests of materials are of outstanding importance. In the fields of power plant construction (among others, components for nuclear power plants) as well as petrochemical and chemical industry considered here, almost all nondestructive test methods are applied. This paper discusses not so much theoretical testing problems, but rather test objects as well as specifications and testing equipment. (orig./HP) [de

  18. Power of tests for comparing trend curves with application to national immunization survey (NIS).

    Science.gov (United States)

    Zhao, Zhen

    2011-02-28

    To develop statistical tests for comparing trend curves of study outcomes between two socio-demographic strata across consecutive time points, and compare statistical power of the proposed tests under different trend curves data, three statistical tests were proposed. For large sample size with independent normal assumption among strata and across consecutive time points, the Z and Chi-square test statistics were developed, which are functions of outcome estimates and the standard errors at each of the study time points for the two strata. For small sample size with independent normal assumption, the F-test statistic was generated, which is a function of sample size of the two strata and estimated parameters across study period. If two trend curves are approximately parallel, the power of Z-test is consistently higher than that of both Chi-square and F-test. If two trend curves cross at low interaction, the power of Z-test is higher than or equal to the power of both Chi-square and F-test; however, at high interaction, the powers of Chi-square and F-test are higher than that of Z-test. The measurement of interaction of two trend curves was defined. These tests were applied to the comparison of trend curves of vaccination coverage estimates of standard vaccine series with National Immunization Survey (NIS) 2000-2007 data. Copyright © 2011 John Wiley & Sons, Ltd.

  19. Roles of the Outer Membrane Protein AsmA of Salmonella enterica in the Control of marRAB Expression and Invasion of Epithelial Cells▿

    OpenAIRE

    Prieto, Ana I.; Hernández, Sara B.; Cota, Ignacio; Pucciarelli, M. Graciela; Orlov, Yuri; Ramos-Morales, Francisco; García-del Portillo, Francisco; Casadesús, Josep

    2009-01-01

    A genetic screen for suppressors of bile sensitivity in DNA adenine methylase (dam) mutants of Salmonella enterica serovar Typhimurium yielded insertions in an uncharacterized locus homologous to the Escherichia coli asmA gene. Disruption of asmA suppressed bile sensitivity also in phoP and wec mutants of S. enterica and increased the MIC of sodium deoxycholate for the parental strain ATCC 14028. Increased levels of marA mRNA were found in asmA, asmA dam, asmA phoP, and asmA wec strains of S....

  20. Power supply for the Spanish stellarator TJ-II, design, construction, and tests

    Energy Technology Data Exchange (ETDEWEB)

    Perez, A.; Lucia, C.; Alberdi, B.; Del Rio, J.M. [JEMA SA, Lasarte-Oria (Spain); Almoguera, L.; Blaumoser, M.; Kirpitchev, I.; Mendez, P. [Asociacion EURATOM-CIEMAT para Fusion, Madrid (Spain)

    1995-12-31

    Most of the components of the electrical power supply system of the new TJ-II stellarator, which is under construction in Madrid (Spain), are now constructed and tested. The flywheel synchronous generator is still under construction and its tests are planned for the end of 1995. The power plant is described in detail as well as the tests which have been carried out and their results.

  1. Evaluation of load rejection to house load test at 50% power for UCN 3

    International Nuclear Information System (INIS)

    Lee, Chang Gyun; Sohn, Suk Whun; Sohn, Jong Joo; Seo, Jong Tae; Lee, Sang Keun; Kim, Yong Sung; Nam, Kyu Won; Jung, Yang Mook; Chae, Kyeong Sik; Koh, Bum Jae; Oh, Chul Sung; Park, Hee Chool

    1998-01-01

    The Load Rejection to House Load test at 50% power was successfully performed during the UCN 3 PAT period. In this test, all plant control systems automatically controlled the plant from 50% power to house load operation mode. The KISPAC code, which was used in the performance analysis during the design process of UCN 3 and 4, predictions of the test agreed with the measured data demonstrating the validity of the code as well as completeness of the plant design

  2. Cyclin D1 in ASM Cells from Asthmatics Is Insensitive to Corticosteroid Inhibition.

    Science.gov (United States)

    Allen, Jodi C; Seidel, Petra; Schlosser, Tobias; Ramsay, Emma E; Ge, Qi; Ammit, Alaina J

    2012-01-01

    Hyperplasia of airway smooth muscle (ASM) is a feature of the remodelled airway in asthmatics. We examined the antiproliferative effectiveness of the corticosteroid dexamethasone on expression of the key regulator of G(1) cell cycle progression-cyclin D1-in ASM cells from nonasthmatics and asthmatics stimulated with the mitogen platelet-derived growth factor BB. While cyclin D1 mRNA and protein expression were repressed in cells from nonasthmatics in contrast, cyclin D1 expression in asthmatics was resistant to inhibition by dexamethasone. This was independent of a repressive effect on glucocorticoid receptor translocation. Our results corroborate evidence demonstrating that corticosteroids inhibit mitogen-induced proliferation only in ASM cells from subjects without asthma and suggest that there are corticosteroid-insensitive proliferative pathways in asthmatics.

  3. Corticosteroids reduce IL-6 in ASM cells via up-regulation of MKP-1.

    Science.gov (United States)

    Quante, Timo; Ng, Yee Ching; Ramsay, Emma E; Henness, Sheridan; Allen, Jodi C; Parmentier, Johannes; Ge, Qi; Ammit, Alaina J

    2008-08-01

    The mechanisms by which corticosteroids reduce airway inflammation are not completely understood. Traditionally, corticosteroids were thought to inhibit cytokines exclusively at the transcriptional level. Our recent evidence, obtained in airway smooth muscle (ASM), no longer supports this view. We have found that corticosteroids do not act at the transcriptional level to reduce TNF-alpha-induced IL-6 gene expression. Rather, corticosteroids inhibit TNF-alpha-induced IL-6 secretion by reducing the stability of the IL-6 mRNA transcript. TNF-alpha-induced IL-6 mRNA decays at a significantly faster rate in ASM cells pretreated with the corticosteroid dexamethasone (t(1/2) = 2.4 h), compared to vehicle (t(1/2) = 9.0 h; P ASM cells.

  4. Basic requirements of mechanical properties for nuclear pressure vessel materials in ASME-BPV code

    International Nuclear Information System (INIS)

    Ning Dong; Yao Weida

    2011-01-01

    The four basic aspects of strengths, ductility, toughness and fatigue strengths can be summarized for overall mechanical properties requirements of materials for nuclear pressure-retaining vessels in ASME-BPV code. These mechanical property indexes involve in the factors of melting, manufacture, delivery conditions, check or recheck for mechanical properties and chemical compositions, etc. and relate to degradation and damage accumulation during the use of materials. This paper specifically accounts for the basic requirements and theoretic basis of mechanical properties for nuclear pressure vessel materials in ASME-BPV code and states the internal mutual relationships among the four aspects of mechanical properties. This paper focuses on putting forward at several problems on mechanical properties of materials that shall be concerned about during design and manufacture for nuclear pressure vessels according to ASME-BPV code. (author)

  5. Metabolic Activity Interferometer: A Powerful Tool for Testing Antibiotics

    Directory of Open Access Journals (Sweden)

    Rachel R. P. Machado

    2012-01-01

    Full Text Available It is demonstrated that the efficiency of antibiotics can be tested using an interferometric method. Two antibiotics were used as models to show that an interferometric method to monitor the metabolic activity of slowly growing bacteria can be a safer method to judge antimicrobial properties of substances than conventional methods. The susceptibility of Mycobacterium bovis to hexane extract of Pterodon emarginatus and to the well-known antibiotic rifampicin was tested with the interferometric method and with the conventional microplate method. The microplate method revealed a potential activity of hexane extract against M. bovis. However, the interferometric method showed that the action of this substance is rather limited. Also in the case of rifampicin, the interferometric method was able to detect resistant bacteria.

  6. Experience with RTD response time testing in nuclear power plants

    International Nuclear Information System (INIS)

    Hashemian, H.M.; Kerlin, T.W.

    1985-01-01

    The reactor coolant temperatures in pressurized water reactors are measured with platinum resistance temperature detectors (RTDs). The information furnished by these RTDs is used for plant protection as well as control. As a part of the plant protection system, the RTDs must respond to temperature changes in a timely fashion. The RTD response time requirements are different for the various plant types. These requirements are specified in the plant technical specifications in terms of an RTd time constant. The current time constant requirements for nuclear plant RTDs varies from 0.5 seconds to 13.0 seconds depending on the type of the plant. Therefore, different types of RTDs are used in different plants to achieve the required time constants. In addition, in-situ response time tests are periodically performed on protective system RTDs to ensure that the in-service time constants are within acceptable limits as the plant is operating. The periodic testing is important because response time degradation may occur while the RTD ages in the process. Recent response time tests in operating plants revealed unacceptable time constants for several protection system RTDs. As a result, these plants had to be shut down to resolve the problem which in one case was due to improper installation and in another case was because of degradation of a thermal compound used in the thermowell

  7. High-power circulator test results at 350 and 700 MHz

    International Nuclear Information System (INIS)

    Roybal, W.; Bradley, J.T.; Rees, D.E.

    2000-01-01

    The high-power RF systems for the Accelerator Production of Tritium (APT) program require high-power circulators at 350 MHz and 700 MHz to protect 1 MW Continuous Wave (CW) klystrons from reflected power. The 350 MHz circulator is based on the CERN, EXF, and APS designs and has performed very well. The 700 MHz circulator is a new design. Prototype 700 MHz circulators have been high-power tested at Los Alamos National Laboratory (LANL). The first of these circulators has satisfied performance requirements. The circulator requirements, results from the testing, and lessons learned from this development are presented and discussed

  8. Tests of the power PC theory of causal induction with negative contingencies.

    Science.gov (United States)

    Shanks, David R

    2002-01-01

    The power PC theory of causal induction (Cheng, 1997) proposes that causal estimates are based on the power p of a potential cause, where p is the contingency between the cause and effect normalized by the base rate of the effect. Previous tests of this theory have concentrated on generative causes that have positive contingencies with their associated outcomes. Here we empirically test this theory in two experiments using preventive causes that have negative contingencies for their outcomes. Contrary to the power PC theory, the results show that causal judgments vary with contingency across conditions of constant power p. This pattern is consistent, however, with several alternative accounts of causal judgment.

  9. Advanced power cycler with intelligent monitoring strategy of IGBT module under test

    DEFF Research Database (Denmark)

    Choi, U. M.; Blaabjerg, F.; Iannuzzo, F.

    2017-01-01

    and diode, which for the wear-out condition monitoring are presented. This advanced power cycler allows to perform power cycling test cost-effectively under conditions close to real power converter applications. In addition, an intelligent monitoring strategy for the separation of package-related wear......-out failure mechanisms has been proposed. By means of the proposed method, the wear-out failure mechanisms of an IGBT module can be separated without any additional efforts during the power cycling tests. The validity and effectiveness of the proposed monitoring strategy are also verified by experiments....

  10. Design of Jet lower hybrid current drive generator and operation of high power test bed

    International Nuclear Information System (INIS)

    Dobbing, J.A.; Bosia, G.; Brandon, M.; Gammelin, M.; Gormezano, C.; Jacquinot, J.; Jessop, G.; Lennholm, M.; Pain, M.; Sibley, A.

    1989-01-01

    The JET Lower Hybrid Current Drive (LHCD) generator consists of 24 klystrons each rated for 650 KW operating at 3.7 GHz, giving a nominal generator power of 15.6 MW for 10 seconds or 12 MW for 20 seconds. This power will be transmitted through 24 waveguides to a phased array launcher on one of the main ports of the JET machine. In addition, two klystrons are currently being operated on a high power test bed to establish reliable operation of the generators components and test high power microwave components prior to their installation

  11. Grid Integration of PV Power based on PHIL testing using different Interface Algorithms

    DEFF Research Database (Denmark)

    Craciun, Bogdan-Ionut; Kerekes, Tamas; Sera, Dezso

    2013-01-01

    to be more active in grid support. Therefore, a better understanding and detailed analysis of the PV systems interaction with the grid is needed; hence power hardware in the loop (PHIL) testing involving PV power became an interesting subject to look into. To test PV systems for grid code (GC) compliance......Photovoltaic (PV) power among all renewable energies had the most accelerated growth rate in terms of installed capacity in recent years. Transmission System Operators (TSOs) changed their perspective about PV power and started to include it into their planning and operation, imposing PV systems...

  12. The ASM Curriculum Guidelines for Undergraduate Microbiology: A Case Study of the Advocacy Role of Societies in Reform Efforts

    Directory of Open Access Journals (Sweden)

    Rachel E.A. Horak

    2015-03-01

    Full Text Available A number of national reports, including Vision and Change in Undergraduate Biology Education: A Call to Action, have called for drastic changes in how undergraduate biology is taught. To that end, the American Society for Microbiology (ASM developed new Curriculum Guidelines for undergraduate microbiology that outline a comprehensive curriculum for any undergraduate introductory microbiology course or program of study. Designed to foster enduring understanding of core microbiology concepts, the Guidelines work synergistically with backwards course design to focus teaching on student-centered goals and priorities.  In order to qualitatively assess how the ASM Curriculum Guidelines are used by educators and learn more about the needs of microbiology educators, the ASM Education Board distributed two surveys to the ASM education community. In this report, we discuss results of these surveys (353 responses. We found that the ASM Curriculum Guidelines are being implemented in many different types of courses at all undergraduate levels. Educators indicated that the ASM Curriculum Guidelines were very helpful when planning courses and assessments. We discuss some specific ways in which the ASM Curriculum Guidelines have been used in undergraduate classrooms. The survey identified some barriers that microbiology educators faced when trying to adopt the ASM Curriculum Guidelines, including lack of time, lack of financial resources, and lack of supporting resources. Given the self-reported challenges to implementing the ASM Curriculum Guidelines in undergraduate classrooms, we identify here some activities related to the ASM Curriculum Guidelines that the ASM Education Board has initiated to assist educators in the implementation process.

  13. Development and Testing of a Prototype Grid-Tied Photovoltaic Power System

    Science.gov (United States)

    Eichenberg, Dennis J.

    2009-01-01

    The NASA Glenn Research Center (GRC) has developed and tested a prototype 2 kW DC grid-tied photovoltaic (PV) power system at the Center. The PV system has generated in excess of 6700 kWh since operation commenced in July 2006. The PV system is providing power to the GRC grid for use by all. Operation of the prototype PV system has been completely trouble free. A grid-tied PV power system is connected directly to the utility distribution grid. Facility power can be obtained from the utility system as normal. The PV system is synchronized with the utility system to provide power for the facility, and excess power is provided to the utility. The project transfers space technology to terrestrial use via nontraditional partners. GRC personnel glean valuable experience with PV power systems that are directly applicable to various space power systems, and provide valuable space program test data. PV power systems help to reduce harmful emissions and reduce the Nation s dependence on fossil fuels. Power generated by the PV system reduces the GRC utility demand, and the surplus power aids the community. Present global energy concerns reinforce the need for the development of alternative energy systems. Modern PV panels are readily available, reliable, efficient, and economical with a life expectancy of at least 25 years. Modern electronics has been the enabling technology behind grid-tied power systems, making them safe, reliable, efficient, and economical with a life expectancy of at least 25 years. Based upon the success of the prototype PV system, additional PV power system expansion at GRC is under consideration. The prototype grid-tied PV power system was successfully designed and developed which served to validate the basic principles described, and the theoretical work that was performed. The report concludes that grid-tied photovoltaic power systems are reliable, maintenance free, long life power systems, and are of significant value to NASA and the community.

  14. Testing of a Stirling engine for heat + power cogeneration; Test eines Stirlingmotors zur Kraft-Waerme-Kopplung

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, M.; Heinen, J. [RWE Energy AG, Essen (Germany)

    2007-01-15

    As part of a technology evaluation of distributed energy generators, RWE Energy AG extensively tested a micro combined heat and power appliance, powered by a Stirling engine developed by the British firm Microgen Energy Limited. Microgen Energy Limited is a specialist in micro combined heat and power (microCHP) based on unique Free-Piston Stirling generator technology Microgen is working with leading appliance manufacturers to integrate its core technology into a range of innovative microCHP products. The investigations concentrated on the determination of capacity, efficiency and emissions, the grid connection and behaviour at start-up and under varying loads. This article summarises the results of the tests and gives an overview of micro-CHP technologies (CHP=combined heat and power) and their possible significance to the market in the future. (orig.)

  15. Comparison between Standing Broad Jump test and Wingate test for assessing lower limb anaerobic power in elite sportsmen.

    Science.gov (United States)

    Krishnan, Anup; Sharma, Deep; Bhatt, Madhu; Dixit, Apoorv; Pradeep, P

    2017-04-01

    Lower limb explosive power is an important motor quality for sporting performance and indicates use of anaerobic energy systems like stored ATP and Creatine phosphate system. Weightlifting, Fencing and Wrestling use it for monitoring and identification of potential sportsmen. The Wingate test and Standing Broad Jump (SBJ) test are reliable and accurate tests for its assessment. This study conducted on elite Indian sportsmen tries to analyse feasibility of use of the SBJ test in sports and military medicine when Wingate test is impractical. 95 elite sportsmen (51 Fencers, 17 Weight lifters and 27 Wrestlers) of a sports institute were administered Wingate cycle ergometer test and SBJ under standardised conditions. The results were analysed for mass and inter-discipline correlation. Analysis using Pearson's correlation showed significant positive correlation between Peak power ( r  = 0.446, p  Weightlifters. Bland-Altman plot analysis showed that about 94% pairs of peak power and SBJ were within limits of agreement for each discipline as well as among all sportsmen. The test results show definite correlation and SBJ test can be used as a field test in performance monitoring, talent identification, military recruit screening and injury prevention.

  16. Low-Power Built-In Self-Test Techniques for Embedded SRAMs

    Directory of Open Access Journals (Sweden)

    Shyue-Kung Lu

    2007-01-01

    Full Text Available The severity of power consumption during parallel BIST of embedded memory cores is growing significantly. In order to alleviate this problem, a row bank-based precharge technique based on the divided wordline (DWL architecture is proposed for low-power testing of embedded SRAMs. The memory cell array is first divided into row banks. The effectiveness of the row bank-based precharge technique is due to the predictable address sequence during test. In low-power test mode, instead of precharging the entire memory array, only the current accessed row bank is precharged. This will result in significant power saving for the precharge circuitry. The precharge power can be reduced to 1/b of that of the traditional precharge techniques, where b denotes the number of row banks in the memory array. With simple transmission gates and inverters, the modified precharge control circuitry was also designed. The hardware overhead for implementing the low-power technique is almost negligible. Moreover, the corresponding BIST design to implement the low-power technique is almost the same as the conventional BIST designs. It is also notable that the inherent low-power characteristics of the DWL architecture can be preserved. According to experimental results, 48.9% power reduction can be achieved for a 256 × 256 bit-oriented SRAM. The memory is divided into 8 row banks. Moreover, if the number of row banks increases, the power saving will also increase.

  17. A Survey of Variable Extragalactic Sources with XTE's All Sky Monitor (ASM)

    Science.gov (United States)

    Jernigan, Garrett

    1998-01-01

    The original goal of the project was the near real-time detection of AGN utilizing the SSC 3 of the ASM on XTE which does a deep integration on one 100 square degree region of the sky. While the SSC never performed sufficiently well to allow the success of this goal, the work on the project has led to the development of a new analysis method for coded aperture systems which has now been applied to ASM data for mapping regions near clusters of galaxies such as the Perseus Cluster and the Coma Cluster. Publications are in preparation that describe both the new method and the results from mapping clusters of galaxies.

  18. ASM LabCap's contributions to disease surveillance and the International Health Regulations (2005).

    Science.gov (United States)

    Specter, Steven; Schuermann, Lily; Hakiruwizera, Celestin; Sow, Mah-Séré Keita

    2010-12-03

    The revised International Health Regulations [IHR(2005)], which requires the Member States of the World Health Organization (WHO) to develop core capacities to detect, assess, report, and respond to public health threats, is bringing new challenges for national and international surveillance systems. As more countries move toward implementation and/or strengthening of their infectious disease surveillance programs, the strengthening of clinical microbiology laboratories becomes increasingly important because they serve as the first line responders to detect new and emerging microbial threats, re-emerging infectious diseases, the spread of antibiotic resistance, and the possibility of bioterrorism. In fact, IHR(2005) Core Capacity #8, "Laboratory", requires that laboratory services be a part of every phase of alert and response.Public health laboratories in many resource-constrained countries require financial and technical assistance to build their capacity. In recognition of this, in 2006, the American Society for Microbiology (ASM) established an International Laboratory Capacity Building Program, LabCap, housed under the ASM International Board. ASM LabCap utilizes ASM's vast resources and its membership's expertise-40,000 microbiologists worldwide-to strengthen clinical and public health laboratory systems in low and low-middle income countries. ASM LabCap's program activities align with HR(2005) by building the capability of resource-constrained countries to develop quality-assured, laboratory-based information which is critical to disease surveillance and the rapid detection of disease outbreaks, whether they stem from natural, deliberate or accidental causes.ASM LabCap helps build laboratory capacity under a cooperative agreement with the U.S. Centers for Disease Control and Prevention (CDC) and under a sub-contract with the Program for Appropriate Technology in Health (PATH) funded by the United States Agency for International Development (USAID

  19. ASME and RCC-MR comparison for the prevention of fatigue analysis

    International Nuclear Information System (INIS)

    Autrusson, B.; Acker, D.

    1989-01-01

    The purpose of this survey is to compare the simplified methods, without reference to the safety factor allowed for the mechanical properties. An application of both codes, RCC-MR and ASME, on the design of the wall mock-up of the NET project is made and also an estimation with an elastoplastic analysis. In the case of fatigue analysis according to ASME in the plastic field, the elastic stress is magnified by a K e factor derived from stress variation, S n , disregarding geometrical discontinuities. According to RCC-MR, the elastic maximum strain will magnified by two coefficients accounting for plasticity and variation of Poisson ratio

  20. Intersection tests for single marker QTL analysis can be more powerful than two marker QTL analysis

    Directory of Open Access Journals (Sweden)

    Doerge RW

    2003-06-01

    Full Text Available Abstract Background It has been reported in the quantitative trait locus (QTL literature that when testing for QTL location and effect, the statistical power supporting methodologies based on two markers and their estimated genetic map is higher than for the genetic map independent methodologies known as single marker analyses. Close examination of these reports reveals that the two marker approaches are more powerful than single marker analyses only in certain cases. Simulation studies are a commonly used tool to determine the behavior of test statistics under known conditions. We conducted a simulation study to assess the general behavior of an intersection test and a two marker test under a variety of conditions. The study was designed to reveal whether two marker tests are always more powerful than intersection tests, or whether there are cases when an intersection test may outperform the two marker approach. We present a reanalysis of a data set from a QTL study of ovariole number in Drosophila melanogaster. Results Our simulation study results show that there are situations where the single marker intersection test equals or outperforms the two marker test. The intersection test and the two marker test identify overlapping regions in the reanalysis of the Drosophila melanogaster data. The region identified is consistent with a regression based interval mapping analysis. Conclusion We find that the intersection test is appropriate for analysis of QTL data. This approach has the advantage of simplicity and for certain situations supplies equivalent or more powerful results than a comparable two marker test.

  1. Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

    Science.gov (United States)

    Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

    2013-01-01

    Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

  2. A Main Steam Safety Valve (MSSV) With Fixed Blowdown According to ASME Section III,Part NC-7512

    International Nuclear Information System (INIS)

    Follmer, Bernhard; Schnettler, Armin

    2002-01-01

    In 1986, the NRC issued the Information Notice (IN) 86-05 'Main Steam Safety Valve test failures and ring setting adjustments'. Shortly after this IN was issued, the Code was revised to require that a full flow test has to be performed on each CL.2 MSSV by the manufacturer to verify that the valve was adjusted so that it would reach full lift and thus full relieving capacity and would re-close at a pressure as specified in the valve Design Specification. In response to the concern discussed in the IN, the Westinghouse Owners Group (WOG) performed extensive full flow testing on PWR MSSVs and found that each valve required a unique setting of a combination of two rings in order to achieve full lift at accumulation of 3% and re-closing at a blowdown of 5%. The Bopp and Reuther MSSV type SiZ 2507 has a 'fixed blowdown' i.e. without any adjusting rings to adjust the 'blowdown' so that the blowdown is 'fixed'. More than 1000 pieces of this type are successfully in nuclear power plants in operation. Many of them since about 25 years. Therefore it can be considered as a proven design. It is new that an optimization of this MSSV type SiZ 2507 fulfill the requirements of part NC-7512 of the ASME Section III although there are still no adjusting rings in the flow part. In 2000, for the Qinshan Candu unit 1 and 2 full flow tests were performed with 32 MSSV type SiZ 2507 size 8'' x 12'' at 51 bar saturated steam in only 6 days. In all tests the functional performance was very stable. It was demonstrated by recording the signals lift and system pressure that all valves had acceptable results to achieve full lift at accumulation of 3% and to re-close at blowdown of 5%. This is an advantage which gives a reduction in cost for flow tests and which gives more reliability after maintenance work during outage compared to the common MSSV design with an individual required setting of the combination of the two rings. The design of the type SiZ 2507 without any adjusting rings in the

  3. Japanese programme on the development of high duty fuel and related power ramping tests

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, Y

    1983-06-01

    Power ramping tests hitherto planned or carried out in Japan can be classified into two categories: 1) the tests programme by private organizations on fuel behaviour under various conditions of power ramping, in participating international programmes; and 2) a partially government sponsored programme, which was officially inaugurated in 1981 under the title of High-Duty Fuel Development Programme. The latter has been carried out by the Nuclear Power Engineering Test Centre, based on the schedule decided by the MITI Committee (chaired by the author), for a period of 10 years. These programmes will be described with emphasis on the latter (National Programme). (author)

  4. Japanese programme on the development of high duty fuel and related power ramping tests

    International Nuclear Information System (INIS)

    Mishima, Y.

    1983-01-01

    Power ramping tests hitherto planned or carried out in Japan can be classified into two categories: 1) the tests programme by private organizations on fuel behaviour under various conditions of power ramping, in participating international programmes; and 2) a partially government sponsored programme, which was officially inaugurated in 1981 under the title of High-Duty Fuel Development Programme. The latter has been carried out by the Nuclear Power Engineering Test Centre, based on the schedule decided by the MITI Committee (chaired by the author), for a period of 10 years. These programmes will be described with emphasis on the latter (National Programme). (author)

  5. Testing Procedures and Results of the Prototype Fundamental Power Coupler for the Spallation Neutron Source

    International Nuclear Information System (INIS)

    M. Stirbet; I.E. Campisi; E.F. Daly; G.K. Davis; M. Drury; P. Kneisel; G. Myneni; T. Powers; W.J. Schneider; K.M. Wilson; Y. Kang; K.A. Cummings; T. Hardek

    2001-01-01

    High-power RF testing with peak power in excess of 500 kW has been performed on prototype Fundamental Power Couplers (FPC) for the Spallation Neutron Source superconducting (SNS) cavities. The testing followed the development of procedures for cleaning, assembling and preparing the FPC for installation in the test stand. The qualification of the couplers has occurred for the time being only in a limited set of conditions (travelling wave, 20 pps) as the available RF system and control instrumentation are under improvement

  6. Summary revaluation of cold testing of the first block of nuclear power plant Mochovce

    International Nuclear Information System (INIS)

    Miskolci, M.; Sarvaic, I.

    1998-01-01

    The document contents summary revaluation of the stage of cold testing of the first unit of nuclear power plant Mochovce. The valuation is processed in individual systems with safety significance. The process and results of system testing and their conclusions for the block readiness for active testing are summarized in the document. The valuation has been elaborated by a scientific management for start-up of nuclear power plant Mochovce as an independent conductor assistance for activation check from the nuclear safety point of view. The valuation of the activation results of systems in the first unit of nuclear power plant was processed as of 15.3.1998

  7. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  8. Survey of the power ramp performance testing of KWU'S PWR UO 2, fuel

    Science.gov (United States)

    Ga¨rtner, M.; Fischer, G.

    1987-06-01

    To determine the power ramp performance of KWU's PWR UO 2 fuel, 134 fuel rodlets with burnups of up to 46 GWd/ t (U) and several fuel assemblies with 19 to 30 GWd/t (U) burnup were ramped in power in the research reactors HFR Petten/The Netherlands and R2 Studsvik/Sweden and in the power plants KWO and KWB-A/Germany, respectively. The power ramp tests demonstrate decreasing resistance of the PWR fuel rods to PCI (pellet-to-clad interaction) up to fuel burnups of 35 GWd/t (U) and a reversal effect at higher burnups. The fuel rods can be operated free of defects at fast power transients to linear heat generation rates of up to 400 W/cm, at least.Power levels of up to 490 W/cm can be reached without defects by reducing the ramp rate. After reshuffling according to an out-in scheme, 1-cycle fuel assemblies may return to rod powers of up to 480 W/cm with a power increase rate of up to 10 W/(cm min) without fuel rod damage. Set points basing on these test results and incorporated into the power distribution control and power density limitation system of KWU's advanced power plants guarantee safe plant operation under normal and load follow operating conditions.

  9. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  10. Program plan for the Brayton Isotope Power System. Phase I. Design, fabrication and test of the Brayton Isotope Power System

    International Nuclear Information System (INIS)

    1975-01-01

    Phase I of an overall program for the development of a 500 to 2000 W(e) (EOM), 7-y life, power system for space vehicles is discussed. The system uses a closed Brayton dynamic system to convert energy from an isotope heat source at a net efficiency greater than 25 percent. This first phase, a 35-month effort, is for the conceptual design of a 1300 W(e), 450 lb flight system and the design, fabrication, and test of a ground demonstration system. The flight system will use, for the baseline design, two of the multihundred-watt (MHW) heat sources being developed. The Ground Demonstration System will simulate, as closely as possible, the Brayton Isotope Power Flight System and will utilize components and technology being developed for the Mini-Brayton rotating unit, recuperator and heat source assembly, respectively. The Ground Demonstration System includes a performance test and a 1000-h endurance test

  11. A testing procedure for wind turbine generators based on the power grid statistical model

    DEFF Research Database (Denmark)

    Farajzadehbibalan, Saber; Ramezani, Mohammad Hossein; Nielsen, Peter

    2017-01-01

    In this study, a comprehensive test procedure is developed to test wind turbine generators with a hardware-in-loop setup. The procedure employs the statistical model of the power grid considering the restrictions of the test facility and system dynamics. Given the model in the latent space...

  12. Power properties of invariant tests for spatial autocorrelation in linear regression

    NARCIS (Netherlands)

    Martellosio, F.

    2006-01-01

    Many popular tests for residual spatial autocorrelation in the context of the linear regression model belong to the class of invariant tests. This paper derives a number of exact properties of the power function of such tests. In particular, we extend the work of Krämer (2005, Journal of Statistical

  13. 40 CFR 86.129-00 - Road load power, test weight, and inertia weight class determination.

    Science.gov (United States)

    2010-07-01

    ... inertia weight class determination. 86.129-00 Section 86.129-00 Protection of Environment ENVIRONMENTAL... power, test weight, and inertia weight class determination. Applicability. Section 86.129-94 (a) applies... testing using paragraphs (e)(1) and (e)(2) of this section. (f)(1) Required test dynamometer inertia...

  14. Modeling a constant power load for nickel-hydrogen battery testing using SPICE

    Science.gov (United States)

    Bearden, Douglas B.; Lollar, Louis F.; Nelms, R. M.

    1990-01-01

    The effort to design and model a constant power load for the HST (Hubble Space Telescope) nickel-hydrogen battery tests is described. The constant power load was designed for three different simulations on the batteries: life cycling, reconditioning, and capacity testing. A dc-dc boost converter was designed to act as this constant power load. A boost converter design was chosen because of the low test battery voltage (4 to 6 VDC) generated and the relatively high power requirement of 60 to 70 W. The SPICE model was shown to consistently predict variations in the actual circuit as various designs were attempted. It is concluded that the confidence established in the SPICE model of the constant power load ensures its extensive utilization in future efforts to improve performance in the actual load circuit.

  15. Structural and functional analysis of the ASM p.Ala359Asp mutant that causes acid sphingomyelinase deficiency.

    Science.gov (United States)

    Acuña, Mariana; Castro-Fernández, Víctor; Latorre, Mauricio; Castro, Juan; Schuchman, Edward H; Guixé, Victoria; González, Mauricio; Zanlungo, Silvana

    2016-10-21

    Niemann-Pick disease (NPD) type A and B are recessive hereditary disorders caused by deficiency in acid sphingomyelinase (ASM). The p.Ala359Asp mutation has been described in several patients but its functional and structural effects in the protein are unknown. In order to characterize this mutation, we modeled the three-dimensional ASM structure using the recent available crystal of the mammalian ASM as a template. We found that the p.Ala359Asp mutation is localized in the hydrophobic core and far from the sphingomyelin binding site. However, energy function calculations using statistical potentials indicate that the mutation causes a decrease in ASM stability. Therefore, we investigated the functional effect of the p.Ala359Asp mutation in ASM expression, secretion, localization and activity in human fibroblasts. We found a 3.8% residual ASM activity compared to the wild-type enzyme, without changes in the other parameters evaluated. These results support the hypothesis that the p.Ala359Asp mutation causes structural alterations in the hydrophobic environment where ASM is located, decreasing its enzymatic activity. A similar effect was observed in other previously described NPDB mutations located outside the active site of the enzyme. This work shows the first full size ASM mutant model describe at date, providing a complete analysis of the structural and functional effects of the p.Ala359Asp mutation over the stability and activity of the enzyme. Copyright © 2016 Elsevier Inc. All rights reserved.

  16. Sex Ratios, Economic Power, and Women's Roles: A Theoretical Extension and Empirical Test.

    Science.gov (United States)

    South, Scott J.

    1988-01-01

    Tested hypotheses concerning sex ratios, women's roles, and economic power with data from 111 countries. Found undersupply of women positively associated with proportion of women who marry and fertility rate; inversely associated with women's average age at marriage, literacy rate, and divorce rate. Suggests women's economic power may counteract…

  17. 77 FR 73056 - Initial Test Programs for Water-Cooled Nuclear Power Plants

    Science.gov (United States)

    2012-12-07

    ... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... (DG), DG-1259, ``Initial Test Programs for Water-Cooled Nuclear Power Plants.'' This guide describes... (ITPs) for light water cooled nuclear power plants. DATES: Submit comments by January 31, 2013. Comments...

  18. The Relationship between Field Tests of Anaerobic Power and 10-km Run Performance.

    Science.gov (United States)

    Sinnett, Aaron M.; Berg, Kris; Latin, Richard W.; Noble, John M.

    2001-01-01

    Investigated the relationship between several field tests of anaerobic power (e.g., +various sprints, vertical jumps, and a plyometric leap) and distance running performance in trained adult male and female runners. Results indicate that anaerobic power is significantly related to distance running performance and may explain a meaningful…

  19. Apparatus and method for heat-run test on high-power PWM ...

    Indian Academy of Sciences (India)

    Due to its bi-directional power flow capability, this is used in a ..... generation, and selection of controller parameters for the voltage and current loops is presented ..... For example, the total power consumed for heat-run test on converters at a.

  20. Risk-informed appendices G and E for section XI of the ASME Boiler and Pressure Vessel Code

    International Nuclear Information System (INIS)

    Carter, B; Spanner, J.; Server, W.; Gamble, R.; Bishop, B.; Palm, N.; Heinecke, C.

    2011-01-01

    Full text of publication follows: The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, contains two appendices (G and E) related to reactor pressure boundary integrity. Appendix G provides procedures for defining Service Level A and B pressure temperature limits for ferritic components in the reactor coolant pressure boundary. Recently, an alternative risk informed methodology has been developed for ASME Section XI, Appendix G. The alternative methodology provides simple procedures to define risk informed pressure temperature limits for Service Level A and B events, including leak testing and reactor start up and shut down for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). Risk informed pressure temperature limits provide more operational flexibility, particularly for reactor pressure vessels (RPV) with relatively high irradiation levels and radiation sensitive materials. Appendix E of Section XI provides a methodology for assessing conditions when the Appendix G limits are exceeded. A similar risk informed methodology is being considered for Appendix E. The probabilistic fracture mechanics evaluations used to develop the risk informed relationships included appropriate material properties for the range of RPV materials in operating plants in the United States and operating history and system operational constraints in both BWRs and PWRs. The analysis results were used to define pressure temperature relationships that provide an acceptable level of risk, consistent with safety goals defined by the U.S. Nuclear Regulatory Commission. The alternative methodologies for Appendices G and E will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low temperature over pressurization for PWRs and BWR leak testing. Overall, application of the risk informed appendices can result in increased plant