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Sample records for asme code section

  1. 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

  2. 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)

  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. 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)

  5. 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

  6. 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)

  7. 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)

  8. 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)

  9. 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...

  10. 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...

  11. 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

  12. 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)

  13. Evaluation of conservatisms and environmental effects in ASME Code, Section III, Class 1 fatigue analysis

    International Nuclear Information System (INIS)

    Deardorff, A.F.; Smith, J.K.

    1994-08-01

    This report documents the results of a study regarding the conservatisms in ASME Code Section 3, Class 1 component fatigue evaluations and the effects of Light Water Reactor (LWR) water environments on fatigue margins. After review of numerous Class 1 stress reports, it is apparent that there is a substantial amount of conservatism present in many existing component fatigue evaluations. With little effort, existing evaluations could be modified to reduce the overall predicted fatigue usage. Areas of conservatism include design transients considerably more severe than those experienced during service, conservative grouping of transients, conservatisms that have been removed in later editions of Section 3, bounding heat transfer and stress analysis, and use of the ''elastic-plastic penalty factor'' (K 3 ). Environmental effects were evaluated for two typical components that experience severe transient thermal cycling during service, based on both design transients and actual plant data. For all reasonable values of actual operating parameters, environmental effects reduced predicted margins, but fatigue usage was still bounded by the ASME Section 3 fatigue design curves. It was concluded that the potential increase in predicted fatigue usage due to environmental effects should be more than offset by decreases in predicted fatigue usage if re-analysis were conducted to reduce the conservatisms that are present in existing component fatigue evaluations

  14. A study on technical issues of materials and design bases in ASME section III subsection NH code

    International Nuclear Information System (INIS)

    Lee, Hyeong Yeon; Kim, Jong Bum; Yoo, Bong

    2000-12-01

    In this study, an analysis of evaluation report by ORNL on the technical issues of elevated temperatures design guide line, ASME Code Section III Subsection NH was conducted and a brief evaluation procedure of the creep-fatigue damage was presented. ORNL published the report in 1993 and reviewed the issue areas where code rules or regulatory guides may be lacking or inadequate to ensure safe operation over the expected life cycles for liquid metal reactor systems. From historical viewpoint of the ASME NH code development, ASME Code Case 47 was changed much in 1989 edition, which includes the stress relaxation behavior in creep damage evaluation. Afterwards the 1992 version of CC N-47 was upgraded to Subsection NH in 1995 edition, which is the same with that of CC N-47 1992 edition except few material data. This report brings up the technical and regulatory issues that can not guarantee the safe and reliable operation of the ALMR which got the conceptual design certification from NRC. Twenty three technical issues were raised and settlement methodology were proposed. Additionally, the status of items approved by ASME code subgroup of elevated temperature design committee for the revision of the most recent 1998 edition of ASME NH was described

  15. 46 CFR 54.01-2 - Adoption of division 1 of section VIII of the ASME Boiler and Pressure Vessel Code.

    Science.gov (United States)

    2010-10-01

    ... Boiler and Pressure Vessel Code. 54.01-2 Section 54.01-2 Shipping COAST GUARD, DEPARTMENT OF HOMELAND... division 1 of section VIII of the ASME Boiler and Pressure Vessel Code. (a) Pressure vessels shall be designed, constructed, and inspected in accordance with section VIII of the ASME Boiler and Pressure Vessel...

  16. 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.

  17. Development and application of proposed ASME Section XI Code changes for risk-based inspection of piping

    International Nuclear Information System (INIS)

    West, R.A.

    1996-01-01

    This synopsis has been written to describe a perspective on the development and application of ASME Section XI Code changes for risk-based inspection of piping. The content is specifically related to the use of risk-based technology for Inservice Inspection (ISI) of piping and efforts made to support the ASME Research/Westinghouse Owners Group/Millstone Unit 3 approach for use of this technology. The opinions contained herein may or may not reflect those of the ASME Codes and Standards Committees responsible for these activities. In order to take such a detailed technical subject and put it into an understandable format, the author has chosen to provide an analogy to simplify what is actually taking place. Risk-based technology in the ISI of piping can be likened to the process of making and using specifically ground prescription glasses to allow for better vision. It provides a process to develop and use these uniquely ground glasses that will dynamically focus on all the locations and obstacles within a plant's piping systems that could cause that plant to trip and fall; more importantly it identifies the locations where the fall could possibly hurt someone else. In this way, Nuclear Safety is being addressed

  18. 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

  19. 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.)

  20. 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.

  1. 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.

  2. 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.''

  3. 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

  4. Section XI ASME B and PV CODE, future trends, nondestructive examination requirements

    International Nuclear Information System (INIS)

    Cowfer, C.D.

    1984-01-01

    Service Induced Flaws such as intergranular stress corrosion cracking, Round Robin Programs like PISC II, and related developments in UT methodology and signal processing the past one to two years will have major impact on the Code and future NDE requirements. The performance of NDE has become a high exposure item which demands high reliability and accuracy; terms generally not used with field NDE in the past. The trend is out of ''cookbook'' requirements and into performance demonstration for personnel, procedures and equipment. This paper highlights the current major transition in the Code regarding NDE performance from the viewpoint of the author's involvement

  5. 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

  6. 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?

  7. 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)

  8. 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.

  9. Case study of the propagation of a small flaw under PWR loading conditions and comparison with the ASME code design life. Comparison of ASME Code Sections III and XI

    International Nuclear Information System (INIS)

    Yahr, G.T.; Gwaltney, R.C.; Richardson, A.K.; Server, W.L.

    1986-01-01

    A cooperative study was performed by EG and G Idaho, Inc., and Oak Ridge National Laboratory to investigate the degree of conservatism and consistency in the ASME Boiler and Pressure Vessel Code Section III fatigue evaluation procedure and Section XI flaw acceptance standards. A single, realistic, sample problem was analyzed to determine the significance of certain points of criticism made of an earlier parametric study by staff members of the Division of Engineering Standards of the Nuclear Regulatory Commission. The problem was based on a semielliptical flaw located on the inside surface of the hot-leg piping at the reactor vessel safe-end weld for the Zion 1 pressurized-water reactor (PWR). Two main criteria were used in selecting the problem; first, it should be a straight pipe to minimize the computational expense; second, it should exhibit as high a cumulative usage factor as possible. Although the problem selected has one of the highest cumulative usage factors of any straight pipe in the primary system of PWRs, it is still very low. The Code Section III fatigue usage factor was only 0.00046, assuming it was in the as-welded condition, and fatigue crack-growth analyses predicted negligible crack growth during the 40-year design life. When the analyses were extended past the design life, the usage factor was less than 1.0 when the flaw had propagated to failure. The current study shows that the criticism of the earlier report should not detract from the conclusion that if a component experiences a high level of cyclic stress corresponding to a fatigue usage factor near 1.0, very small cracks can propagate to unacceptable sizes

  10. 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

  11. 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)

  12. 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

  13. 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.

  14. 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

  15. 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

  16. Performance Demonstration Initiative U.S. implementation of ASME B and PV code section 11 Appendix 8

    International Nuclear Information System (INIS)

    Becker, F.L.; Ammirato, F.; Huffman, K.

    1994-01-01

    New requirements have now been added to Section 11 as mandatory Appendix 8, ''Performance Demonstration Requirements for Ultrasonic Examination systems''. The appendix was recently published and incorporates performance demonstration requirements for ultrasonic examination equipment, procedures, and personnel. These new requirements will have far reaching and significant impact on the conduct of ISI at all nuclear power plants. For the first time since Section 11 was issued in 1970, the effectiveness of ultrasonic examination procedures and the proficiency of examiners must be demonstrated on reactor pressure vessel (RPV), piping, and bolting markups containing real flaws, Recognizing the importance and complexity of Appendix 8 implementation, representatives from all US nuclear utilities have formed the Performance Demonstration Initiative (PDI) to implement Appendix 8 to provide for uniform implementation

  17. 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

  18. 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.

  19. 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

  20. 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)

  1. 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)

  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. 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)

  5. 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)

  6. 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

  7. 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

  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. 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)

  10. 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

  11. 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

  12. 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

  13. 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

  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. 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

  17. 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)

  18. Twenty years of fracture mechanics and flaw evaluation applications in the ASME Nuclear Code

    International Nuclear Information System (INIS)

    Riccardella, P.C.

    1991-01-01

    The paper presents a retrospective on the development and applications of fracture mechanics-based toughness requirements and flaw evaluation methodology in Sections III and XI of the ASME Code. Section III developments range from the rules and requirements for thick section Class 1 pressure vessels to thinner section components in other Classes. Section XI applications include flaw acceptance standards and evaluation methodology for various components ranging from pressure vessels to thins section piping of carbon and austenitic steels. The experience gained in operating plant applications of these rules and procedures are also discussed

  19. 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)

  20. 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)

  1. The ASME Section 11 Special Working Group On Plant Life Extension

    International Nuclear Information System (INIS)

    Katz, L.R.

    1990-01-01

    The codes and standards applicable to plant life extension have not been identified in the U.S. at this time. However, several initiatives have been taken to establish specific codes and standards pertaining to nuclear plant life extension (PLEX). One of these initiatives, sponsored by ASME, is the Section XI Special Working Group on Plant Life Extension (SWG-PLEX). The SWG-PLEX reports to the ASME Section XI Subcommittee and is responsible for recommending or drafting rules and requirements for modifying Section XI to accommodate age-related degradation to support nuclear plant life extension. This paper summarizes the results and reports the activities of the SWG-PLEX during the 1989/1990 period

  2. 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

  3. The ASME Boiler and Pressure Vessel Code: overview

    International Nuclear Information System (INIS)

    Farr, J.R.

    1987-01-01

    To become familiar with the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers, it is necessary to understand the history, organization, and operation of the Boiler Code Committee as well as to become familiar with the important aspects of each Section of the Code. This chapter will review the background and contents of the Code as well as give a review of the salient contents of most sections. (author)

  4. 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)

  5. 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

  6. 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

  7. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

  8. 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.

  9. 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

  10. Evaluation of ASME code flaw analysis procedure using the influence function method for application to PWR primary piping

    International Nuclear Information System (INIS)

    Hong, S.Y.; Yeater, M.L.

    1985-01-01

    This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)

  11. 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

  12. Nupack, the new ASME code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as Nupack, has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used for the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper

  13. Draft ASME code case on ductile cast iron for transport packaging

    International Nuclear Information System (INIS)

    Saegusa, T.; Arai, T.; Hirose, M.; Kobayashi, T.; Tezuka, Y.; Urabe, N.; Hueggenberg, R.

    2004-01-01

    The current Rules for Construction of ''Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste'' of Division 3 in Section III of ASME Code (2001 Edition) does not include ductile cast iron in its list of materials permitted for use. The Rules specify required fracture toughness values of ferritic steel material for nominal wall thickness 5/8 to 12 inches (16 to 305 mm). New rule for ductile cast iron for transport packaging of which wall thickness is greater than 12 inches (305mm) is required

  14. Enactment of KEPIC MNH Based on 2007 ASME BPVC Section III Division 1, Subsection NH: Class 1 Components in Elevated Temperature Service

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Kim, J. B.; Lee, H. Y.; Park, C. G.

    2008-11-01

    This report is a draft of an enactment of KEPIC MNH based on 2007 ASME Boiler and Pressure Vessel Code, Section III, Division 1 Subsection NH for Class 1 Components in Elevated Temperature Service and contains of ASME Article NH-3000 design, the mandatory appendix I-14, and non-mandatory appendices T and X

  15. Enactment of KEPIC MNH Based on 2007 ASME BPVC Section III Division 1, Subsection NH: Class 1 Components in Elevated Temperature Service

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Kim, J. B.; Lee, H. Y.; Park, C. G

    2008-11-15

    This report is a draft of an enactment of KEPIC MNH based on 2007 ASME Boiler and Pressure Vessel Code, Section III, Division 1 Subsection NH for Class 1 Components in Elevated Temperature Service and contains of ASME Article NH-3000 design, the mandatory appendix I-14, and non-mandatory appendices T and X.

  16. 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

  17. 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

  18. 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.

  19. 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

  20. Elastic-plastic stress analysis and ASME code evaluation of a bottomhead penetration in a reactor pressure vessel

    International Nuclear Information System (INIS)

    Ranganath, S.

    1979-01-01

    Nuclear pressure vessel components are designed to meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Specifically, the design must satisfy the limits on stress range and fatigue usage prescribed in NB-3200, Section III ASME Code for the various design and operating conditions for the component. The Code requirements assure that the component does not experience gross yielding and that in general, elastic shakedown occurs following cyclic loading. When elastic stress analysis is performed this can be shown by meeting the limits in the Code on Primary and Primary plus Secondary (P+Q) stress intensities. However, when the P+Q limits cannot be met and elastic Shakedown cannot be demonstrated, plastic analysis may be performed to meet the requirements of the Code. This paper describes the elastic-plastic stress analysis of a Boiling Water Reactor Vessel bottom head in-core penetration and illustrates how plastic analysis can be used in ASME Code evaluations to show Code compliance. Details of the thermal analysis, elastic-plastic stress analysis and fatigue evaluation are presented and it is shown that the in-core penetration satisfies the code requirements. 6 refs

  1. 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

  2. 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

    robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.

  3. 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

  4. 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)

  5. 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

  6. 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

  7. Comparison of elevated temperature design codes of ASME Subsection NH and RCC-MRx

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon, E-mail: hylee@kaeri.re.kr

    2016-11-15

    Highlights: • Comparison of elevated temperature design (ETD) codes was made. • Material properties and evaluation procedures were compared. • Two heat-resistant materials of Grade 91 steel and austenitic stainless steel 316 are the target materials in the present study. • Application of the ETD codes to Generation IV reactor components and a comparison of the conservatism was conducted. - Abstract: The elevated temperature design (ETD) codes are used for the design evaluation of Generation IV (Gen IV) reactor systems such as sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), and very high temperature reactor (VHTR). In the present study, ETD code comparisons were made in terms of the material properties and design evaluation procedures for the recent versions of the two major ETD codes, ASME Section III Subsection NH and RCC-MRx. Conservatism in the design evaluation procedures was quantified and compared based on the evaluation results for SFR components as per the two ETD codes. The target materials are austenitic stainless steel 316 and Mod.9Cr-1Mo steel, which are the major two materials in a Gen IV SFR. The differences in the design evaluation procedures as well as the material properties in the two ETD codes are highlighted.

  8. Present activity in ASME Section XI regarding risk-informed maintenance

    International Nuclear Information System (INIS)

    Hedden, Owen; Chockie, Alan

    2005-01-01

    Since 1996 Section XI of the ASME Boiler and Pressure Vessel Code has actively incorporated risk-informed concepts. The risk-informed process provides a framework for allocating inspection resources in a cost-effective manner and helps focus inspections where most critical for plant safety. Based on the success of the risk-informed ISI piping applications at US and non-US plants, Section XI has refined existing Code Cases and expanded the use of the risk-informed process to a variety of high-risk components and systems. The risk informed approach started in the area of inspection and is now being expanded to other plant maintenance activities. This article summarizes the Section XI actions and the continued development of the risk-informed process to improve nuclear plant maintenance. (author)

  9. Impact of ACI-ASME code on design and construction of nuclear containment structures

    International Nuclear Information System (INIS)

    Reedy, R.F.

    1978-01-01

    The effect of the ACI-ASME code for design and construction of concrete containment structures on the nuclear and concrete industries is examined. Topics covered include purpose of the code, general requirements, responsibilities and duties, design and construction specifications, quality assurance, inspection, the liner, and stamping

  10. 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

  11. 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)

  12. Nupack, the new Asme code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as 'Nupack', has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper. Participation in the Nupack development work described in this paper was supported by the U.S. Department of Energy. (authors)

  13. Researching on knowledge architecture of design by analysis based on ASME code

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan

    2003-01-01

    The quality of knowledge-based system's knowledge architecture is one of decisive factors of knowledge-based system's validity and rationality. For designing the ASME code knowledge based system, this paper presents a knowledge acquisition method which is extracting knowledge through document analysis consulted domain experts' knowledge. Then the paper describes knowledge architecture of design by analysis based on the related rules in ASME code. The knowledge of the knowledge architecture is divided into two categories: one is empirical knowledge, and another is ASME code knowledge. Applied as the basement of the knowledge architecture, a general procedural process of design by analysis that is met the engineering design requirements and designers' conventional mode is generalized and explained detailed in the paper. For the sake of improving inference efficiency and concurrent computation of KBS, a kind of knowledge Petri net (KPN) model is proposed and adopted in expressing the knowledge architecture. Furthermore, for validating and verifying of the empirical rules, five knowledge validation and verification theorems are given in the paper. Moreover the research production is applicable to design the knowledge architecture of ASME codes or other engineering standards. (author)

  14. Elastic creep-fatigue evaluation for ASME [American Society of Mechanical Engineers] code

    International Nuclear Information System (INIS)

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

    1987-02-01

    Reassessment of past ASME N-47 creep-fatigue rules have been under way by committee members. The new proposed elastic creep-fatigue methods are easier to apply than those previously in the code case. They also provide a wider range of practical application while still providing conservative assessments. It is expected that new N-47 code rules for elastic creep-fatigue evaluation will be adopted in the near future

  15. 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

  16. 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

  17. Procedures of ASME code case N-201 for KALIMER. Reactor internal structures

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Yoo, B.

    2001-02-01

    The main objective of this report is to describe the design procedure of ASME Boiler and Pressure Vessel Code, Code Case N-201-4, which is an elevated temperature structural design code of the Nuclear reactor internal structures, checking the criteria of stress limit, accumulated inelastic strain and deformation, creep-fatigue damage, and buckling limit. As one of examples, the creep-fatigue damage evaluations are carried out for the KALIMER reactor internal structures of baffle annulus. This report is expected to be very useful in evaluating the structural integrity of the liquid metal reactor operating under an elevated temperature

  18. Application of the ASME code in the design of the GA-4 and GA-9 casks

    International Nuclear Information System (INIS)

    Mings, W.J.; Koploy, M.A.

    1992-01-01

    General Atomics (GA) is developing two spent fuel shipping casks for transport by legal weight truck (LWT). The casks are designed to the loading, environmental conditions and safety requirements defined in Title 10 of the Code of Federal Regulations, Part 71 (10CFR71). To ensure that all components of the cask meet the 10CFR71 rules, GA established structural design criteria for each component based on NRC Regulatory Guides and the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paper discusses the criteria used for different cask components, how they were applied and the conservatism and safety margins built into the criteria and assumption

  19. Utility experience in code updating of equipment built to 1974 code, Section 3, Subsection NF

    International Nuclear Information System (INIS)

    Rao, K.R.; Deshpande, N.

    1990-01-01

    This paper addresses changes to ASME Code Subsection NF and reconciles the differences between the updated codes and the as built construction code, of ASME Section III, 1974 to which several nuclear plants have been built. Since Section III is revised every three years and replacement parts complying with the construction code are invariably not available from the plant stock inventory, parts must be procured from vendors who comply with the requirements of the latest codes. Aspects of the ASME code which reflect Subsection NF are identified and compared with the later Code editions and addenda, especially up to and including the 1974 ASME code used as the basis for the plant qualification. The concern of the regulatory agencies is that if later code allowables and provisions are adopted it is possible to reduce the safety margins of the construction code. Areas of concern are highlighted and the specific changes of later codes are discerned; adoption of which, would not sacrifice the intended safety margins of the codes to which plants are licensed

  20. Appropriate nominal stresses for use with ASME Code pressure-loading stress indices for nozzles

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1976-06-01

    This program is part of a cooperative effort with industry to develop and verify analytical methods for assessing the safety of nuclear pressure-vessel and piping-system design. The study of nominal stresses and stress indices described is part of a continuing study of design rules for nozzles in pressure vessels being coordinated by the PVRC Subcommittee on Reinforced Openings and External Loadings. Results from these studies are used by appropriate ASME Code groups in drafting new and improved design rules

  1. Modification of the ASME code z-factor for circumferential surface crack in nuclear ferritic pipings

    International Nuclear Information System (INIS)

    Choi, Young Hwan; Chung, Yon Ki; Koh, Wan Young; Lee, Joung Bae

    1996-01-01

    The purpose of this paper is to modify the ASME Code Z-Factor, which is used in the evaluation of circumferential surface crack in nuclear ferritic pipings. The ASME Code Z-Factor is a load multiplier to compensate plastic load with elasto-plastic load. The current ASME Code Z-Factor underestimates pipe maximum load. In this study, the original SC. TNP method is modified first because the original SC. TNP method has a problem that the maximum allowable load predicted from the original SC. TNP method is slightly higher than that measured from the experiment. Then the new Z-Factor is developed using the modified SC. TNP method. The desirability of both the modified SC. TNP method and the new Z-Factor is examined using the experimental results for the circumferential surface crack in pipings. The results show that (1) the modified SC. TNP method is good for predicting the circumferential surface crack behavior in pipings, and (2) the Z-Factor obtained from the modified SC. TNP method well predicts the behavior of circumferential surface crack in ferritic pipings. 30 refs., 13 figs., 4 tabs. (author)

  2. Recommendations to ASME for code guidelines and criteria for continued operation of equipment

    International Nuclear Information System (INIS)

    Harvey, J.F.

    1993-01-01

    In May 1988, the American Society of Mechanical Engineers, ASME, asked the Pressure Vessel Research Council, PVRC, to review the part it should play in the continued operation of equipment originally designed and fabricated to the ASME codes and rules. This was prompted solely by an economic opportunity in which the capital expenditures to replace plants was far more costly than evaluating, repairing, and extending the nominal design life of the individual component. For instance, nuclear plants are normally designed for a life of 40 years, while fossil fired facilities may have been designed for other time lives, yet at the end of their original design life may actually have many useful years remaining. While this action was economically prompted, it inherently involved a two-fold one; namely, (1) safety, (2) legal. There is no question of safety to operating personnel. While codes for fossil components do not specify design lives, their adoption by many states provides a legal means of procedure in event of a mishap. This recognizes a cradle-to-grave safety responsibility. It is toward maintaining ASMEs leadership as a code authority that this report has been prepared

  3. Materials and design bases issues in ASME Code Case N-47

    International Nuclear Information System (INIS)

    Huddleston, R.L.; Swindeman, R.W.

    1993-04-01

    A preliminary evaluation of the design bases (principally ASME Code Case N-47) was conducted for design and operation of reactors at elevated temperatures where the time-dependent effects of creep, creep-fatigue, and creep ratcheting are significant. Areas where Code rules or regulatory guides may be lacking or inadequate to ensure the operation over the expected life cycles for the next-generation advanced high-temperature reactor systems, with designs to be certified by the US Nuclear Regulatory Commission, have been identified as unresolved issues. Twenty-two unresolved issues were identified and brief scoping plans developed for resolving these issues

  4. 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

  5. New methods of analysis of materials strength data for the ASME Boiler and Pressure Vessel Code

    International Nuclear Information System (INIS)

    Booker, M.K.; Booker, B.L.P.

    1980-01-01

    Tensile and creep data of the type used to establish allowable stress levels for the ASME Boiler and Pressure Vessel Code have been examined for type 321H stainless steel. Both inhomogeneous, unbalanced data sets and well-planned homogeneous data sets have been examined. Data have been analyzed by implementing standard manual techniques on a modern digital computer. In addition, more sophisticated techniques, practical only through the use of the computer, have been applied. The result clearly demonstrates the efficacy of computerized techniques for these types of analyses

  6. Progress toward NuPack, the ASME code for Type B containments

    International Nuclear Information System (INIS)

    Turula, P.

    1995-01-01

    This paper presented a brief status report on the development of an ASME Code Division for nuclear packaging and discussed some of the more interesting policy decisions as to what is and is not covered in terms of analytical methods, criteria, scope, and other aspects. The process of the development of this Division has been very slow and inconsistent. There were many participants with many diverse interests. The Division 3 rules are close to being ready to be issued. They are a compromise between many needs and the result is certainly not perfect. Opportunities for fine tuning and expanding this document will present themselves after it is issued as future needs become clear

  7. 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)

  8. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and GEN IV

    International Nuclear Information System (INIS)

    O'Donnell, William J.; Griffin, Donald S.

    2007-01-01

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  9. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    Energy Technology Data Exchange (ETDEWEB)

    William J. O’Donnell; Donald S. Griffin

    2007-05-07

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  10. Comparison of ASME Code NB-3200 and NB-3600 results for fatigue analysis of B31.1 branch nozzles

    International Nuclear Information System (INIS)

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

    1996-01-01

    Fatigue analyses wre conducted on two reactor coolant system branch nozzles in an operating PWR designed to the B31.1 Code, for which no explicit fatigue analysis was required by the licensing basis. These analyses were performed as part of resolving issues connected with NRC's Fatigue Action Plan to determine if the cumulative usage factor (CUF) for these nozzles, using the 1992 ASME Code and representative PWR transients, were comparable to nozzles designed and analyzed to the ASME Code. Both NB-3200 and NB-3600 ASME Code methods were used. NB-3200 analyses included the development of finite element models for each nozzle. Although detailed thermal transients were not available for the plant analyzed, representative transients from similar PWRs were applied in each method. CUFs calculated using NB-3200 methods were significantly less than using NB-3600. The paper points out differences in analysis methods and highlights difficulties and unknowns in performing more detailed analyses to reduce conservative assumptions

  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. Code on nuclear air and gas treatment ASME/ANSI AG-1

    International Nuclear Information System (INIS)

    Miller, W.H. Jr.

    1993-01-01

    The focus of this panel is on equipment code section work over the past two years. Major topics include changes in Filter Code Sections, revamping of the Ductwork Code Section, and emergence of an improved I ampersand C Code Section. Actual applications of AG-1 are to be discussed by CONAGT members. Remaining time will be devoted to fielding questions concerning ASMA/ANSI AG-1

  13. 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

  14. 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.

  15. 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)

  16. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Hee-Jin [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Ha, Min-Su, E-mail: msha12@nfri.re.kr [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Sa-Woong; Jung, Hun-Chea [ITER Korea, National Fusion Research Institute, 169-148 Gwahak-Ro, Yuseong-Gu, Daejeon (Korea, Republic of); Kim, Duck-Hoi [ITER Organization, Route de Vinon sur Verdon - CS 90046, 13067 Sant Paul Lez Durance (France)

    2016-11-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K{sub e} factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  17. Comparative evaluation of structural integrity for ITER blanket shield block based on SDC-IC and ASME code

    International Nuclear Information System (INIS)

    Shim, Hee-Jin; Ha, Min-Su; Kim, Sa-Woong; Jung, Hun-Chea; Kim, Duck-Hoi

    2016-01-01

    Highlights: • The procedure of structural integrity and fatigue assessment was described. • Case studies were performed according to both SDC-IC and ASME Sec. • III codes The conservatism of the ASME code was demonstrated. • The study only covers the specifically comparable case about fatigue usage factor. - Abstract: The ITER blanket Shield Block is a bulk structure to absorb radiation and to provide thermal shielding to vacuum vessel and external vessel components, therefore the most significant load for Shield Block is the thermal load. In the previous study, the thermo-mechanical analysis has been performed under the inductive operation as representative loading condition. And the fatigue evaluations were conducted to assure structural integrity for Shield Block according to Structural Design Criteria for In-vessel Components (SDC-IC) which provided by ITER Organization (IO) based on the code of RCC-MR. Generally, ASME code (especially, B&PV Sec. III) is widely applied for design of nuclear components, and is usually well known as more conservative than other specific codes. For the view point of the fatigue assessment, ASME code is very conservative compared with SDC-IC in terms of the reflected K_e factor, design fatigue curve and other factors. Therefore, an accurate fatigue assessment comparison is needed to measure of conservatism. The purpose of this study is to provide the fatigue usage comparison resulting from the specified operating conditions shall be evaluated for Shield Block based on both SDC-IC and ASME code, and to discuss the conservatism of the results.

  18. 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)

  19. 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

  20. Fatigue-crack propagation behavior of steels in vacuum, and implications for ASME Section 11 crack growth analyses

    International Nuclear Information System (INIS)

    James, L.A.

    1985-08-01

    Section XI of the ASME Boiler and Pressure Vessel Code provides rules for the analysis of structures for which cracks or crack-like flaws have been discovered during inservice inspection. The Code provides rules for the analysis of both surface flaws as well as flaws that are embedded within the wall of the pressure vessel. In the case of surface flaws, the Code provides fatigue crack growth rate relationships for typical nuclear pressure vessel steels (e.g., ASTM A508-2 and A533-B) cycled in water environments typical of those in light-water reactors (LWR). However, for the case of embedded cracks, the Code provides crack growth relationships based on results from specimens that were cycled in an elevated temperature air environment. Although these latter relationships are often referred to as applying to ''inert'' environments, the results of this paper will show that an elevated temperature air environment is anything but inert, and that use of such relationships can result in overly pessimistic estimates of fatigue-crack growth lifetimes of embedded cracks. The reason, of course, is that embedded cracks grow in an environment that is probably much closer to a vacuum than an air environment

  1. U.F.F.A.: A numerical procedure for fatigue analysis according to ASME code

    International Nuclear Information System (INIS)

    Bellettato, W.; Ticozzi, C.; Zucchini, C.

    1981-01-01

    A new procedure is developed, which employs some already used methodologies and brings some new concepts. The computer code UFFA employs the so obtained procedure. This paper in the first part describes the methodology used for the usage factor calculation, in the second part carries a general description of the code and in the third part shows some example and their respective results. We suppose an elastic behaviour of the materials and we do not consider the effect of the application order of the loads. Moreover, we suppose valid the hypothesis of cumulative damage, that is we apply the Miner's rule. One of the problems in the nuclear components fatigue analysis is that in the load histories there is a high number of operational cycles for which we cannot specify a succession in the time. Therefore, it was introduced the concept of 'level' (or steady working status) by which we can approximate the load conditions in realistic way. As regard the problem of multiaxial cases, it is possible to show that it is not right, an neither conservative, to make a distinguished analysis of the 3 stress differences and then take the maximum of the 3 compuoted usage factors as component usage factor. Indeed, as the stresses act on the structure at the same time, it is necessary a contemporary analysis of the 3 stress difference curves. The computer code can deal as well with the case of sher stresses (varying principal stress directions) through the ASME 'normalization' procedure. The results of the UFFA program, compared with the results of other programs used at present, come up to the expectations. (orig./HP)

  2. Developments on ASME Code Cases to Risk-Informed Repair/Replacement Activities in Support of Risk-Informed Regulation Initiatives

    International Nuclear Information System (INIS)

    Balkey, Kenneth R.; Holston, William C.

    2002-01-01

    ASME Code Case N-658, 'Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities' and Code Case N-660, 'Alternative Repair/Replacement Requirements For Items Classified In Accordance With Risk-Informed Processes' are being completed to expand the breadth of risk-informed requirements for pressure-retaining items. This initiative, which is built from prior ASME Section XI risk-informed inservice inspection developments over the past decade, has been undertaken in conjunction with U.S. risk-informed regulation efforts. The U.S. Nuclear Regulatory Commission (NRC) is working with the industry on risk informing Title 10 Code of Federal Regulations Part 50 (10CFR50). The Nuclear Regulatory Commission's basic proposal is to allow modification of some of the special treatment requirements of 10CFR50. Their effort is proceeding via an Advanced Notice of Public Rulemaking, March 3, 2000, and an announcement of Availability of Draft Rule Wording, November 29, 2001, to add 10 CFR 50.69, 'Risk-Informed Treatment of Structures, Systems and Components'. A parallel task by the Nuclear Energy Institute (NEI) to develop a guideline on how to implement the results of the rulemaking is also well underway via NEI 00-04 (Draft Revision B), 'Option 2 Implementation Guideline', May 2001. This paper summarizes the content and status of approval of the proposed ASME Code Cases, including how they relate to the above NRC and NEI efforts. Some initial results from trial application of the Code Cases will also be cited. (authors)

  3. Applicability of ASME sections III and VIII and of B31.1 and B31.3 to DOE facilities

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1993-01-01

    DOE order 6430.1A Section 1300-3.2 requires that open-quotes....safety class items shall be designed to the ASME Boiler and Pressure Vessel Code (ASME Section III) or to other comparable safety-related codes and standards...close quotes. This requirement raises a host of technical and practical questions which, to the author's knowledge, have not been fully addressed in the past. This paper attempts to cover the following essential points, in order: Evolution of industry reference codes, Code scope, Safety margins, Logistical considerations, Costs, Backfit considerations. These points are covered in the context of a reference safety class piping and vessel system at a DOE facility which processes radioactive fluids, and which this paper calls the open-quotes reference DOE nuclear facilityclose quotes. In the conclusion, the author proposes three alternatives for code applicability which are ranked technically as open-quotes goodclose quotes, open-quotes closer to 6430.1Aclose quotes and open-quotes closest to 6430.1Aclose quotes. It is however questionable whether the alternatives which are labeled open-quotes closerclose quotes and open-quotes closestclose quotes are practically viable, as will be discussed

  4. 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

  5. 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

  6. Report on the FY17 Development of Computer Program for ASME Section III, Division 5, Subsection HB, Subpart B Rules

    Energy Technology Data Exchange (ETDEWEB)

    Swindeman, M. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Jetter, R. I. [Argonne National Lab. (ANL), Argonne, IL (United States); Sham, T. -L. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-01-01

    One of the objectives of the high temperature design methodology activities is to develop and validate both improvements and the basic features of ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB, Subpart B (HBB). The overall scope of this task is to develop a computer program to aid assessment procedures of components under specified loading conditions in accordance with the elevated temperature design requirements for Division 5 Class A components. There are many features and alternative paths of varying complexity in HBB. The initial focus of this computer program is a basic path through the various options for a single reference material, 316H stainless steel. However, the computer program is being structured for eventual incorporation all of the features and permitted materials of HBB. This report will first provide a description of the overall computer program, particular challenges in developing numerical procedures for the assessment, and an overall approach to computer program development. This is followed by a more comprehensive appendix, which is the draft computer program manual for the program development. The strain limits rules have been implemented in the computer program. The evaluation of creep-fatigue damage will be implemented in future work scope.

  7. Comparisons of ratchetting analysis methods using RCC-M, RCC-MR and ASME codes

    International Nuclear Information System (INIS)

    Yang Yu; Cabrillat, M.T.

    2005-01-01

    The present paper compares the simplified ratcheting analysis methods used in RCC-M, RCC-MR and ASME with some examples. Firstly, comparisons of the methods in RCC-M and efficiency diagram in RCC-MR are investigated. A special method is used to describe these two methods with curves in one coordinate, and the different conservation is demonstrated. RCC-M method is also be interpreted by SR (second ratio) and v (efficiency index) which is used in RCC-MR. Hence, we can easily compare the previous two methods by defining SR as abscissa and v as ordinate and plotting two curves of them. Secondly, comparisons of the efficiency curve in RCC-MR and methods in ASME-NH APPENDIX T are investigated, with significant creep. At last, two practical evaluations are performed to show the comparisons of aforementioned methods. (authors)

  8. Application of the ASME-code-case N 47 to a typical thickwalled HTR-component made of Incoloy 800

    International Nuclear Information System (INIS)

    Kemter, F.; Schmidt, A.

    Several components of the HTR-plant are exposed to temperatures beyond 500 0 C, i.e. within the high-temperature range. The service life of those components is not only limited by fatigue damage but also mainly by creep damage and accumulated inelastic strain. These can be conservatively estimated according to the ASME-Code (high temperature part CC N47) by means of the results of elastic calculations, yet this simplified method to provide evidence often leads to calculated overloads such as the present case of the live steam collector of the steam generator of a HTR. For providing the evidence that the actual loads of the component are within permissible limits, comprehensive inelastic analyses have to be referred to in such a case. The two-dimensional inelastic analysis which is reported here in detail shows that the creep and fatigue failure as well as the inelastic extensions of the live steam collectors accumulated during the service time are below the permissible limit stated in the ASME-Code and failure of those components while used in the reactor can this be excluded. (orig.) [de

  9. IE Information Notice No. 85-33: Undersized nozzle-to-shell welded joints in tanks and heat exchangers constructed under the rules of the ASME boiler and vessel code

    International Nuclear Information System (INIS)

    Jordan, E.L.

    1993-01-01

    During the CAT (Construction Appraisal Team) inspections conducted at the River Bend, Shearon Harris, and Braidwood nuclear power projects, the NRC identified undersized nozzle-to-shell welded joints (ASME Category D joints) in tanks and heat exchangers manufactured by various vendors. Specifically, four main steam isolation valve air accumulator tanks were found to have undersized nozzle-to-shell joints at the River Bend plant; seven tanks were found to have undersized nozzle-to-shell weld reinforcements at the Shearon Harris Station; eight tanks and two heat exchangers were found to have undersized nozzle-to-shell weld reinforcements at Braidwood Station. These tanks and heat exchangers were Code stamped and certified as being constructed in accordance with the requirements of the ASME Code. The ASME Code, Section III (NX-3352.4) requires that nozzle-to-shell welded joints have reinforcement (t c ) of 0.7t p or 1/4 inch, whichever is less, where t p is the thickness of the penetrating part. Some of the inspected welded joints did not have the minimum weld reinforcement (t c ) required by the Code. Other joints had the minimum weld reinforcement (t c ) required by the Code, but were found to be undersized with respect to the sizes specified on the applicable construction drawings

  10. Review of ASME code criteria for control of primary loads on nuclear piping system branch connections and recommendations for additional development work

    International Nuclear Information System (INIS)

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

    1993-11-01

    This report collects and uses available data to reexamine the criteria for controlling primary loads in nuclear piping branch connections as expressed in Section III of the ASME Boiler and Pressure Vessel Code. In particular, the primary load stress indices given in NB-3650 and NB-3683 are reexamined. The report concludes that the present usage of the stress indices in the criteria equations should be continued. However, the complex treatment of combined branch and run moments is not supported by available information. Therefore, it is recommended that this combined loading evaluation procedure be replaced for primary loads by the separate leg evaluation procedure specified in NC/ND-3653.3(c) and NC/ND-3653.3(d). No recommendation is made for fatigue or secondary load evaluations for Class 1 piping. Further work should be done on the development of better criteria for treatment of combined branch and run moment effects

  11. 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)

  12. 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)

  13. Probabilistic evaluation of design S-N curve and reliability assessment of ASME code-based evaluation

    International Nuclear Information System (INIS)

    Zhao Yongxiang

    1999-01-01

    A probabilistic evaluating approach of design S-N curve and a reliability assessment approach of the ASME code-based evaluation are presented on the basis of Langer S-N model-based P-S-N curves. The P-S-N curves are estimated by a so-called general maximum likelihood method. This method can be applied to deal with the virtual stress amplitude-crack initial life data which have a characteristics of double random variables. Investigation of a set of the virtual stress amplitude-crack initial life (S-N) data of 1Cr18Ni9Ti austenitic stainless steel-welded joint reveals that the P-S-N curves can give a good prediction of scatter regularity of the S-N data. Probabilistic evaluation of the design S-N curve with 0.9999 survival probability has considered various uncertainties, besides of the scatter of the S-N data, to an appropriate extent. The ASME code-based evaluation with 20 reduction factor on the mean life is much more conservative than that with 2 reduction factor on the stress amplitude. Evaluation of the latter in 666.61 MPa virtual stress amplitude is equivalent to 0.999522 survival probability and in 2092.18 MPa virtual stress amplitude equivalent to 0.9999999995 survival probability. This means that the evaluation in the low loading level may be non-conservative and in contrast, too conservative in the high loading level. Cause is that the reduction factors are constants and the factors can not take into account the general observation that scatter of the N data increases with the loading level decreasing. This has indicated that it is necessary to apply the probabilistic approach to the evaluation of design S-N curve

  14. Evaluation of the plastic characteristics of piping products in relation to ASME code criteria

    International Nuclear Information System (INIS)

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

    1978-07-01

    Theories and test data relevant to the plastic characteristics of piping products are presented and compared with Code Equations in NB-3652 for Class 1 piping; in NC/ND-3652.2 for Class 2 and Class 3 piping. Comparisons are made for (a) straight pipe, (b) elbows, (c) branch connections, and (d) tees. The status of data (or lack of data) for other piping components is discussed. Comparisons are made between available data and the Code equations for two typical piping materials, SA106 Grade B and SA312 TP304, for Code Design Limits, and Service Limits A, B, C, and D. Conditions under which the Code Limits cannot be shown to be conservative from available data are pointed out. Based on the results of the study, recommendations for Code revisions are presented, along with recommendations for additional work

  15. 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

  16. 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 ...

  17. CEASEMT system: the TEDEL code. Pipings - Plasticity - Dynamics - Statics - Buckling - Thermoplasticity - Creep - Large displacements - FLUIDS - SEISMS - ASME

    International Nuclear Information System (INIS)

    Hoffmann, Alain; Jeanpierre, Francoise; Axisa, Francois; Chevalier, Gerard; Lepareux, Michel.

    1977-01-01

    The TEDEL code is intended for elastic and plastic computation of three-dimensional pipes and frames with possible junction to shells. The structures are described with using assemblies of beam elements, or piping elements such as, curved pipes, 90 0 elbows, tees, any elements, the stiffness properties of which are given to TEDEL. TEDEL allows the dynamic computation of the structures: search of eigenfrequencies and eigenmodes of vibration, time response to any time-dependent canvassing. This response can be obtained either from recombining a number of eigenmodes, or from a direct numerical integration of the dynamics equation. In these last two cases TEDEL accounts for some possible damping. A TEDEL option allows critical buckling loads to be computed (Euler). The structures can offer any shapes comprising any number of materials. The data are readout without any format, and distributed in optional commands with a precise physical meaning: GEOMETRY, MATERIALS, LOAD, COMPUTATION, END. A dynamical memory control allows the size of the routine to be adapted to the problem to be treated. For pipings, an option is intended for an automatic checking of the stress level with regard to the limiting values of the ASME. Geometrical data, node positions, element numbering are given by COCO which also delivers perspective drawings for the structure to be studied. The results on magnetic tapes can be treated by the subroutines ESPACE-VISU-TEMPS [fr

  18. 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...

  19. 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

  20. 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.

  1. 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

  2. ASME section XI - design and access requirements for in-service inspection

    International Nuclear Information System (INIS)

    Davis, D.D.

    1982-01-01

    The Owner of a nuclear power plant has the regulatory commitment to perform Section XI in-service inspection throughout the service life of a plant. In anticipation of what will be needed to perform adequately the required examinations and tests, sub-article IWA-1500 of Section XI not only requires that sufficient access be provided to accommodate equipment and inspection personnel but also requires that other provisions be considered such as: component surface preparations, material selections, shielding, removal and storage of hardware, handling equipment, and provisions for repairs and replacements. It is, therefore, the owner's and the architect engineer's responsibility to ensure that proper design and access provisions are incorporated to enable the owner to meet his commitments. Since the architect engineer usually has the prime responsibility for the implementation of design criteria, the owner must ensure that these provisions be considered in each phase of design and construction. The benefits of this can result in shorter outages, more meaningful examinations and tests and less radiation exposure of inspection personnel. This paper will address in detail those topics that affect design and access provisions which need to be considered during the design and construction of a nuclear power plant. (author)

  3. 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.

  4. 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

  5. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    International Nuclear Information System (INIS)

    Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

    2007-01-01

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  6. ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

    International Nuclear Information System (INIS)

    Berry, Jan; Ferrada, Juan J.; Curd, Warren; Dell Orco, Giovanni; Barabash, Vladimir; Kim, Seokho H.

    2011-01-01

    During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

  7. 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

  8. Replacement of radiography with ultrasonic phased array for feeder tubes in CANDU reactors using ASME code case N-659-2

    International Nuclear Information System (INIS)

    Simmons, R.; Bower, Q.; Arseneau, S.

    2013-01-01

    In this paper we will discuss phased array technology for the replacement of radiography on new construction projects in the nuclear industry. Specifically, through the implementation of A.S.M.E. code N-659-2 and MetaPhase phased array services. Phased Array is not considered a new technique on in service welds in the nuclear industry; however it was unprecedented on new construction welds and required significant investment in regulatory approval (C.N.S.C.), technology research and development, regulatory, client and technician training for successful service implementation. This paper will illustrate the abilities and limitations associated in replacing radiography with MetaPhase, as well as the substantial benefits relative to increased production, improved weld quality, enhanced safety and overall project cost savings. (author)

  9. Automatic examination of nuclear reactor vessels with focused search units. Status and typical application to inspections performed in accordance with ASME code

    International Nuclear Information System (INIS)

    Verger, B.; Saglio, R.

    1981-05-01

    The use of focused search units in nuclear reactor vessel examinations has significantly increased the capability of flaw indication detection and characterization. These search units especially allow a more accurate sizing of indications and a more efficient follow up of their history. In this aspect, they are a unique tool in the area of safety and reliability of installations. It was this type of search unit which was adopted to perform the examinations required within the scope of inservice inspections of all P.W.R. reactors of the French nuclear program. This paper summarizes the results gathered through the 4l examinations performed over the last five years. A typical application of focused search units in automated inspections performed in accordance with ASME code requirements on P.W.R. nuclear reactor vessels is then described

  10. 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

  11. 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.

  12. 78 FR 79363 - Hazardous Materials: Adoption of ASME Code Section XII and the National Board Inspection Code

    Science.gov (United States)

    2013-12-30

    ... practices, improved materials, advances in welding, examination and testing. Notably, fracture mechanics did... materials, design, fabrication, examination, inspection, testing, certification, and over-pressure protection. \\2\\ ``Continued service'' is an all-inclusive term referring to inspection, testing, repair...

  13. 75 FR 80765 - Hazardous Materials: Adoption of ASME Code Section XII and the National Board Inspection Code

    Science.gov (United States)

    2010-12-23

    ... updates the NBIC and presents the updates on the National Board's website for public review in April-May... the design, construction, and certification of cargo tank motor vehicles, cryogenic portable tanks and....). You may review DOT's complete Privacy Act Statement in the Federal Register published on April 11...

  14. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

  15. The safety relief valve handbook design and use of process safety valves to ASME and International codes and standards

    CERN Document Server

    Hellemans, Marc

    2009-01-01

    The Safety Valve Handbook is a professional reference for design, process, instrumentation, plant and maintenance engineers who work with fluid flow and transportation systems in the process industries, which covers the chemical, oil and gas, water, paper and pulp, food and bio products and energy sectors. It meets the need of engineers who have responsibilities for specifying, installing, inspecting or maintaining safety valves and flow control systems. It will also be an important reference for process safety and loss prevention engineers, environmental engineers, and plant and process designers who need to understand the operation of safety valves in a wider equipment or plant design context. . No other publication is dedicated to safety valves or to the extensive codes and standards that govern their installation and use. A single source means users save time in searching for specific information about safety valves. . The Safety Valve Handbook contains all of the vital technical and standards informat...

  16. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    International Nuclear Information System (INIS)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik

    2017-01-01

    In this paper, the stress intensity limits, Sm and St of HT-9 were built for the structural criteria of an SFR fuel assembly. Sm is obtained from the ultimate strength. As for St, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of Smt, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as Smt under the temperature about 470 .deg. C which is relatively low temperature range and over 470 .deg. C with relatively short time duration as 1000 hours. And the St is adopted as Smt at over 470 .deg. C and long time duration over 34800 hours, and over 520 .deg. C and 104 hours too. And at over 570 .deg. C and 1000 hours, and at over 630 .deg. C and 100 hours, St is also adopted for Smt. To use the present result as design criteria, a stringent examination needs to be carried out, because those are calculated from the formulae of HT-9 without an experimental validation. Therefore, an experimental work on the mechanical properties of HT-9 will be necessary.

  17. 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

  18. International pressure vessels and piping codes and standards. Volume 2: Current perspectives; PVP-Volume 313-2

    International Nuclear Information System (INIS)

    Rao, K.R.; Asada, Yasuhide; Adams, T.M.

    1995-01-01

    The topics in this volume include: (1) Recent or imminent changes to Section 3 design sections; (2) Select perspectives of ASME Codes -- Section 3; (3) Select perspectives of Boiler and Pressure Vessel Codes -- an international outlook; (4) Select perspectives of Boiler and Pressure Vessel Codes -- ASME Code Sections 3, 8 and 11; (5) Codes and Standards Perspectives for Analysis; (6) Selected design perspectives on flow-accelerated corrosion and pressure vessel design and qualification; (7) Select Codes and Standards perspectives for design and operability; (8) Codes and Standards perspectives for operability; (9) What's new in the ASME Boiler and Pressure Vessel Code?; (10) A look at ongoing activities of ASME Sections 2 and 3; (11) A look at current activities of ASME Section 11; (12) A look at current activities of ASME Codes and Standards; (13) Simplified design methodology and design allowable stresses -- 1 and 2; (14) Introduction to Power Boilers, Section 1 of the ASME Code -- Part 1 and 2. Separate abstracts were prepared for most of the individual papers

  19. Rulemaking efforts on codes and standards

    International Nuclear Information System (INIS)

    Millman, G.C.

    1992-01-01

    Section 50.55a of the NRC regulations provides a mechanism for incorporating national codes and standards into the regulatory process. It incorporates by reference ASME Boiler and Pressure Vessel Code (ASME B and PV Code) Section 3 rules for construction and Section 11 rules for inservice inspection and inservice testing. The regulation is periodically amended to update these references. The rulemaking process, as applied to Section 50.55a amendments, is overviewed to familiarize users with associated internal activities of the NRC staff and the manner in which public comments are integrated into the process. The four ongoing rulemaking actions that would individually amend Section 50.55a are summarized. Two of the actions would directly impact requirements for inservice testing. Benefits accrued with NRC endorsement of the ASME B and PV Code, and possible future endorsement of the ASME Operations and Maintenance Code (ASME OM Code), are identified. Emphasis is placed on the need for code writing committees to be especially sensitive to user feedback on code rules incorporated into the regulatory process to ensure that the rules are complete, technically accurate, clear, practical, and enforceable

  20. Methodology and guidelines for evaluation of welded attachments on ASME Class 1,2, or 3 piping

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.; Rodabaugh, E.C.

    1985-01-01

    The ASME Boiler and Pressure Vessel Code, Section III Subsection NB/NC/ND-3600 provides simplified rules for the evaluation and qualification of piping components. The use of rectangular and hollow, circular welded attachments (hereafter called lugs and trunions) is sometimes necessary in order to provide support for piping systems. The Code provides a set of simple and conservative equations, the associated stress indicies, and specified limitations on their applicability for lugs on Class 1 and Class 2/3 piping in Code Cases N-122 and N-318 respectively. Two new ASME Section III Code Cases, N-391 and N-392, have been prepared to provide the corresponding design guidelines for specific trunion configurations. This paper presents the background on the major concepts involved in the development of these Code Cases and provides some general guidelines to the analysts and designers for the qualification of the attachments not covered by the Code Cases

  1. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  2. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  3. Draft ASME Boiler and Pressure Vessel Code Section III, Division 5, Section HB, Subsection B, Code Case for Alloy 617 and Background Documentation

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Julie Knibloe [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Alloy 617 is the leading candidate material for an intermediate heat exchanger for the very high temperature reactor. To evaluate the behavior of this material in the expected service conditions, strain controlled cyclic tests that include long hold times up to 240 minutes at maximum tensile strain were conducted at 850°C. In terms of the total number of cycles to failure, the fatigue resistance decreased when a hold time was added at peak tensile strain. Increases in the tensile hold duration degraded the creep fatigue resistance, at least to the investigated strain controlled hold time of up to 60 minutes at the 0.3% strain range and 240 minutes at the 1.0% strain range. The creep fatigue deformation mode is considered relative to the lack of saturation, or continually decreasing number of cycles to failure with increasing hold times. Additionally, preliminary values from the 850°C creep fatigue data are calculated for the creep fatigue damage diagram and have higher values of creep damage than those from tests at 950°C.

  4. Absorptive coding metasurface for further radar cross section reduction

    Science.gov (United States)

    Sui, Sai; Ma, Hua; Wang, Jiafu; Pang, Yongqiang; Feng, Mingde; Xu, Zhuo; Qu, Shaobo

    2018-02-01

    Lossless coding metasurfaces and metamaterial absorbers have been widely used for radar cross section (RCS) reduction and stealth applications, which merely depend on redirecting electromagnetic wave energy into various oblique angles or absorbing electromagnetic energy, respectively. Here, an absorptive coding metasurface capable of both the flexible manipulation of backward scattering and further wideband bistatic RCS reduction is proposed. The original idea is carried out by utilizing absorptive elements, such as metamaterial absorbers, to establish a coding metasurface. We establish an analytical connection between an arbitrary absorptive coding metasurface arrangement of both the amplitude and phase and its far-field pattern. Then, as an example, an absorptive coding metasurface is demonstrated as a nonperiodic metamaterial absorber, which indicates an expected better performance of RCS reduction than the traditional lossless coding metasurface and periodic metamaterial-absorber. Both theoretical analysis and full-wave simulation results show good accordance with the experiment.

  5. Update on ASME rules for spent nuclear fuel and high level radioactive material and waste storage containments

    International Nuclear Information System (INIS)

    Ralph S. Hill III; Foster, G.M.

    2005-01-01

    In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) was published. The Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. The Code Case has been incorporated into Section III of the Code as Division 3, Subsection WC, Class SC Storage Containments, and will be published in the 2005 Addenda. This paper provides an informative background and insight for these rules to provide Owners, regulators, designers, and fabricators with a more comprehensive understanding of the technical basis for these rules. (authors)

  6. Hauser*5, a computer code to calculate nuclear cross sections

    International Nuclear Information System (INIS)

    Mann, F.M.

    1979-07-01

    HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables

  7. Library of neutron cross sections of the Thermos code

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-10-01

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  8. PCS a code system for generating production cross section libraries

    International Nuclear Information System (INIS)

    Cox, L.J.

    1997-01-01

    This document outlines the use of the PCS Code System. It summarizes the execution process for generating FORMAT2000 production cross section files from FORMAT2000 reaction cross section files. It also describes the process of assembling the ASCII versions of the high energy production files made from ENDL and Mark Chadwick's calculations. Descriptions of the function of each code along with its input and output and use are given. This document is under construction. Please submit entries, suggestions, questions, and corrections to (ljc at sign llnl.gov) 3 tabs

  9. Report No. 3 -- Background of the factors of safety used in Divisions 1 of Sections III and XI of the ASME rules for nuclear vessels

    International Nuclear Information System (INIS)

    Cooper, W.E.

    1993-01-01

    This paper was prepared in 1984 as a quick background summary of the factors of safety used in the various Code rules as they apply to Class 1 reactor pressure vessels (RPVs). This document should be interpreted as the individual viewpoint of the author on the matters under discussion. The general approach followed is to cite basic reference documents, to lead the reader to the more important aspects which are pertinent to the subject material, and to provide a viewpoint on additional rules which might be either needed or desirable. After a discussion of the differences between Sections 3 and 11 in categorizing the loadings to be considered, the sequence of consideration is as follows: (1) The rules of the code that control initial construction, Section 3, Division (Construction = materials + design + examination + testing + inspection + certification). (2) The rules of Section 3 that specifically affect operation, Appendix G, which forms the basis of the pressure-temperature limits of the Technical Specifications. (3) The rules of Section 11 as they affect ''Acceptance by Examination'' and ''Acceptance by Evaluation''. (4) The special considerations of: (a) Limits for Emergency and Faulted Conditions (b) Elastic-Plastic Fracture Mechanics (c) Low Upper-Shelf Energy (USE) Materials (d) Pressure Thermal Shock (PTS)

  10. Relationship between various pressure vessel and piping codes

    International Nuclear Information System (INIS)

    Canonico, D.A.

    1976-01-01

    Section VIII of the ASME Code provides stress allowable values for material specifications that are provided in Section II Parts A and B. Since the adoption of the ASME Code over 60 years ago the incidence of failure has been greatly reduced. The Codes are currently based on strength criteria and advancements in the technology of fracture toughness and fracture mechanics should permit an even greater degree of reliability and safety. This lecture discusses the various Sections of the Code. It describes the basis for the establishment of design stress allowables and promotes the idea of the use of fracture mechanics

  11. Section 60 revisited (The Ontario Electrical Safety Code)

    Energy Technology Data Exchange (ETDEWEB)

    Olechna, T.

    2003-06-01

    Recent changes to the Ontario Electrical Safety Code (OESC), specifically the deletion of Section 60, Electrical Communication Systems, are discussed in an effort to explain the history behind the decision and the time frame of the changes. Communication systems include telephone, telegraph, data communications, intercommunications, wired music and paging systems. In brief, the deletion of Section 60 occurred in 1983, and resulted from the fact that communication-type wiring was historically the property of the communications utility and under federal jurisdiction. Since such equipment was under federal jurisdiction, they were not inspected in Ontario, hence the deletion of Section 60 from the Ontario Code. It should be noted that although Section 60 is deleted, a number of rules applicable to communications circuits are spread throughout various sections of the Code, notably in Rule 1-032 dealing with damage and interference, Rule 4-022 involving harmonics issues, Rule 12-904(2) regulates the use of conductors that are of different sources of voltage, and Rule 10-708 which specifies the spacing and bonding requirements for communications systems. The end result is that even though Section 60 was deleted, there are these and other rules in the OESC that have direct impact on communications circuits and in effect help to protect the integrity of the system.

  12. SCAMPI: A code package for cross-section processing

    International Nuclear Information System (INIS)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-01-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis

  13. SCAMPI: A code package for cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-04-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

  14. 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.)

  15. 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

  16. 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

  17. 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.

  18. Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2017-01-01

    Highlights: • P-T limits based on ASME K_I_a curve, K_I_C curve and RI method are presented. • Probabilistic and deterministic methods are used to evaluate P-T limits on RPV. • The feasibility of substituting P-T curves with more operational is demonstrated. • Warm-prestressing effect is critical in determining the fracture probability. - Abstract: The ASME Code Section XI-Appendix G defines the normal reactor startup (heat-up) and shut-down (cool-down) operation limits according to the fracture toughness requirement of reactor pressure vessel (RPV) materials. This paper investigates the effects of different pressure-temperature limit operations on structural integrity of a Taiwan domestic pressurized water reactor (PWR) pressure vessel. Three kinds of pressure-temperature limits based on different fracture toughness requirements – the K_I_a fracture toughness curve of ASME Section XI-Appendix G before 1998 editions, the K_I_C fracture toughness curve of ASME Section XI-Appendix G after 2001 editions, and the risk-informed revision method supplemented in ASME Section XI-Appendix G after 2013 editions, respectively, are established as the loading conditions. A series of probabilistic fracture mechanics analyses for the RPV are conducted employing ORNL’s FAVOR code considering various radiation embrittlement levels under these pressure-temperature limit conditions. It is found that the pressure-temperature operation limits which provide more operational flexibility may lead to higher fracture risks to the RPV. The cladding-induced shallow surface breaking flaws are the most critical and dominate the fracture probability of the RPV under pressure-temperature limit transients. Present study provides a risk-informed reference for the operation safety and regulation viewpoint of PWRs in Taiwan.

  19. 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.)

  20. Cracking the Sugar Code by Mass Spectrometry - An Invited Perspective in Honor of Dr. Catherine E. Costello, Recipient of the 2017 ASMS Distinguished Contribution Award

    Science.gov (United States)

    Mirgorodskaya, Ekaterina; Karlsson, Niclas G.; Sihlbom, Carina; Larson, Göran; Nilsson, Carol L.

    2018-04-01

    The structural study of glycans and glycoconjugates is essential to assign their roles in homeostasis, health, and disease. Once dominated by nuclear magnetic resonance spectroscopy, mass spectrometric methods have become the preferred toolbox for the determination of glycan structures at high sensitivity. The patterns of such structures in different cellular states now allow us to interpret the sugar codes in health and disease, based on structure-function relationships. Dr. Catherine E. Costello was the 2017 recipient of the American Society for Mass Spectrometry's Distinguished Contribution Award. In this Perspective article, we describe her seminal work in a historical and geographical context and review the impact of her research accomplishments in the field. 8[Figure not available: see fulltext.

  1. 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

  2. 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.

  3. 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)

  4. Benchmark of neutron production cross sections with Monte Carlo codes

    Science.gov (United States)

    Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

    2018-02-01

    Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first

  5. MUXS: a code to generate multigroup cross sections for sputtering calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1982-10-01

    This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

  6. Preliminary application of the draft code case for alloy 617 for a high temperature component

    International Nuclear Information System (INIS)

    Lee, Hyeong Yeon; Kim, Yong Wan; Song, Kee Nam

    2008-01-01

    The ASME draft Code Case for Alloy 617 was developed in the late 1980s for the design of very-high-temperature gas cooled reactors. The draft Code Case was patterned after the ASME Code Section III Subsection NH and was intended to cover Ni-Cr-Co-Mo Alloy 617 to 982 .deg. C (1800 .deg. F). But the draft Code Case is still in an incomplete status, lacking necessary material properties and design data. In this study, a preliminary evaluation on the creep-fatigue damage for a high temperature hot duct pipe structure has been carried out according to the draft Code Case. The evaluation procedures and results according to the draft Code Case for Alloy 617 material were compared with those of the ASME Subsection NH and RCC-MR for Alloy 800H material. It was shown that many data including material properties, fatigue and creep data should be supplemented for the draft Code Case. However, when the evaluation results on the creep-fatigue damage according to the draft Code Case, ASME-NH and RCC-MR were compared based on the preliminary evaluation, it was shown that the Alloy 617 results from the draft Code Case tended to be more resistant to the creep damage while less resistant to the fatigue damage than those from the ASME-NH and RCC-MR

  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. An intercomparison of medium energy cross-section codes

    International Nuclear Information System (INIS)

    Pearlstein, S.

    1988-05-01

    Five medium energy proton reaction cases are selected for benchmarking nuclear model codes. The quantities calculated are isotopic activation yields for 180 MeV protons on Al and 40-200 MeV protons on Co, and double differential neutron emission spectra from Al, Zr-90 and Pb-208 for 35, 80, 160, 318, and 800 presented consist of three types: a closed form preequilibrium plus evaporation model, an intranuclear-cascade and evaporation model, and a model relying on nuclear systematics. The characteristics of each code are described. There are orders of magnitude differences in the time for each type of code to calculate neutron emission spectra, with codes using systematics, preequilibrium and intranuclear-cascade models requiring seconds, minutes and hours, respectively. Calculations are not compared with experiment in this initial study. For double differential neutron emission spectra, there is good overall agreement in magnitude among the different types of codes at forward angles. Differences where they occur at forward angles are greatest for the mid-energy neutrons emitted. At back angles the incident energy at which the best overall agreement is obtained is 160 MeV and the material for which the best overall agreement is obtained is Al. 4 refs., 7 tabs

  9. 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

  10. Office of Codes and Standards resource book. Section 1, Building energy codes and standards

    Energy Technology Data Exchange (ETDEWEB)

    Hattrup, M.P.

    1995-01-01

    The US Department of Energy`s (DOE`s) Office of Codes and Standards has developed this Resource Book to provide: A discussion of DOE involvement in building codes and standards; a current and accurate set of descriptions of residential, commercial, and Federal building codes and standards; information on State contacts, State code status, State building construction unit volume, and State needs; and a list of stakeholders in the building energy codes and standards arena. The Resource Book is considered an evolving document and will be updated occasionally. Users are requested to submit additional data (e.g., more current, widely accepted, and/or documented data) and suggested changes to the address listed below. Please provide sources for all data provided.

  11. Generation of neutron cross sections library for the Thermos code of the Fuel management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Viais J, J.

    1990-10-01

    There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)

  12. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  13. Elaboration of data and documents intended to complement and expand the German series of nuclear engineering codes. 4. Technical report. Current knowledge of the mechanisms of corrosion fatigue in ferritic materials

    International Nuclear Information System (INIS)

    Frank, J.

    1997-01-01

    Concerning the processes of crack initiation under cyclic loading in a high-temperature water environment, examinations are performed in the USA for amending the design-basis curves published in ASME section III, which are comparable to those contained in the KTA code 3201.2. A revision of ASME section III has not yet been published. The same applies to the processes of crack propagation under cyclic loading in a high-temperature water environment, US experts examining for the purpose of amendment the limiting curves of cyclic crack propagation rates of carbon and low-alloyed steels in a water environment published in ASME section XI. (Orig./CB) [de

  14. NSLINK, Coupling of NJOY Cross-Sections Generator Code to SCALE-3 System

    International Nuclear Information System (INIS)

    De Leege, P.F.A

    1991-01-01

    1 - Description of program or function: NSLINK (NJOY - SCALE - LINK) is a set of computer codes to couple the NJOY cross-section generation code to the SCALE-3 code system (using AMPX-2 master library format) retaining the Nordheim resolved resonance treatment option. 2 - Method of solution: The following module and codes are included in NSLINK: XLACSR: This module is a stripped-down version of the XLACS-2 code. The module passes all l=0 resonance parameters as well as the contribution from all other resonances to the group cross-sections, the contribution from the wings of the l=0 resonances, the background cross-section and possible interference for multilevel Breit-Wigner resonance parameters. The group cross-sections are stored in the appropriate 1-D cross-section arrays. The output file has AMPX-2 master format. The original NJOY code is used to calculate all other data. The XLACSR module is included in the NJOY code. MILER: This code converts NJOY output (GENDF format) to AMPX-2 master format. The code is an extensively revised version of the original MILER code. In addition, the treatment of thermal scattering matrices at different temperatures is included. UNITABR: This code is a revised version of the UNITAB code. It merges the output of XLACSR and MILER in such a way that contributions from the bodies of the l=0 resonances in the resolved energy range, calculated by XLACSR, are subtracted from the 1-D group cross-section arrays for fission (MT=18) and neutron capture (MT=102). The l=0 resonance parameters and the contributions from the bodies of these resonances are added separately (MT=1023, 1022 and 1021). The total cross-section (MT=1), the absorption cross- section (MT=27) and the neutron removal cross-section (MT=101) values are adjusted. In the case of Bondarenko data, infinite dilution values of the cross-sections (MT=1, 18 and 102) are changed in the same way as the 1-D cross-section. The output file of UNITABR is in AMPX-2 master format and

  15. 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

  16. 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

  17. Process of cross section generation for radiation shielding calculations, using the NJOY code

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1986-10-01

    The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

  18. RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Anaf, J.

    1988-02-01

    An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab

  19. HADES. A computer code for fast neutron cross section from the Optical Model

    International Nuclear Information System (INIS)

    Guasp, J.; Navarro, C.

    1973-01-01

    A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs

  20. Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

    International Nuclear Information System (INIS)

    Dunn, M.E.; Greene, N.M.; Leal, L.C.

    1999-01-01

    Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

  1. Validation of the WIMSD4M cross-section generation code with benchmark results

    International Nuclear Information System (INIS)

    Deen, J.R.; Woodruff, W.L.; Leal, L.E.

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

  2. Validation of the WIMSD4M cross-section generation code with benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

  3. 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)

  4. MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-11-08

    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

  5. Design evaluation on sodium piping system and comparison of the design codes

    International Nuclear Information System (INIS)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon

    2015-01-01

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  6. Design evaluation on sodium piping system and comparison of the design codes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won; Jeong, Ji Young; Lee, Yong Bum; Lee, Hyeong Yeon [KAERI, Daejeon (Korea, Republic of)

    2015-03-15

    A large-scale sodium test loop of STELLA-1 (Sodium integral effect test loop for safety simulation and assessment) with two main piping systems has been installed at KAERI. In this study, design evaluations on the main sodium piping systems in STELLA-1 have been conducted according to the DBR (design by rule) codes of the ASME B31.1 and RCC-MRx RB-3600. In addition, design evaluations according to the DBA (design by analysis) code of the ASME Section III Subsection NB-3200 have been conducted. The evaluation results for the present piping systems showed that results from the DBR codes were more conservative than those from the DBA code, and among the DBR codes, the non-nuclear code of the ASME B31.1 was more conservative than the French nuclear DBR code of the RCC-MRx RB-3600. The conservatism on the DBR codes of the ASME B31.1 and RCC-MRx RB-3600 was quantified based on the present sodium piping analyses.

  7. Report on FY15 alloy 617 code rules development

    Energy Technology Data Exchange (ETDEWEB)

    Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jetter, Robert I [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hollinger, Greg [Becht Engineering Co., Inc., Liberty Corner, NJ (United States); Pease, Derrick [Becht Engineering Co., Inc., Liberty Corner, NJ (United States); Carter, Peter [Stress Engineering Services, Inc., Houston, TX (United States); Pu, Chao [Univ. of Tennessee, Knoxville, TN (United States); Wang, Yanli [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Due to its strength at very high temperatures, up to 950°C (1742°F), Alloy 617 is the reference construction material for structural components that operate at or near the outlet temperature of the very high temperature gas-cooled reactors. However, the current rules in the ASME Section III, Division 5 Subsection HB, Subpart B for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 650°C (1200°F) (Corum and Brass, Proceedings of ASME 1991 Pressure Vessels and Piping Conference, PVP-Vol. 215, p.147, ASME, NY, 1991). The rationale for this exclusion is that at higher temperatures it is not feasible to decouple plasticity and creep, which is the basis for the current simplified rules. This temperature, 650°C (1200°F), is well below the temperature range of interest for this material for the high temperature gas-cooled reactors and the very high temperature gas-cooled reactors. The only current alternative is, thus, a full inelastic analysis requiring sophisticated material models that have not yet been formulated and verified. To address these issues, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (EPP) analysis methods applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature (Carter, Jetter and Sham, Proceedings of ASME 2012 Pressure Vessels and Piping Conference, papers PVP 2012 28082 and PVP 2012 28083, ASME, NY, 2012), and have been recently revised to incorporate comments and simplify their application. Background documents have been developed for these two code cases to support the ASME Code committee approval process. These background documents for the EPP strain limits and creep-fatigue code cases are documented in this report.

  8. 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

  9. Modification in the CITATION computer code: change of microscopic cross sections by zone

    International Nuclear Information System (INIS)

    Yamaguchi, M.; Kosaka, N.

    1983-01-01

    Some modifications done in the CITATION computer code are presented, aiming to calculate the accumulated burnup for each reactor zone in each step of burnup and allow changing the microscopic cross sections for each zone in accordance to the burnup accumulated after each step of burnup. Some input data were put in the computer code. The alterations were tested and the results were compared with and without modifications. (E.G.) [pt

  10. GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra

    International Nuclear Information System (INIS)

    Young, P.G.; Arthur, E.D.

    1977-11-01

    A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables

  11. PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Tsuneo; Sugi, Teruo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iijima, Shungo; Nishigori, Takeo

    1999-06-01

    The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,{gamma}), (n,n`), (n,p), (n,{alpha}), (n,d), (n,t), (n,{sup 3}He), (n,2n), (n,n`p), (n,n`{alpha}), (n,n`d), (n,n`t), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)

  12. A computer code for calculating neutron cross-sections from resonance parameter data

    International Nuclear Information System (INIS)

    Mill, A.J.

    1979-08-01

    A computer code, XSEC, has been written which calculates neutron cross-sections from resonance data. Although the program was originally written in order to identify neutron 'windows' in enriched nuclides, it may be used to evaluate the total neutron cross-section of any medium mass nuclide at intermediate energies. XSEC has proved very useful in identifying suitable nuclides for use as neutron filters at intermediate energies. (author)

  13. Generation of cross-sections and reference solutions using the code Serpent

    International Nuclear Information System (INIS)

    Gomez T, A. M.; Delfin L, A.; Del Valle G, E.

    2012-10-01

    Serpent is a code that solves the neutron transport equations using the Monte Carlo method that besides generating reference solutions in stationary state for complex geometry problems, has been specially designed for physical applications of cells, what includes the generation of homogenized cross-sections for several energy groups. In this work a calculation methodology is described using the code Serpent to generate the necessary cross-sections to carry out calculations with the code TNXY, developed in 1993 in the Nuclear Engineering Department of the Instituto Politecnico Nacional (Mexico) by means of an interface programmed in Octave. The computation program TNXY solves the neutron transport equations for several energy groups in stationary state and geometry X Y using the Discreet Ordinates method (S N ). To verify and to validate the methodology the results of TNXY were compared with those calculated by Serpent giving minor differences to 0.55% in the value of the multiplication factor. (Author)

  14. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    International Nuclear Information System (INIS)

    Brenner, D.J.; Prael, R.E.; Little, R.C.

    1987-01-01

    Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

  15. On the use of the Serpent Monte Carlo code for few-group cross section generation

    International Nuclear Information System (INIS)

    Fridman, E.; Leppaenen, J.

    2011-01-01

    Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and

  16. ASME codification of ductile cast iron cask for transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Saegusa, Toshiari; Arai, Taku

    2012-01-01

    The CRIEPI has been executing research and development on ductile cast iron cask for transport and storage of spent nuclear fuel in order to diversify options of the casks. Based on the research results, the CRIEPI proposed materials standards (Section II) and structural design standards (Section III) for the ductile cast iron cask to the authoritative and international ASME (American Society of Mechanical Engineers) Codes. For the Section II, the CRIEPI proposed the JIS G 5504 material with additional requirement prohibiting repair of cast body by welding, etc. as well as the ASTM A874 material to the Part A. In addition, the CRIEPI proposed design stress allowables, physical properties (thermal conductivity, modulus of elasticity, etc.), and external pressure chart to the Part D. For the Section III, the CRIEPI proposed a fracture toughness requirement of the ductile cast iron cask at -40degC to WB and WC of Division 3. Additionally, the CRIEPI proposed a design fatigue curve of the ductile cast iron cask to Appendix of Division 1. This report describes the outline of the proposed standards, their bases, and the deliberation process in order to promote proper usage of the code, future improvement, etc. (author)

  17. PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping

    International Nuclear Information System (INIS)

    Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.

    1975-01-01

    To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report

  18. SENSIT: a cross-section and design sensitivity and uncertainty analysis code

    International Nuclear Information System (INIS)

    Gerstl, S.A.W.

    1980-01-01

    SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE

  19. Aquelarre. A computer code for fast neutron cross sections from the statistical model

    International Nuclear Information System (INIS)

    Guasp, J.

    1974-01-01

    A Fortran V computer code for Univac 1108/6 using the partial statistical (or compound nucleus) model is described. The code calculates fast neutron cross sections for the (n, n'), (n, p), (n, d) and (n, α reactions and the angular distributions and Legendre moments.for the (n, n) and (n, n') processes in heavy and intermediate spherical nuclei. A local Optical Model with spin-orbit interaction for each level is employed, allowing for the width fluctuation and Moldauer corrections, as well as the inclusion of discrete and continuous levels. (Author) 67 refs

  20. Code boiler and pressure vessel life assessment

    International Nuclear Information System (INIS)

    Farr, J.R.

    1992-01-01

    In the United States of America and in Canada, laws and controls for determining life assessment for continued operation of equipment exist only for those pressure vessels built to Section III and evaluated according to Section XI. In this presentation, some of those considerations which are made in the USA and Canada for deciding on life or condition assessment of boilers and pressure vessels designed and constructed to other sections of the ASME Boiler and Pressure Vessel Code are reviewed. Life assessment or condition assesssment is essential in determining what is necessary for continued operation. With no ASME rules being adopted by laws or regulations, other than OSHA in the USA and similar environmental controls in Canada, to control life assessment for continued operation, the equipment owner must decide if assessment is to be done and how much to do. Some of those considerations are reviewed along with methods and procedures to make an assessment along with a discussion of where the ASME B and PV Code currently stands regarding continued operation. (orig.)

  1. Dynamic Model for the Z Accelerator Vacuum Section Based on Transmission Line Code%Dynamic Model for the Z Accelerator Vacuum Section Based on Transmission Line Code

    Institute of Scientific and Technical Information of China (English)

    呼义翔; 雷天时; 吴撼宇; 郭宁; 韩娟娟; 邱爱慈; 王亮平; 黄涛; 丛培天; 张信军; 李岩; 曾正中; 孙铁平

    2011-01-01

    The transmission-line-circuit model of the Z accelerator, developed originally by W. A. STYGAR, P. A. CORCORAN, et al., is revised. The revised model uses different calculations for the electron loss and flow impedance in the magnetically insulated transmission line system of the Z accelerator before and after magnetic insulation is established. By including electron pressure and zero electric field at the cathode, a closed set of equations is obtained at each time step, and dynamic shunt resistance (used to represent any electron loss to the anode) and flow impedance are solved, which have been incorporated into the transmission line code for simulations of the vacuum section in the Z accelerator. Finally, the results are discussed in comparison with earlier findings to show the effectiveness and limitations of the model.

  2. A wide-range model of two-group gross sections in the dynamics code HEXTRAN

    International Nuclear Information System (INIS)

    Kaloinen, E.; Peltonen, J.

    2002-01-01

    In dynamic analyses the thermal hydraulic conditions within the reactor core may have a large variation, which sets a special requirement on the modeling of cross sections. The standard model in the dynamics code HEXTRAN is the same as in the static design code HEXBU-3D/MODS. It is based on a linear and second order fitting of two-group cross sections on fuel and moderator temperature, moderator density and boron density. A new, wide-range model of cross sections developed in Fortum Nuclear Services for HEXBU-3D/MOD6 has been included as an option into HEXTRAN. In this model the nodal cross sections are constructed from seven state variables in a polynomial of more than 40 terms. Coefficients of the polynomial are created by a least squares fitting to the results of a large number of fuel assembly calculations. Depending on the choice of state variables for the spectrum calculations, the new cross section model is capable to cover local conditions from cold zero power to boiling at full power. The 5. dynamic benchmark problem of AER is analyzed with the new option and results are compared to calculations with the standard model of cross sections in HEXTRAN (Authors)

  3. 76 FR 37034 - Certain Employee Remuneration in Excess of $1,000,000 Under Internal Revenue Code Section 162(m)

    Science.gov (United States)

    2011-06-24

    ... Certain Employee Remuneration in Excess of $1,000,000 Under Internal Revenue Code Section 162(m) AGENCY... remuneration in excess of $1,000,000 under the Internal Revenue Code (Code). The proposed regulations clarify... stock options, it is intended that the directors may retain discretion as to the exact number of options...

  4. Summary of design of nuclear vessels and piping to ASME III (NB, NC, ND) and vessels to BS 5500

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1992-01-01

    There is a hierarchy of design code requirements for pressurised components, starting with non-nuclear codes as the minimum and progressing through the ASME III nuclear Classes 3, 2, 1. In establishing and assessing the safety justifications of nuclear plants it is important to have an appreciation of the gradation of requirements in the ASME III design rules and how these go beyond non-nuclear component design rules. There are two broad aspects to the structural integrity of pressurised components, namely the achievement of integrity and the demonstration of integrity. The technical requirements of design codes are associated with achieving integrity while the documentary aspects are usually associated with demonstrating integrity. In practice documents also have a part in achieving integrity in the communication of information between different organisations and personnel involved in the design process. It is not possible to assign simple numerical measures to the relative integrity afforded by non-nuclear codes and the three Classes of ASME III. Instead it is necessary to compare the different requirements of the rules for the various technical and documentary aspects. This paper summarises the most important technical and documentary aspects of the three Classes of the ASME III Code for vessels and the non-nuclear code BS 5500. A similar summary is also provided for the three Classes of ASME III rules for piping. The intention is that the paper provides a basis for appreciating the relative integrity afforded by these various rules. (author)

  5. Comparison of Neutron Cross-Sections Using IAEA Nuclear Codes ''ABAREX'' and ''SCAT2''

    International Nuclear Information System (INIS)

    Myint Myint Moe; Win Sin; Sein Htoon

    2004-05-01

    Moel calculations can be used to provide nuclear data for applications in science and technology. The energy averaged neutron induced nuclear reaction cross-sections particular for Al-27, Mg-24, Cr-52, Mn-55, Zn-64 and U-238 with neutrons of energy (0.005 to 10 MeV) are calculated using IAEA nuclear codes ''ABAREX'' and ''SCAT2''. The results are compared with those given in ENDF- 3 nuclear data

  6. STEEP4 code for computation of specific thermonuclear reaction rates from pointwise cross sections

    International Nuclear Information System (INIS)

    Harris, D.R.; Dei, D.E.; Husseiny, A.A.; Sabri, Z.A.; Hale, G.M.

    1976-05-01

    A code module, STEEP4, is developed to calculate the fusion reaction rates in terms of the specific reactivity [sigma v] which is the product of cross section and relative velocity averaged over the actual ion distributions of the interacting particles in the plasma. The module is structured in a way suitable for incorporation in thermonuclear burn codes to provide rapid and yet relatively accurate on-line computation of [sigma v] as a function of plasma parameters. Ion distributions are modified to include slowing-down contributions which are characterized in terms of plasma parameters. Rapid and accurate algorithms are used for integrating [sigma v] from cross sections and spectra. The main program solves for [sigma v] by the method of steepest descent. However, options are provided to use Gauss-Hermite and dense trapezoidal quadrature integration techniques. Options are also provided for rapid calculation of screening effects on specific reaction rates. Although such effects are not significant in cases of plasmas of laboratory interest, the options are included to increase the range of applicability of the code. Gamow penetration form, log-log interpolation, and cubic interpolation routines are included to provide the interpolated values of cross sections

  7. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-11-01

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  8. 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.

  9. 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

  10. 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

  11. RSAP - A Code for Display of Neutron Cross Section Data and SAMMY Fit Results

    International Nuclear Information System (INIS)

    Sayer, R.O.

    2001-01-01

    RSAP is a computer code for display of neutron cross section data and selected SAMMY output. SAMMY is a multilevel R-matrix code for fitting neutron time-of-flight cross-section data using Bayes' method. RSAP, which runs on the Digital Unix Alpha platform, reads ORELA Data Files (ODF) created by SAMMY and uses graphics routines from the PLPLOT package. In addition, RSAP can read data and/or computed values from ASCII files with a format specified by the user. Plot output may be displayed in an X window, sent to a postscript file (rsap.ps), or sent to a color postscript file (rsap.psc). Thirteen plot types are supported, allowing the user to display cross section data, transmission data, errors, theory, Bayes fits, and residuals in various combinations. In this document the designations theory and Bayes refer to the initial and final theoretical cross sections, respectively, as evaluated by SAMMY. Special plot types include Bayes/Data, Theory--Data, and Bayes--Data. Output from two SAMMY runs may be compared by plotting the ratios Theory2/Theory1 and Bayes2/Bayes1 or by plotting the differences (Theory2-Theory1) and (Bayes2-Bayes1)

  12. Section 525(a) of the bankruptcy code plainly does not apply to Medicare provider agreements.

    Science.gov (United States)

    Sperow, E H

    2001-01-01

    Section 525(a) of the Bankruptcy Code prevents government entities from discriminating against debtors based on the debtor's bankruptcy filing. This Article analyzes how this provision is applied to healthcare providers who file for bankruptcy. Some commentators have expressed concerns that because of Section 525, the federal government is unable to deny a bankrupt provider a new Medicare provider agreement due to the debtor's failure to pay debts discharged during bankruptcy. This Article, however, argues that Section 525 does not apply to a provider agreements because it is not a "license, permit, charter, franchise, or other similar grant" as defined by the statute. Therefore, the author concludes that debtor healthcare providers should not be allowed back into the Medicare program without first paying their statutorily required debts.

  13. PHOBINS: an index file of photon production cross section data and its utility code system

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Koyama, Kinji; Ido, Masaru; Hotta, Masakazu; Miyasaka, Shun-ichi

    1978-08-01

    The code System PHOBINS developed for reference of photon production cross sections is described in detail. The system is intended to grasp the present status of photon production data and present the information of available data. It consists of four utility routines, CREA, UP-DT, REF and BACK, and data files. These utility routines are used for making an index file of the photon production cross sections, updating the index file, searching the index file and producing a back-up file of the index file. In the index file of the photon production cross sections, a data base system is employed for efficient data management in economical storage, ease of updating and efficient reference. The present report is a reference manual of PHOBINS. (author)

  14. SCATLAW: a code of scattering law and cross sections calculation for liquids and solids

    International Nuclear Information System (INIS)

    Padureanu, I.; Rapeanu, S.; Rotarascu, G.; Craciun, C.

    1978-11-01

    A code for calculation of the scattering law S(Q,ω), differential and double differential cross sections and scattering kernels in the energy range E(0 - 683 meV) and wave-vector transfer Q(0 - 40 A -1 ) is presented. The code can be used both for solids and liquids which are coherent or incoherent scatterer. For liquids the calculations are based on the most recent theoretical models involving the correlation functions and generalized field approach. The phonon expansion model and the free gas model are also analysed in term of frequency spectra obtained from inelastic neutron scattering using time-of-flight technique. Several results on liquid sodium at T = 233 deg C and on liquid bismuth at T = 286 deg C and T = 402 deg C are presented. (author)

  15. One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

    1988-11-01

    We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

  16. Applications of American design codes for elevated temperature environment

    International Nuclear Information System (INIS)

    Severud, L.K.

    1980-03-01

    A brief summary of the ASME Code rules of Case N-47 is presented. An overview of the typical procedure used to demonstrate Code compliance is provided. Application experience and some examples of detailed inelastic analysis and simplified-approximate methods are given. Recent developments and future trends in design criteria and ASME Code rules are also presented

  17. EMPIRE-II 2.18, Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections

    International Nuclear Information System (INIS)

    Herman, Michal Wladyslaw; Panini, Gian Carlo

    2003-01-01

    1 - Description of program or function: EMPIRE-II is a flexible code for calculation of nuclear reactions in the frame of combined optical, Multi-step Direct (TUL), Multi-step Compound (NVWY) and statistical (Hauser-Feshbach) models. Incident particle can be a nucleon or any nucleus(Heavy Ion). Isomer ratios, residue production cross sections and emission spectra for neutrons, protons, alpha-particles, gamma-rays, and one type of Light Ion can be calculated. The energy range starts just above the resonance region for neutron induced reactions and extends up to several hundreds of MeV for the Heavy Ion induced reactions. IAEA1169/06: This version corrects an error in the Absoft compile procedure. 2 - Method of solution: For projectiles with A<5 EMPIRE calculates fusion cross section using spherical optical model transmission coefficients. In the case of Heavy Ion induced reactions the fusion cross section can be determined using various approaches including simplified coupled channels method (code CCFUS). Pre-equilibrium emission is treated in terms of quantum-mechanical theories (TUL-MSD and NVWY-MSC). MSC contribution to the gamma emission is taken into account. These calculations are followed by statistical decay with arbitrary number of subsequent particle emissions. Gamma-ray competition is considered in detail for every decaying compound nucleus. Different options for level densities are available including dynamical approach with collective effects taken into account. EMPIRE contains following third party codes converted into subroutines: - SCAT2 by O. Bersillon, - ORION and TRISTAN by H. Lenske and H. Wolter, - CCFUS by C.H. Dasso and S. Landowne, - BARMOM by A. Sierk. 3 - Restrictions on the complexity of the problem: The code can be easily adjusted to the problem by changing dimensions in the dimensions.h file. The actual limits are set by the available memory. In the current formulation up to 4 ejectiles plus gamma are allowed. This limit can be relaxed

  18. 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....

  19. Code implementation of partial-range angular scattering cross sections: GAMMER and MORSE

    International Nuclear Information System (INIS)

    Ward, J.T. Jr.

    1978-01-01

    A partial-range (finite-element) method has been previously developed for representing multigroup angular scattering in Monte Carlo photon transport. Computer application of the method, with preliminary quantitative results is discussed here. A multigroup photon cross section processing code, GAMMER, was written which utilized ENDF File 23 point data and the Klein--Nishina formula for Compton scattering. The cross section module of MORSE, along with several execution routines, were rewritten to permit use of the method with photon transport. Both conventional and partial-range techniques were applied for comparison to calculating angular and spectral penetration of 6-MeV photons through a six-inch iron slab. GAMMER was found to run 90% faster than SMUG, with further improvement evident for multiple-media situations; MORSE cross section storage was reduced by one-third; cross section processing, greatly simplified; and execution time, reduced by 15%. Particle penetration was clearly more forward peaked, as moment accuracy is retained to extremly high order. This method of cross section treatment offers potential savings in both storage and handling, as well as improved accuracy and running time in the actual execution phase. 3 figures, 4 tables

  20. MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

    International Nuclear Information System (INIS)

    Henryson, H. II; Toppel, B.J.; Stenberg, C.G.

    1976-06-01

    MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers

  1. Up to date cross sections library for Thermos and Record codes

    International Nuclear Information System (INIS)

    Hernandez Lopez, H.

    1993-01-01

    Reactor cell analysis is the first step in determining reactor core behavior and is required in the reload licensing process. For best results, reactor cell analysis should be carried out with libraries of up to date, accurate cross sections produced with well described methods from standard evaluated nuclear data. At first step in this work were determined the library structure for RECORD and THERMOS and were prepared the cross sections libraries using the NJOY nuclear data processing system and the ENDF-B/IV evaluated nuclear data. These libraries were used by the codes and some samples were perform, the result show some differences against the results obtained using the previous libraries. By other hand the libraries contain various adjustments to correct for deficiencies in nuclear data or analytical methods. These adjustments doesn't have any documentation, although some of them were identified in this work. (Author). 25 refs, 78 figs, 55 tabs

  2. 12G: code for conversion of isotope-ordered cross-section libraries into group-ordered cross-section libraries

    International Nuclear Information System (INIS)

    Resnik, W.M. II; Bosler, G.E.

    1977-09-01

    Many current reactor physics codes accept cross-section libraries in an isotope-ordered form, convert them with internal preprocessing routines to a group-ordered form, and then perform calculations using these group-ordered data. Occasionally, because of storage and time limitations, the preprocessing routines in these codes cannot convert very large multigroup isotope-ordered libraries. For this reason, the I2G code, i.e., ISOTXS to GRUPXS, was written to convert externally isotope-ordered cross section libraries in the standard file format called ISOTXS to group-ordered libraries in the standard format called GRUPXS. This code uses standardized multilevel data management routines which establish a strategy for the efficient conversion of large libraries. The I2G code is exportable contingent on access to, and an intimate familiarization with, the multilevel routines. These routines are machine dependent, and therefore must be provided by the importing facility. 6 figures, 3 tables

  3. Validation of the XLACS code related to contribution of resolved and unresolved resonances and background cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1990-01-01

    The procedures for calculating contributions of resolved and unresolved resonances and background cross sections, in XLACS code, were revised. Constant weighting function and zero Kelvin temperature were considered. Discrepancies found were corrected and now the validated XLACS code generates results that are correct and in accordance with its originally established procedures. (author)

  4. Storage Tanks - Selection Of Type, Design Code And Tank Sizing

    International Nuclear Information System (INIS)

    Shatla, M.N; El Hady, M.

    2004-01-01

    The present work gives an insight into the proper selection of type, design code and sizing of storage tanks used in the Petroleum and Process industries. In this work, storage tanks are classified based on their design conditions. Suitable design codes and their limitations are discussed for each tank type. The option of storage under high pressure and ambient temperature, in spherical and cigar tanks, is compared to the option of storage under low temperature and slight pressure (close to ambient) in low temperature and cryogenic tanks. The discussion is extended to the types of low temperature and cryogenic tanks and recommendations are given to select their types. A study of pressurized tanks designed according to ASME code, conducted in the present work, reveals that tanks designed according to ASME Section VIII DIV 2 provides cost savings over tanks designed according to ASME Section VIII DlV 1. The present work is extended to discuss the parameters that affect sizing of flat bottom cylindrical tanks. The analysis shows the effect of height-to-diameter ratio on tank instability and foundation loads

  5. Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

    International Nuclear Information System (INIS)

    Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.

    2007-01-01

    Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr

  6. New evaluated neutron cross section libraries for the GEANT4 code

    International Nuclear Information System (INIS)

    Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.

    2012-04-01

    The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)

  7. Development of a code FITWR for nuclear cross section statistical analysis

    International Nuclear Information System (INIS)

    Alrwashdeh, Mohammad; Kan, Wang

    2014-01-01

    Highlights: • We used the weighted least square with nonlinear regression method to fit experimental nuclear data. • The FITWR code has been successful applied for both light and heavy nuclei with many resonance points. • More improvements will be applied in the future, by including a new methods for nuclear data fitting. - Abstract: A computer program named FITWR has been developed and applied to the experimental total cross sections for MEV incident energy particles such as neutron and proton. The computer program FITWR adapted the weighted least square method with weighted mathematical models with nonlinear regression applied to high order fitting polynomial, in order to meet the growing demands of the experimental nuclear data. The computer program FITWR deals with variance and covariance data provided along with experimental data and yields those for the evaluated ones

  8. 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

  9. 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.

  10. 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

  11. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  12. Interpreting ASME limits and philosophy in FEA of pressure vessel parts

    International Nuclear Information System (INIS)

    Bezerra, L.M.; Cruz, J.R.B.; Miranda, C.A.J.; Neto, M.M.

    1995-01-01

    In recent years there has been an effort to interpret finite element (FE) stress results on the light of the ASME B and PV rules and philosophy. Many task groups have issued guidelines on stress linearization and classifications. All those attempts have come up trying to cope modern FE techniques with the rules imposed by the ASME Code. This paper is an independent contribution to the Pressure Vessel Research Council (PVRC) groups which are studying the stress classification and the failure mechanism in a FE framework. This work tries to complement the interesting work by Hollinger and Hechmer presented in the PVP-94 in Minneapolis. In that paper, the authors examined a typical support skirt and showed relations between the skirt collapse load obtained by finite element analysis and the loads allowed from the ASME stress limits. To complement such paper, in the present article, different skirt geometry configurations are analyzed. The configurations here investigated consist of similar support skirts but with different angles of attachments between cylinder and cone parts. It will be possible to observe the influence of the bending stress in the collapse load and its relation to the allowable loads inferred from the ASME limits. A pressure vessel with torispherical head under internal pressure is also examined. Using elastic and limit load FEA, the present paper determines the collapse loads of the configurations. It sets up the relations between these collapse loads, stress categories, and limits dictated by the ASME Code Subsection NB. On the light of NB rules and philosophy, this paper shows how different methods of stress assessment, classification, and limits may influence in the design of a pressure vessel

  13. Comparison of the ASME Environmental Fatigue Design Curve with the Leax' Low Bound Model

    International Nuclear Information System (INIS)

    Jeong, Ill Seok; Kim, Wan Jae; Jun, Hyun Ik

    2010-01-01

    Environmental fatigue issue long time argued between industry and regulator. The issues of the debates are about environmental fatigue data only from experiment laboratories, no evidences in fields, and over conservatism. However, NRC issued the requirement to implement it to the construction design prior to industry practical design code. American Society of Mechanical Engineers (ASME) determined to issue non-mandatory code cases of environmental fatigue design. This paper evaluated the conservatism of the ASME proposed environmental fatigue design curve in comparison with the Leax' low bound approach model of environmental fatigue curve. A group of CF8M cast austenitic stainless steel (CASS) produced in KEPCO Research Center was introduced in the evaluation

  14. 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

  15. Design criteria and pressure vessel codes - an American view

    International Nuclear Information System (INIS)

    Tuppeny, W.H.

    1975-01-01

    To the pressure vessel designer, codes and criteria represent the common ground where the stress analyst and the metallurgist must interact and evolve rules and procedures which will ensure safety and open-ended responsiveness to technological, economic, and environmental change. The paper briefly discusses the evolution and rationale behind the current ASME code sections -emphasizing those portions applicable to designs operating in the creep range. The author then proposes a plan of action so that the analysts and materials people can make optimum use of time and resources, and evolve data and design criteria which will be responsive to changing technology and the economic and safety requirements of the future. (author)

  16. Library of neutron cross sections of the Thermos code; Biblioteca de secciones eficaces de neutrones del codigo Thermos

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-10-15

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  17. A fortran code CVTRAN to provide cross-section file for TWODANT by using macroscopic file written by SRAC

    International Nuclear Information System (INIS)

    Yamane, Tsuyoshi; Tsuchihashi, Keichiro

    1999-03-01

    A code CVTRAN provides the macroscopic cross-sections in the format of XSLIB file which is one of Standard interface files for a two-dimensional Sn transport code TWODANT by reading a macroscopic cross section file in the PDS format which is prepared by SRAC execution. While a two-dimensional Sn transport code TWOTRAN published by LANL is installed as a module in the SRAC code system, several functions such as alpha search, concentration search, zone thickness search and various edits are suppressed. Since the TWODANT code was released from LANL, its short running time, stable convergence and plenty of edits have attracted many users. The code CVTRAN makes the TWODANT available to the SRAC user by providing the macroscopic cross-sections on a card-image file XSLIB. The CVTRAN also provides material dependent fission spectra into a card-image format file CVLIB, together with group velocities, group boundary energies and material names. The user can feed them into the TWODANT input, if necessary, by cut-and-paste command. (author)

  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. ORLIB: a computer code that produces one-energy group, time- and spatially-averaged neutron cross sections

    International Nuclear Information System (INIS)

    Blink, J.A.; Dye, R.E.; Kimlinger, J.R.

    1981-12-01

    Calculation of neutron activation of proposed fusion reactors requires a library of neutron-activation cross sections. One such library is ACTL, which is being updated and expanded by Howerton. If the energy-dependent neutron flux is also known as a function of location and time, the buildup and decay of activation products can be calculated. In practice, hand calculation is impractical without energy-averaged cross sections because of the large number of energy groups. A widely used activation computer code, ORIGEN2, also requires energy-averaged cross sections. Accordingly, we wrote the ORLIB code to collapse the ACTL library, using the flux as a weighting function. The ORLIB code runs on the LLNL Cray computer network. We have also modified ORIGEN2 to accept the expanded activation libraries produced by ORLIB

  20. New Standard Evaluated Neutron Cross Section Libraries for the GEANT4 Code and First Verification

    CERN Document Server

    Mendoza, Emilio; Koi, Tatsumi; Guerrero, Carlos

    2014-01-01

    The Monte Carlo simulation of the interaction of neutrons with matter relies on evaluated nuclear data libraries and models. The evaluated libraries are compilations of measured physical parameters (such as cross sections) combined with predictions of nuclear model calculations which have been adjusted to reproduce the experimental data. The results obtained from the simulations depend largely on the accuracy of the underlying nuclear data used, and thus it is important to have access to the nuclear data libraries available, either of general use or compiled for specific applications, and to perform exhaustive validations which cover the wide scope of application of the simulation code. In this paper we describe the work performed in order to extend the capabilities of the GEANT4 toolkit for the simulation of the interaction of neutrons with matter at neutron energies up to 20 MeV and a first verification of the results obtained. Such a work is of relevance for applications as diverse as the simulation of a n...

  1. Benchmark test of MORSE-DD code using double-differential form cross sections

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa; Ishiguro, Yukio

    1985-02-01

    The multi-group double-differential form cross sections (DDX) and the three dimensional Monte Carlo code MORSE-DD devised to utilize the DDX, which were developed for the fusion neutronics analysis, have been validated through many benchmark tests. All the problems tested have a 14 MeV neutron source. To compare the calculated results with the measured values, the following experiments were adopted as the benchmark problems; leakage neutron spectra from spheres composed of nine kinds of materials measured at LLNL, neutron angular spectra from the Li 2 O slab measured at FNS in JAERI, tritium production rate (TPR) in the graphite-reflected Li 2 O sphere measured at FNS and the TPR in the metallic Li sphere measured at KfK. In addition in order to test an accuracy of the calculation method in detail, spectra of neutrons scattered from a small sample and various reaction rates in a Li 2 O cylinder were compared between the present method and the continuous energy Monte Carlo method. The nuclear data files used are mainly ENDF/B4 and partly JENDL-3PR1. The tests were carried out through a comparison with the measured values and also with the results obtained from the conventional Legendre expansion method and the continuous energy Monte Carlo method. It is found that the results by the present method are more accurate than those by the conventional one and agree well with those by the continuous energy Monte Carlo calculations. Discrepancies due to the nuclear data are also discussed. (author)

  2. Manual phased arrays for weld inspections using North American codes

    International Nuclear Information System (INIS)

    Moles, Michael

    2008-01-01

    Phased arrays are primarily a method of generating and receiving ultrasound, not a new technology. In addition, the physics of ultrasound generated by phased arrays is identical to that from conventional monocrystals. Not surprisingly, all the major North American (and some European) codes accept phased arrays, either explicitly or implicitly. However, the technique and procedures needs to be proven, typically by a Performance Demonstration. The ASME (AmeicanSociety for Mechanical Engineers) Section V and API RP2X explicitly accept phased arrays. Three ASME code cases have been written specifically fo manual phased array: Code Cases 2541. 2557 and 2558. Over and above the general requirements of Article 4, these Code Cases require full waveform calibration. This is echoed in ASTM E-2491, a Standard Guide for setting up phased arrays. In addition. details such as focusing and reporting are addressed. The American Petroleum Institute QUTE procedure did not need any modifications to be compatible with manual phased arrays. The American Welding Society (AWS) Structural Welding Code D1.1 implicitly accepts phased arrays. New technologies such as phased arrays can be proven using Annex K. Nonetheless, a manual phased array unit using the standard AWS probe and displaying 45, 60 and 70degrees waveforms would be acceptable for D1.1 a s is . Overall, most major North American codes accept phased arrays, however, the technique and procedures must be proven, often using a Performance Demonstration. (author)

  3. Decree of 19 November 1981 listing the insanitary industries in accordance with Section 216 of the health Code

    International Nuclear Information System (INIS)

    1981-01-01

    This Decree issued by the Minister of Health approves an amended list of insanitary industries which are subject to certain obligations under Section 216 of the Health Code of 1934. The amendments concern certain nuclear plants and laboratories. The 1981 Decree modifies a previous Decree of 1976. (NEA) [fr

  4. Comparison of Fatigue crack growth rate of Type 347 stainless steel with ASME and JSME models

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Jeon, Soon-Hyeok; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, the FCGR of 347SS was evaluated in modified PWR high temperature water conditions. The FCGRs of 347SS under modified pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO), and it were compared with other models proposed by ASME and Japanese groups. Corrosion fatigue is main factor of environmental fatigue effect. Increase of DO level in water induced more corrosion damage, and it accelerated FCGR in PWR and FCGR of 347SS in PWR water condition was faster than reference curves in J-PWR and ASME draft code case derived by 304 and 316 stainless steel, but it was slower than J-BWR reference curve. Using J-BWR model for estimating the FCGR of 347SS under PWR might be conservative.

  5. Reliability based code calibration of fatigue design criteria of nuclear Class-1 piping

    International Nuclear Information System (INIS)

    Mishra, J.; Balasubramaniyan, V.; Chellapandi, P.

    2016-01-01

    Fatigue design of Class-l piping of NPP is carried out using Section-III of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel code. The fatigue design criteria of ASME are based on the concept of safety factor, which does not provide means for the management of uncertainties for consistently reliable and economical designs. In this regards, a work is taken up to estimate the implicit reliability level associated with fatigue design criteria of Class-l piping specified by ASME Section III, NB-3650. As ASME fatigue curve is not in the form of analytical expression, the reliability level of pipeline fittings and joints is evaluated using the mean fatigue curve developed by Argonne National Laboratory (ANL). The methodology employed for reliability evaluation is FORM, HORSM and MCS. The limit state function for fatigue damage is found to be sensitive to eight parameters, which are systematically modelled as stochastic variables during reliability estimation. In conclusion a number of important aspects related to reliability of various piping product and joints are discussed. A computational example illustrates the developed procedure for a typical pipeline. (author)

  6. 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...

  7. 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)

  8. 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.

  9. SENSIT: a cross-section and design sensitivity and uncertainty analysis code. [In FORTRAN for CDC-7600, IBM 360

    Energy Technology Data Exchange (ETDEWEB)

    Gerstl, S.A.W.

    1980-01-01

    SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE.

  10. Comparison of Computational Electromagnetic Codes for Prediction of Low-Frequency Radar Cross Section

    National Research Council Canada - National Science Library

    Lash, Paul C

    2006-01-01

    .... The goal of this research is to compare the capabilities of three computational electromagnetic codes for use in production of RCS signature assessments at low frequencies in terms of performance...

  11. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  12. HAUFES : a FORTRAN code for the calculation of compound nuclear cross-sections by Hauser-Feshbach theory

    International Nuclear Information System (INIS)

    Viyogi, Y.P.; Ganguly, N.K.

    1975-01-01

    The FORTRAN code described in the report has been developed for the BESM-6 computer with a view to calculate the cross-section of reactions proceeding via the formation of compound nucleus for all open two-body reaction channels using Hauser-Feshbach theory with Moldauer's correction for the fluctuation of level widths. The code can also be used to analyse data from 'crystal blocking' experiments to obtain nuclear level densities. The report describes the input-output specifications along with a short account of the algorithm of the program. (author)

  13. ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes

    International Nuclear Information System (INIS)

    Kodeli, Ivo; Aldama, Daniel L.; Leege, Piet F.A. de; Legrady, David; Hoogenboom, J. Eduard

    2007-01-01

    1 - Description: Format: MATXS and ACE; Number of groups: 175 neutron, 45 gamma-ray; Nuclides: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Mn-55, Fe-54, -56, -57, -58, I-127, W-nat. Origin: ENDF/B-VI.8; Weighting spectrum: Fission and fusion peak at high energies and a 1/E + thermal Maxwellian extension at low energies. The following materials/nuclides are included in the library: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Fe-54, -56, -57, -58, Mn-55, I-127, W-nat. ZZ-BOREHOLE-EB6.8-MG is a multigroup cross section library for deterministic (DOORS, DANTSYS) and Monte Carlo (MCNP) transport codes developed for the oil well logging applications. The library is based on the ENDF/B-VI.8 evaluation and was processed by the NJOY-99 code. The cross sections are given in the 175 neutron and 45 gamma ray group structure. The MATXS format library can be directly used in TRANSX code to prepare the multigroup self-shielded cross sections for deterministic discrete ordinates codes like DOORS and DANTSYS. The data provided in the GROUPR and GAMINR format were converted to the MCNP ACE format by the NSLINK, SCALE and CRSRD codes. IAEA1398/03: Multigroup cross section data for Mn-55 were added in TRANSX format

  14. 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)

  15. Organization of cross-section data in the Monte Carlo code SPARTAN

    International Nuclear Information System (INIS)

    Bending, R.C.

    1974-01-01

    The Monte Carlo code SPARTAN is a general-purpose code intended for neutron or gamma transport calculations. The code is designed to accept physics data from a number of external libraries (which may be used singly or in combination) and to use this data with as little alteration as possible. Data obtained from one or several libraries is placed in an interface file on magnetic tape or disk, using a general hierarchical structure which allows particular data items to be assessed in a straightforward way. The interface file, with or without additional data from cards, is regarded as a data source for the main Monte Carlo calculation. A summary of the functional forms, sampling distributions, and particle interaction laws which are available at present, and some of the mathematical methods used are described. 5 references. (U.S.)

  16. Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

    International Nuclear Information System (INIS)

    Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk

    1992-01-01

    A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)

  17. Development of a one-group cross section data base of the ORIGEN2 computer code for research reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Hwang, Won Guk [Kyung Hee University, Seoul (Korea, Republic of)

    1992-03-01

    A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author).

  18. MICROX-2: an improved two-region flux spectrum code for the efficient calculation of group cross sections

    International Nuclear Information System (INIS)

    Mathews, D.; Koch, P.

    1979-12-01

    The MICROX-2 code is an improved version of the MICROX code. The improvements allow MICROX-2 to be used for the efficient and rigorous preparation of broad group neutron cross sections for poorly moderated systems such as fast breeder reactors in addition to the well moderated thermal reactors for which MICROX was designed. MICROX-2 is an integral transport theory code which solves the neutron slowing down and thermalization equations on a detailed energy grid for two-region lattice cells. The fluxes in the two regions are coupled by transport corrected collision probabilities. The inner region may include two different types of grains (particles). Neutron leakage effects are treated by performing B 1 slowing down and P 0 plus DB 2 thermalization calculations in each region. Cell averaged diffusion coefficients are prepared with the Benoist cell homogenization prescription

  19. The activation cross section library UKACT1 and the inventory code FISPACT

    International Nuclear Information System (INIS)

    Forrest, R.A.

    1989-01-01

    The UK activation library for fusion applications, UKACT1, supersedes the existing UKCTRIIIA library. It contains neutron induced reaction data for 8719 reactions on 625 target nuclides. The library is used by the inventory code FISPACT which is a modified version of the existing code FISPIN. A library of decay information for all the 1314 nuclides involved is also required for calculations and this is also briefly described. UKACT1 will be used for irradiation calculations and as the starting point for a new version which will contain improved data for the most important reactions. These will be identified using the sensitivity subroutine in FISPACT. 16 refs, 1 fig., 2 tabs

  20. 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.

  1. Improvements to the nuclear model code GNASH for cross section calculations at higher energies

    International Nuclear Information System (INIS)

    Young, P.G.; Chadwick, M.B.

    1994-01-01

    The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV

  2. 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.)

  3. Analysis of reaction cross-section production in neutron induced fission reactions on uranium isotope using computer code COMPLET.

    Science.gov (United States)

    Asres, Yihunie Hibstie; Mathuthu, Manny; Birhane, Marelgn Derso

    2018-04-22

    This study provides current evidence about cross-section production processes in the theoretical and experimental results of neutron induced reaction of uranium isotope on projectile energy range of 1-100 MeV in order to improve the reliability of nuclear stimulation. In such fission reactions of 235 U within nuclear reactors, much amount of energy would be released as a product that able to satisfy the needs of energy to the world wide without polluting processes as compared to other sources. The main objective of this work is to transform a related knowledge in the neutron-induced fission reactions on 235 U through describing, analyzing and interpreting the theoretical results of the cross sections obtained from computer code COMPLET by comparing with the experimental data obtained from EXFOR. The cross section value of 235 U(n,2n) 234 U, 235 U(n,3n) 233 U, 235 U(n,γ) 236 U, 235 U(n,f) are obtained using computer code COMPLET and the corresponding experimental values were browsed by EXFOR, IAEA. The theoretical results are compared with the experimental data taken from EXFOR Data Bank. Computer code COMPLET has been used for the analysis with the same set of input parameters and the graphs were plotted by the help of spreadsheet & Origin-8 software. The quantification of uncertainties stemming from both experimental data and computer code calculation plays a significant role in the final evaluated results. The calculated results for total cross sections were compared with the experimental data taken from EXFOR in the literature, and good agreement was found between the experimental and theoretical data. This comparison of the calculated data was analyzed and interpreted with tabulation and graphical descriptions, and the results were briefly discussed within the text of this research work. Copyright © 2018 The Authors. Published by Elsevier Ltd.. All rights reserved.

  4. Verification of SIGACE code for generating ACE format cross-section files with continuous energy at high temperature

    International Nuclear Information System (INIS)

    Li Zhifeng; Yu Tao; Xie Jinsen; Qin Mian

    2012-01-01

    Based on the recently released ENDF/B-VII. 1 library, high temperature neutron cross-section files are generated through SIGACE code using low temperature ACE format files. To verify the processed ACE file of SIGACE, benchmark calculations are performed in this paper. The calculated results of selected ICT, standard CANDU assembly, LWR Doppler coefficient and SEFOR benchmarks are well conformed with reference value, which indicates that high temperature ACE files processed by SIGACE can be used in related neutronics calculations. (authors)

  5. MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-12-01

    MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual

  6. 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.

  7. Creation of problem-dependent Doppler-broadened cross sections in the KENO Monte Carlo code

    International Nuclear Information System (INIS)

    Hart, Shane W.D.; Celik, Cihangir; Maldonado, G. Ivan; Leal, Luiz

    2016-01-01

    Highlights: • A quick method of Doppler broadening one- and two-dimensional cross sections has been added to KENO. • The method uses a finite difference method to broaden data to user defined temperatures. • Various problems and benchmarks were run to showcase results. • Results with the Doppler broadened cross sections are closer to benchmark results. - Abstract: This paper introduces a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. The problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.

  8. 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

  9. CRSEC: a general purpose Hauser--Feshbach code for the calculation of nuclear cross-sections and thermonuclear reaction rates

    International Nuclear Information System (INIS)

    Woosley, S.; Fowler, W.A.

    1977-09-01

    CRSEC is a FORTRAN IV computer code designed for the efficient calculation of average nuclear cross sections in situations where a statistical theory of nuclear reactions is applicable and where compound nuclear formation is the dominant reaction mechanism. This code generates cross sections of roughly factor of 2 accuracy for incident particle energies in the range of 10 keV to 10 MeV for most target nuclei from magnesium to bismuth. Exceptions usually involve reactions that enter the compound nucleus at such a low energy that fewer than 10 levels are present in the ''energy window of interest.'' The incident particle must be a neutron, proton, or alpha particle, and only binary reactions resulting in the emission of a single n, p, α, or γ (cascade) are calculated. CRSEC is quite fast, a complete calculation of 12 different reactions over a grid of roughly 150 energy points and the generation of Maxwellian averaged rates taking about 30 seconds of CDC7600 time. Also the semi-empirical parameterization of nuclear properties contained in CRSEC is very general. Greater accuracy may be obtained, however, by furnishing specific low-lying excited states, level density parameterization, and nuclear strength functions. A more general version of CRSEC, called CRSECI, is available that conserves isospin properly in all reactions and allows the user to specify a given degree of isospin mixing in the highly excited states of the compound nucleus. Besides the cross section as a function of center-of-mass energy, CRSEC also generates the Maxwell--Boltzmann averaged thermonuclear reaction rate and temperature dependent nuclear partition function for a grid of temperatures from 10 8 to 10 10 0 K. Sections of this report describe in greater detail the physics employed in CRSEC and how to use the code. 2 tables

  10. Computer calculation of neutron cross sections with Hauser-Feshbach code STAPRE incorporating the hybrid pre-compound emission model

    International Nuclear Information System (INIS)

    Ivascu, M.

    1983-10-01

    Computer codes incorporating advanced nuclear models (optical, statistical and pre-equilibrium decay nuclear reaction models) were used to calculate neutron cross sections needed for fusion reactor technology. The elastic and inelastic scattering (n,2n), (n,p), (n,n'p), (n,d) and (n,γ) cross sections for stable molybdenum isotopes Mosup(92,94,95,96,97,98,100) and incident neutron energy from about 100 keV or a threshold to 20 MeV were calculated using the consistent set of input parameters. The hydrogen production cross section which determined the radiation damage in structural materials of fusion reactors can be simply deduced from the presented results. The more elaborated microscopic models of nuclear level density are required for high accuracy calculations

  11. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

    International Nuclear Information System (INIS)

    Yang, W.S.; Lee, C.H.

    2008-01-01

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

  12. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

    2008-05-16

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

  13. High temperature structural integrity evaluation method and application studies by ASME-NH for the next generation reactor design

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, Jae Han

    2006-01-01

    The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500 .deg. C and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated

  14. Application of adjusted subpixel method (ASM) in HRCT measurements of the bronchi in bronchial asthma patients and healthy individuals.

    Science.gov (United States)

    Mincewicz, Grzegorz; Rumiński, Jacek; Krzykowski, Grzegorz

    2012-02-01

    Recently, we described a model system which included corrections of high-resolution computed tomography (HRCT) bronchial measurements based on the adjusted subpixel method (ASM). To verify the clinical application of ASM by comparing bronchial measurements obtained by means of the traditional eye-driven method, subpixel method alone and ASM in a group comprised of bronchial asthma patients and healthy individuals. The study included 30 bronchial asthma patients and the control group comprised of 20 volunteers with no symptoms of asthma. The lowest internal and external diameters of the bronchial cross-sections (ID and ED) and their derivative parameters were determined in HRCT scans using: (1) traditional eye-driven method, (2) subpixel technique, and (3) ASM. In the case of the eye-driven method, lower ID values along with lower bronchial lumen area and its percentage ratio to total bronchial area were basic parameters that differed between asthma patients and healthy controls. In the case of the subpixel method and ASM, both groups were not significantly different in terms of ID. Significant differences were observed in values of ED and total bronchial area with both parameters being significantly higher in asthma patients. Compared to ASM, the eye-driven method overstated the values of ID and ED by about 30% and 10% respectively, while understating bronchial wall thickness by about 18%. Results obtained in this study suggest that the traditional eye-driven method of HRCT-based measurement of bronchial tree components probably overstates the degree of bronchial patency in asthma patients. Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

  15. Application of adjusted subpixel method (ASM) in HRCT measurements of the bronchi in bronchial asthma patients and healthy individuals

    International Nuclear Information System (INIS)

    Mincewicz, Grzegorz; Rumiński, Jacek; Krzykowski, Grzegorz

    2012-01-01

    Background: Recently, we described a model system which included corrections of high-resolution computed tomography (HRCT) bronchial measurements based on the adjusted subpixel method (ASM). Objective: To verify the clinical application of ASM by comparing bronchial measurements obtained by means of the traditional eye-driven method, subpixel method alone and ASM in a group comprised of bronchial asthma patients and healthy individuals. Methods: The study included 30 bronchial asthma patients and the control group comprised of 20 volunteers with no symptoms of asthma. The lowest internal and external diameters of the bronchial cross-sections (ID and ED) and their derivative parameters were determined in HRCT scans using: (1) traditional eye-driven method, (2) subpixel technique, and (3) ASM. Results: In the case of the eye-driven method, lower ID values along with lower bronchial lumen area and its percentage ratio to total bronchial area were basic parameters that differed between asthma patients and healthy controls. In the case of the subpixel method and ASM, both groups were not significantly different in terms of ID. Significant differences were observed in values of ED and total bronchial area with both parameters being significantly higher in asthma patients. Compared to ASM, the eye-driven method overstated the values of ID and ED by about 30% and 10% respectively, while understating bronchial wall thickness by about 18%. Conclusions: Results obtained in this study suggest that the traditional eye-driven method of HRCT-based measurement of bronchial tree components probably overstates the degree of bronchial patency in asthma patients.

  16. Generation of one energy group cross section library with MC2 computer code

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da; Souza, A.L. de.

    1982-01-01

    One group temperature dependent cross sections are generated via MC 2 for Pu-242, Ni-58, Fe-56, U-235, U-238, Pu-239, Pu-240, Pu-241, Be-9 e Th-232. The influence of the buckling and the weighting functions is studied throught calculations of an important integral parameter: the critical radius. (author) [pt

  17. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  18. An overview and recent changes in Section I

    International Nuclear Information System (INIS)

    Bernstein, M.D.

    1995-01-01

    Power Boilers, Section 1 of the ASME B and PV Code, is the first and oldest of the Code sections, originally published in 1915. Eighty years later Section 1 is well established, and changes are typically incremental rather than dramatic. Before reviewing some of these recent changes, it is useful to provide an overview of Section 1 rules, which were originally, and are still today, standard rules for the construction of steam boilers. Changes of the last few years affect the following portions of Section 1: The Foreword; PG-39 rules regarding small nozzles which can be attached without postweld heat treatment; rules of PG-67 and PG-70 regarding safety valve relieving capacity; PW-28 rules covering weld procedure qualification for certain nonpressure parts; PW-39 rules for postweld heat treatment; a major revision to the PW-43 rules for calculating loads permitted on structural attachments to tube walls; and some ongoing revisions to the Manufacturer's Data Report Forms

  19. Sensitivity analysis of U238 cross sections in fast nuclear systems-SENSEAV-R computer code

    International Nuclear Information System (INIS)

    Amorim, E.S. de; D'Oliveira, A.B.; Oliveira, E.C. de

    1981-01-01

    For many performance parameters of reactors the tabulated ratio calculation/experiment indicate that some potential problems may exist either in the cross section data or in the calculation models used to investigate the critical experimental data. A first step toward drawing a more definite conclusion is to perform a selective analysis of sensitivity profiles and covariance data files for the cross section data used in the calculation. Many works in the current literature show that some of these uncertainties come from uncertainties in 238 U(n,γ), 238 U(n,f) 239 Pu(n,f). Perturbation methods were developed to analyze the effects of finite changes in a large number of cross sections and summarize the investigation by a group dependent sensitivity coefficient. As an application, the results of this investigation indicates that improvements should be done only on the medium and low energy ranges of 238 U(n,γ) based on an analysis of cost and economic benefits. (Author) [pt

  20. EDITAR: a module for reaction rate editing and cross-section averaging within the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1986-03-01

    The EDITAR module of the AUS neutronics code system edits one and two-dimensional flux data pools produced by other AUS modules to form reaction rates for materials and their constituent nuclides, and to average cross sections over space and energy. The module includes a Bsub(L) flux calculation for application to cell leakage. The STATUS data pool of the AUS system is used to enable the 'unsmearing' of fluxes and nuclide editing with minimal user input. The module distinguishes between neutron and photon groups, and printed reaction rates are formed accordingly. Bilinear weighting may be used to obtain material reactivity worths and to average cross sections. Bilinear weighting is at present restricted to diffusion theory leakage estimates made using mesh-average fluxes

  1. 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.

  2. HADES. A computer code for fast neutron cross section from the Optical Model; HADES. Un programa numerico para el calculo de seccciones eficaces neutronicas mediante el modelo optico

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J; Navarro, C

    1973-07-01

    A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs.

  3. Quality of recording of diabetes in the UK: how does the GP's method of coding clinical data affect incidence estimates? Cross-sectional study using the CPRD database.

    Science.gov (United States)

    Tate, A Rosemary; Dungey, Sheena; Glew, Simon; Beloff, Natalia; Williams, Rachael; Williams, Tim

    2017-01-25

    To assess the effect of coding quality on estimates of the incidence of diabetes in the UK between 1995 and 2014. A cross-sectional analysis examining diabetes coding from 1995 to 2014 and how the choice of codes (diagnosis codes vs codes which suggest diagnosis) and quality of coding affect estimated incidence. Routine primary care data from 684 practices contributing to the UK Clinical Practice Research Datalink (data contributed from Vision (INPS) practices). Incidence rates of diabetes and how they are affected by (1) GP coding and (2) excluding 'poor' quality practices with at least 10% incident patients inaccurately coded between 2004 and 2014. Incidence rates and accuracy of coding varied widely between practices and the trends differed according to selected category of code. If diagnosis codes were used, the incidence of type 2 increased sharply until 2004 (when the UK Quality Outcomes Framework was introduced), and then flattened off, until 2009, after which they decreased. If non-diagnosis codes were included, the numbers continued to increase until 2012. Although coding quality improved over time, 15% of the 666 practices that contributed data between 2004 and 2014 were labelled 'poor' quality. When these practices were dropped from the analyses, the downward trend in the incidence of type 2 after 2009 became less marked and incidence rates were higher. In contrast to some previous reports, diabetes incidence (based on diagnostic codes) appears not to have increased since 2004 in the UK. Choice of codes can make a significant difference to incidence estimates, as can quality of recording. Codes and data quality should be checked when assessing incidence rates using GP data. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

  4. Nationwide Risk-Based PCB Remediation Waste Disposal Approvals under Title 40 of the Code of Federal Regulations (CFR) Section 761.61(c)

    Science.gov (United States)

    This page contains information about Nationwide Risk-Based Polychlorinated Biphenyls (PCBs) Remediation Waste Disposal Approvals under Title 40 of the Code of Federal Regulations (CFR) Section 761.61(c)

  5. Nationwide Enviro Jet PCB Decontamination Approval and Notifications under Title 40 of the Code of Federal Regulations (CFR) Section 761.79(h)

    Science.gov (United States)

    This page contains information about approvals and notifications for Enviro Jet to Decontaminate PCB-contaminated natural gas pipelines under Title 40 of the Code of Federal Regulations (CFR) Section 761.79(h)

  6. 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

  7. URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

    International Nuclear Information System (INIS)

    Leal, L.C.; de Saussure, G.; Perez, R.B.

    1990-01-01

    The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables

  8. ASME stress linearization and classification - a discussion based on a case study

    International Nuclear Information System (INIS)

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

    2011-01-01

    The ASME code, specially in its Nuclear Division (Subsection NB - Class I Components), gives some recommendations to the structural analyst on how to perform the verifications required to prove the design as good as the by-analysis prevented failures modes. Each of these failure modes has specific stress limits which are established based on simple but conservative hypothesis like the material perfectly plastic behavior and the shell theory with its typical membrane and bending stresses with linear distribution along the thickness. Other detail to keep in mind is the code distinction between primary and secondary stresses (respectively, stress that came due to equilibrium and due to displacement compatibility). In general, the numerical models used in the analyses are developed with plane or 3D solid elements and due this fact no direct comparison with the code limits can be done and, besides that, the programs do not distinguish between primary and secondary stresses. Mostly, the later are produced due to the temperature variation but they also appear near discontinuities. Sometimes, this classification is not so clear or direct. To perform the required ASME Code verifications the analyst should obtain the membrane and bending stresses from the plane or 3-D model which is called stress linearization and, also, should classify them as primary and secondary. (The excess between the maximum stress at a point and the sum of these linearized values is called peak stress and is included in the fatigue verification.) This task, most of the time is not a simple one due to the nature of the involved load and/or the complex geometry under analysis. In fact, there are several studies discussing on how to perform these stress classification and linearization. The present paper shows a discussion on how to perform these verifications based on a generic geometry found in many plants, from petrochemical to nuclear, which emphasizes some of theses issues. (author)

  9. ASME stress linearization and classification - a discussion based on a case study

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, Carlos A. de J.; Faloppa, Altair A.; Mattar Neto, Miguel; Fainer, Gerson, E-mail: cmiranda@ipen.b, E-mail: afaloppa@ipen.b, E-mail: mmattar@ipen.b, E-mail: gfainer@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The ASME code, specially in its Nuclear Division (Subsection NB - Class I Components), gives some recommendations to the structural analyst on how to perform the verifications required to prove the design as good as the by-analysis prevented failures modes. Each of these failure modes has specific stress limits which are established based on simple but conservative hypothesis like the material perfectly plastic behavior and the shell theory with its typical membrane and bending stresses with linear distribution along the thickness. Other detail to keep in mind is the code distinction between primary and secondary stresses (respectively, stress that came due to equilibrium and due to displacement compatibility). In general, the numerical models used in the analyses are developed with plane or 3D solid elements and due this fact no direct comparison with the code limits can be done and, besides that, the programs do not distinguish between primary and secondary stresses. Mostly, the later are produced due to the temperature variation but they also appear near discontinuities. Sometimes, this classification is not so clear or direct. To perform the required ASME Code verifications the analyst should obtain the membrane and bending stresses from the plane or 3-D model which is called stress linearization and, also, should classify them as primary and secondary. (The excess between the maximum stress at a point and the sum of these linearized values is called peak stress and is included in the fatigue verification.) This task, most of the time is not a simple one due to the nature of the involved load and/or the complex geometry under analysis. In fact, there are several studies discussing on how to perform these stress classification and linearization. The present paper shows a discussion on how to perform these verifications based on a generic geometry found in many plants, from petrochemical to nuclear, which emphasizes some of theses issues. (author)

  10. 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.

  11. 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...

  12. Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

    Science.gov (United States)

    Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

    2009-11-01

    Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

  13. 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

  14. 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.)

  15. 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

  16. THEMIS-4: a coherent punctual and multigroup cross section library for Monte Carlo and SN codes from ENDF/B4

    International Nuclear Information System (INIS)

    Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

    1983-05-01

    The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented

  17. GNASH: a preequilibrium, statistical nuclear-model code for calculation of cross sections and emission spectra. [In FORTRAN for CDC 7600

    Energy Technology Data Exchange (ETDEWEB)

    Young, P.G.; Arthur, E.D.

    1977-11-01

    A new multistep Hauser--Feshbach code that includes corrections for preequilibrium effects is described. The code can calculate up to 60 decay reactions (cross sections and energy spectra) in one computation, and thereby provide considerable flexibility for handling processes with complicated reaction chains. Input parameter setup, problem output, and subroutine descriptions are given along with a sample problem calculation. A brief theoretical description is also included. 8 figures, 3 tables.

  18. Structural integrity assessment of a pressure container component. Design and service code implementation. Case studies

    International Nuclear Information System (INIS)

    Sanzi, H.C.

    2006-01-01

    In the present work, the most important results of the local stresses occurred in the cracked pipes with a axial through-wall crack (outer), produced during operation of a Petrochemical Plant, using finite elements method, are presented. As requested, the component has been verified based 3D FE plastic analysis, under the postulated failure loading, assuring with this method a high degree of accuracy in the results. Codes used by Design and Service, as ASME Section VIII Div. 2 and API 579, have been used in the analysis. (author) [es

  19. On-line monitoring and inservice inspection in codes; Betriebsueberwachung und wiederkehrende Pruefungen in den Regelwerken

    Energy Technology Data Exchange (ETDEWEB)

    Bartonicek, J.; Zaiss, W. [Gemeinschaftskernkraftwerk Neckar GmbH, Neckarwestheim (Germany); Bath, H.R. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany). Geschaeftsstelle des Kerntechnischen Ausschusses (KTA)

    1999-08-01

    The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [Deutsch] Fuer die Komponentenintegritaet sind die Schaedigungsmechanismen mit dem nach den Regelwerken einzuhaltenden Abstand abzusichern. Dabei ist die jeweils vorhandene (Ist-) Qualitaet als Ausgangspunkt entscheidend. Die Absicherung der vorhandenen Qualitaet im weiteren Betrieb erfolgt durch geeignete Betriebsueberwachung und wiederkehrende Pruefungen. Die Anforderungen der Regelwerke sind vergleichbar, wobei die Bestimmung der vorhandenen Qualitaet nach einer bestimmten Betriebszeit sowie deren Absicherung im weiteren Betrieb am vollstaendigsten auf Basis des KTA-Regelwerkes moeglich ist. Die Absicherung der Komponentenintegritaet im Betrieb beruht in deutschen konventionellen Regelwerken nur auf den wiederkehrenden Pruefungen (hauptsaechlich Druckpruefungen und Sichtpruefungen). Das KTA-Regelwerk forderte hier schon immer qualifizierte

  20. On-line monitoring and inservice inspection in codes

    International Nuclear Information System (INIS)

    Bartonicek, J.; Zaiss, W.; Bath, H.R.

    1999-01-01

    The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

  1. 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

  2. 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.

  3. 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...

  4. 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 ...

  5. URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

    International Nuclear Information System (INIS)

    Leal, L.C.; de Saussure, G.; Perez, R.B.

    1989-01-01

    The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs

  6. Current and proposed revisions, changes, and modifications to American codes and standards to address packaging, handling, and transportation of radioactive materials and how they relate to comparable international regulations

    International Nuclear Information System (INIS)

    Borter, W.H.; Froehlich, C.H.

    2004-01-01

    This paper addresses current and proposed revisions, additions, and modifications to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) (i.e., ''ASMEthe Code'') Section III, Division 3 and American National Standards Institute (ANSI)/ASME N14.6. It provides insight into the ongoing processes of the associated committees and highlights important revisions, changes, and modifications to this Code and Standard. The ASME Code has developed and issued Division 3 to address items associated with the transportation and storage of radioactive materials. It currently only addresses ''General Requirements'' in Subsections WA and ''Class TP (Type B) Containments'' (Transportation Packages) in Subsection WB, but is in the process of adding a new Subsection WC to address ''Class SC'' (Storage Containments). ANSI/ASME Standard N14.6 which interacts with components constructed to Division 3 by addressinges special lifting devices for radioactive material shipping containers. This Standard is in the process of a complete re-write. This Code and Standard can be classified as ''dynamic'' in that their committees meet at least four times a year to evaluate proposed modifications and additions that reflect current safety practices in the nuclear industry. These evaluations include the possible addition of new materials, fabrication processes, examination methods, and testing requirements. An overview of this ongoing process is presented in this paper along with highlights of the more important proposed revisions, changes, and modifications and how they relate to United States (US) and international regulations and guidance like International Atomic Energy Agency (IAEA) Requirement No. TS-R-1

  7. Evaluation of ETOG-3Q, ETOG-3, FLANGE-II, XLACS, NJOY and LINEAR/RECENT/GROUPIE computer codes concerning to the resonance contribution and background cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-12-01

    The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analysed. (author) [pt

  8. Evaluation of ETOG-3Q/ETOG-3, FLANGE-II, XLACS, NJOY and linear/recent/groupie codes for calculations of resonance and reference cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1991-01-01

    The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analyzed. (author)

  9. The Sections Relating to Death Penalty in Pakistan Penal Code as Compared with Shari'a (Islamic Law (A Comparative Study (Urdu

    Directory of Open Access Journals (Sweden)

    Dr. Abzahir Khan

    2016-01-01

    Full Text Available Law plays a pivotal role in the establishment of any peaceful society.islam, being proactive, devised important rules about 1400 years back for the safety of Deen, life, wealth, wisdom and Generation. Qatal (murder is a crime of taking soul of a humanbeing, about which Islam has announced Qisas i.e to do with assissinater what he has done it to killed human being. In the same manner Pakistan penal Code has gathered rules about crimes steped out in Pakistan. So Pakistan penal code, under several sections has the same punishment. This artcle throws light on Pakistan penal code sections about death Senctance in perspective of Islamic imperium, order and explanation.

  10. Cross sections in 25 groups obtained from ENDF/B-IV and ENDL/78 libraries, processed with GALAXY and NJOY computer codes

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Corcuera, R.P.

    1982-01-01

    The discrepancies existing between ENDF/B-IV and ENDL/78 libraries, in diferent energy regions are identified, and the order of the differences in multigroup sections are determined, when GALAXY or NJOY computer codes are used. (E.G.) [pt

  11. A Life-Cycle Risk-Informed Systems Structured Nuclear Code

    International Nuclear Information System (INIS)

    Hill, Ralph S. III

    2002-01-01

    Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The design code is a separate volume from the code for inservice inspections and both are separate from the standards for operations and maintenance. The ASME code for inservice inspections and code for nuclear plant operations and maintenance have adopted risk-informed methodologies for inservice inspection, preventive maintenance, and repair and replacement decisions. The American Institute of Steel Construction and the American Concrete Institute have incorporated risk-informed probabilistic methodologies into their design codes. It is proposed that the ASME nuclear code should undergo a planned evolution that integrates the various nuclear codes and standards and adopts a risk-informed approach across a facility life-cycle - encompassing design, construction, operation, maintenance and closure. (author)

  12. Activated sludge model (ASM) based modelling of membrane bioreactor (MBR) processes: a critical review with special regard to MBR specificities.

    Science.gov (United States)

    Fenu, A; Guglielmi, G; Jimenez, J; Spèrandio, M; Saroj, D; Lesjean, B; Brepols, C; Thoeye, C; Nopens, I

    2010-08-01

    Membrane bioreactors (MBRs) have been increasingly employed for municipal and industrial wastewater treatment in the last decade. The efforts for modelling of such wastewater treatment systems have always targeted either the biological processes (treatment quality target) as well as the various aspects of engineering (cost effective design and operation). The development of Activated Sludge Models (ASM) was an important evolution in the modelling of Conventional Activated Sludge (CAS) processes and their use is now very well established. However, although they were initially developed to describe CAS processes, they have simply been transferred and applied to MBR processes. Recent studies on MBR biological processes have reported several crucial specificities: medium to very high sludge retention times, high mixed liquor concentration, accumulation of soluble microbial products (SMP) rejected by the membrane filtration step, and high aeration rates for scouring purposes. These aspects raise the question as to what extent the ASM framework is applicable to MBR processes. Several studies highlighting some of the aforementioned issues are scattered through the literature. Hence, through a concise and structured overview of the past developments and current state-of-the-art in biological modelling of MBR, this review explores ASM-based modelling applied to MBR processes. The work aims to synthesize previous studies and differentiates between unmodified and modified applications of ASM to MBR. Particular emphasis is placed on influent fractionation, biokinetics, and soluble microbial products (SMPs)/exo-polymeric substances (EPS) modelling, and suggestions are put forward as to good modelling practice with regard to MBR modelling both for end-users and academia. A last section highlights shortcomings and future needs for improved biological modelling of MBR processes. (c) 2010 Elsevier Ltd. All rights reserved.

  13. Monitoring compliance with the International Code of Marketing of Breastmilk Substitutes in west Africa: multisite cross sectional survey in Togo and Burkina Faso.

    Science.gov (United States)

    Aguayo, Victor M; Ross, Jay S; Kanon, Souleyman; Ouedraogo, Andre N

    2003-01-18

    To monitor compliance with the International Code of Marketing of Breastmilk Substitutes in health systems, sales outlets, distribution points, and the news media in Togo and Burkina Faso, west Africa. Multisite cross sectional survey. Staff at 43 health facilities and 66 sales outlets and distribution points, 186 health providers, and 105 mothers of infants aged market commercial breast milk substitutes were found in 29 (44%) sales and distribution points. Forty commercial breast milk substitutes violated the labelling standards of the code: 21 were manufactured by Danone, 11 by Nestlé, and eight by other national and international manufacturers. Most (148, 90%) health providers had never heard of the code, and 66 mothers (63%) had never received any counselling on breast feeding by their health providers. In west Africa manufacturers are violating the code of marketing of breast milk substitutes. Comparable levels of code violations are observed with (Burkina Faso) or without (Togo) regulating legislation. Legislation must be accompanied by effective information, training, and monitoring systems to ensure that healthcare providers and manufacturers comply with evidence based practice and the code.

  14. Development of the next generation code system as an engineering modeling language (6). Development of a cross section adjustment and nuclear design accuracy evaluation solver

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Numata, Kazuyuki

    2008-01-01

    A new cross section adjustment and nuclear design accuracy evaluation solver was developed as a set of modules for MARBLE (multi-purpose advanced reactor physics analysis system based on language of engineering). In order to enhance the system extendibility and flexibility, the object-oriented design and analysis technique was adopted to the development. In the new system, it is easy to add a new design accuracy evaluation method because a new numerical calculation module is independent from other modules. Further, several new functions such as searching and editing calculation data are provided in the new solver. These functions can be easily customised by users because they are designed to work cooperatively with Python scripting language, which is used as a user interface of the MARBLE system. In order to validate the new solver, a test calculation was performed for a realistic calculation case of creating a new unified cross section library. In the test calculation, results calculated by the new solver agreed well with those by the conventional code system. In addition, it is possible to reuse existing input data files prepared for the conventional code system because the new solver utilities support the conventional formats. Because the new solver implements all main functions of the conventional code system, MARBLE can be used as a new calculation code system for cross section adjustment and nuclear design accuracy evaluation

  15. Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

    International Nuclear Information System (INIS)

    Cai, Li

    2014-01-01

    In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

  16. 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

  17. 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

  18. 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...

  19. Theoretical calculations of the reaction cross-sections for proton-induced reactions on natural copper using ALICE-IPPE code

    International Nuclear Information System (INIS)

    Alharbi, A.A.; Azzam, A.

    2012-01-01

    A theoretical study of the nuclear-reaction cross sections for proton-induced reactions on 63 Cu and 65 Cu was performed in the proton energy range from threshold values up to 50 MeV. The produced nuclei were different isotopes of Zn, Cu, Ni, Co and Mn, some of which have important applications. The reaction cross-section calculations were performed using the ALICE-IPPE code, which depends on the pre-equilibrium compound nucleus model. This code is suitable for the studied energy and isotopic mass ranges. Approximately 14 excitation functions for the different reactions have been constructed from the calculated cross-section values. The excitation function curves for the proton reactions with natural copper targets have been constructed from those for enriched targets using the natural abundance of the copper isotopes. Comparisons between the calculated excitation functions with those previously experimentally measured are given whenever the experimental values were available. Some statistical parameters were introduced to control the quality of the fitting between both the experimental and the theoretical calculated cross-section values. - Highlights: ► We performed reaction cross section calculations using ALICE-IPPE code. ► We constructed 14 excitation functions for nat Cu(p,xn)Zn,Cu,Ni,Co,Mn reactions. ► The available experimental data were fitted to the performed ALICE-IPPE calculations. ► Statistical parameters were introduced to control the quality of the fitting. ► The code failed to fit the experimental data for reactions with large nucleon emissions.

  20. Design validation of the ITER EC upper launcher according to codes and standards

    Energy Technology Data Exchange (ETDEWEB)

    Spaeh, Peter, E-mail: peter.spaeh@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Aiello, Gaetano [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Gagliardi, Mario [Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); F4E, Fusion for Energy, Joint Undertaking, Barcelona (Spain); Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro [Karlsruhe Institute of Technology, Institute for Applied Materials, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Weinhorst, Bastian [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

  1. Design validation of the ITER EC upper launcher according to codes and standards

    International Nuclear Information System (INIS)

    Spaeh, Peter; Aiello, Gaetano; Gagliardi, Mario; Grossetti, Giovanni; Meier, Andreas; Scherer, Theo; Schreck, Sabine; Strauss, Dirk; Vaccaro, Alessandro; Weinhorst, Bastian

    2015-01-01

    Highlights: • A set of applicable codes and standards has been chosen for the ITER EC upper launcher. • For a particular component load combinations, failure modes and stress categorizations have been determined. • The design validation was performed in accordance with the “design by analysis”-approach of the ASME boiler and pressure vessel code section III. - Abstract: The ITER electron cyclotron (EC) upper launcher has passed the CDR (conceptual design review) in 2005 and the PDR (preliminary design review) in 2009 and is in its final design phase now. The final design will be elaborated by the European consortium ECHUL-CA with contributions from several research institutes in Germany, Italy, the Netherlands and Switzerland. Within this consortium KIT is responsible for the design of the structural components (the upper port plug, UPP) and also the design integration of the launcher. As the selection of applicable codes and standards was under discussion for the past decade, the conceptual and the preliminary design of the launcher structure were not elaborated in straight accordance with a particular code but with a variety of well-acknowledged engineering practices. For the final design it is compulsory to validate the design with respect to a typical engineering code in order to be compliant with the ITER quality and nuclear requirements and to get acceptance from the French regulator. This paper presents typical design validation of the closure plate, which is the vacuum and Tritium barrier and thus a safety relevant component of the upper port plug (UPP), performed with the ASME boiler and pressure vessel code. Rationales for choosing this code are given as well as a comparison between different design methods, like the “design by rule” and the “design by analysis” approach. Also the selections of proper load specifications and the identification of potential failure modes are covered. In addition to that stress categorizations, analyses

  2. Identifying Adverse Events Using International Classification of Diseases, Tenth Revision Y Codes in Korea: A Cross-sectional Study

    Directory of Open Access Journals (Sweden)

    Minsu Ock

    2018-01-01

    Full Text Available Objectives The use of administrative data is an affordable alternative to conducting a difficult large-scale medical-record review to estimate the scale of adverse events. We identified adverse events from 2002 to 2013 on the national level in Korea, using International Classification of Diseases, tenth revision (ICD-10 Y codes. Methods We used data from the National Health Insurance Service-National Sample Cohort (NHIS-NSC. We relied on medical treatment databases to extract information on ICD-10 Y codes from each participant in the NHIS-NSC. We classified adverse events in the ICD-10 Y codes into 6 types: those related to drugs, transfusions, and fluids; those related to vaccines and immunoglobulin; those related to surgery and procedures; those related to infections; those related to devices; and others. Results Over 12 years, a total of 20 817 adverse events were identified using ICD-10 Y codes, and the estimated total adverse event rate was 0.20%. Between 2002 and 2013, the total number of such events increased by 131.3%, from 1366 in 2002 to 3159 in 2013. The total rate increased by 103.9%, from 0.17% in 2002 to 0.35% in 2013. Events related to drugs, transfusions, and fluids were the most common (19 446, 93.4%, followed by those related to surgery and procedures (1209, 5.8% and those related to vaccines and immunoglobulin (72, 0.3%. Conclusions Based on a comparison with the results of other studies, the total adverse event rate in this study was significantly underestimated. Improving coding practices for ICD-10 Y codes is necessary to precisely monitor the scale of adverse events in Korea.

  3. Status of reactor physics activities on cross section generation and functionalization for the prismatic very high temperature reactor, and development of spatially-heterogeneous codes

    International Nuclear Information System (INIS)

    Lee, C. H.; Zhong, Z.; Taiwo, T. A.; Yang, W. S.; Smith, M. A.; Palmiotti, G.

    2006-01-01

    The cross section generation methodology and procedure for design and analysis of the prismatic Very High Temperature Gas-cooled Reactor (VHTR) core have been addressed for the DRAGON and REBUS-3/DIF3D code suite. Approaches for tabulation and functionalization of cross sections have been investigated and implemented. The cross sections are provided at different burnup and fuel and moderator temperature states. In the tabulation approach, the multigroup cross sections are tabulated as a function of the state variables so that a cross section file is able to cover the range of core operating conditions. Cross sections for points between tabulated data points are fitted simply by linear interpolation. For the functionalization approach, an investigation of the applicability of quadratic polynomials and linear coupling for fuel and moderator temperature changes has been conducted, based on the observation that cross sections are monotonically changing with fuel or moderator temperatures. Preliminary results show that the functionalization makes it possible to cover a wide range of operating temperature conditions with only six sets of data per burnup, while maintaining a good accuracy and significantly reducing the size of the cross section file. In these approaches, the number of fission products has been minimized to a few nuclides (I/Xe/Pm/Sm and a lumped fission product) to reduce the overall computation time without sacrificing solution accuracy. Discontinuity factors (DFs) based on nodal equivalence theory have been introduced to accurately represent the significant change in neutron spectrum at the interface of the fuel and reflector regions as well as between different fuel blocks (e.g., fuel elements with burnable poisons or control rods). Using the DRAGON code, procedures have been established for generating cross sections for fuel and reflector blocks with and without control absorbers. The preliminary results indicate that the solution accuracy is improved

  4. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  5. 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.

  6. 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.

  7. Monitoring compliance with the International Code of Marketing of Breastmilk Substitutes in west Africa: multisite cross sectional survey in Togo and Burkina Faso

    Science.gov (United States)

    Aguayo, Victor M; Ross, Jay S; Kanon, Souleyman; Ouedraogo, Andre N

    2003-01-01

    Objectives To monitor compliance with the International Code of Marketing of Breastmilk Substitutes in health systems, sales outlets, distribution points, and the news media in Togo and Burkina Faso, west Africa. Design Multisite cross sectional survey. Participants Staff at 43 health facilities and 66 sales outlets and distribution points, 186 health providers, and 105 mothers of infants aged ⩽5 months in 16 cities. Results Six (14%) health facilities had received donations of breast milk substitutes. All donations were being given to mothers free of charge. Health providers in five (12%) health facilities had received free samples of breast milk substitutes for purposes other than professional research or evaluation. Health professionals in five (12%) health facilities had received promotional gifts from manufacturers. Promotional materials of commercial breast milk substitutes were found in seven (16%) health facilities. Special displays to market commercial breast milk substitutes were found in 29 (44%) sales and distribution points. Forty commercial breast milk substitutes violated the labelling standards of the code: 21 were manufactured by Danone, 11 by Nestlé, and eight by other national and international manufacturers. Most (148, 90%) health providers had never heard of the code, and 66 mothers (63%) had never received any counselling on breast feeding by their health providers. Conclusion In west Africa manufacturers are violating the code of marketing of breast milk substitutes. Comparable levels of code violations are observed with (Burkina Faso) or without (Togo) regulating legislation. Legislation must be accompanied by effective information, training, and monitoring systems to ensure that healthcare providers and manufacturers comply with evidence based practice and the code. What is already known on this topicAll member states of the World Health Assembly have reaffirmed their support for the International Code of Marketing of Breastmilk

  8. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.; De Crecy, F.

    1994-01-01

    This paper deals with the application of the adjoint sensitivity method (ASM) to thermal hydraulic codes. The advantage of the method is to use small central processing unit time in comparison with the usual approach requiring one complete code run per sensitivity determination. In the first part the mathematical aspects of the problem are treated, and the applicability of the method of the functional-type response of a thermal hydraulic model is demonstrated. On a simple example of non-linear hyperbolic equation (Burgers equation) the problem has been analysed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the continuous ASM and the discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the discrete ASM constitutes a practical solution for thermal hydraulic codes. The application of the discrete ASM to the thermal hydraulic safety code CATHARE is then presented for two examples. They demonstrate that the discrete ASM constitutes an efficient tool for the analysis of code sensitivity. ((orig.))

  9. [Quantitative analysis of the structure of neuronal dendritic spines in the striatum using the Leitz-ASM system].

    Science.gov (United States)

    Leontovich, T A; Zvegintseva, E G

    1985-10-01

    Two principal classes of striatum long axonal neurons (sparsely ramified reticular cells and densely ramified dendritic cells) were analyzed quantitatively in four animal species: hedgehog, rabbit, dog and monkey. The cross section area, total dendritic length and the area of dendritic field were measured using "LEITZ-ASM" system. Classes of neurons studied were significantly different in dogs and monkeys, while no differences were noted between hedgehog and rabbit. Reticular neurons of different species varied much more than dendritic ones. Quantitative analysis has revealed the progressive increase in the complexity of dendritic tree in mammals from rabbit to monkey.

  10. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS); Generacion de la biblioteca de secciones eficaces de neutrones para el codigo Record del Sistema de Administracion de Combustible (FMS)

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-11-15

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  11. Program POD; A computer code to calculate nuclear elastic scattering cross sections with the optical model and neutron inelastic scattering cross sections by the distorted-wave born approximation

    International Nuclear Information System (INIS)

    Ichihara, Akira; Kunieda, Satoshi; Chiba, Satoshi; Iwamoto, Osamu; Shibata, Keiichi; Nakagawa, Tsuneo; Fukahori, Tokio; Katakura, Jun-ichi

    2005-07-01

    The computer code, POD, was developed to calculate angle-differential cross sections and analyzing powers for shape-elastic scattering for collisions of neutron or light ions with target nucleus. The cross sections are computed with the optical model. Angle-differential cross sections for neutron inelastic scattering can also be calculated with the distorted-wave Born approximation. The optical model potential parameters are the most essential inputs for those model computations. In this program, the cross sections and analyzing powers are obtained by using the existing local or global parameters. The parameters can also be inputted by users. In this report, the theoretical formulas, the computational methods, and the input parameters are explained. The sample inputs and outputs are also presented. (author)

  12. Section XI -- 25 years of development

    International Nuclear Information System (INIS)

    Hedden, O.F.

    1996-01-01

    The original concept of nuclear power plant designers was that the higher standards of design and fabrication would make inservice inspections unnecessary, and little attention was given to provisions for access. By 1966 the Atomic Energy Commission recognized that a planned program of periodic inservice inspections would be needed. They began development of criteria, and encouraged industry code-writing organizations to do likewise. These groups joined forces in 1968, and their product was published by ASME in 1970 as part of the Boiler and Pressure Code, Section XI, Rules for Inservice Inspection of Nuclear Reactor Coolant Systems. Section XI, 24 pages in 1970, is now 723 pages. While it originally covered only light water reactor Class 1 components and piping, it now includes Class 2, 3, and containment, and liquid metal cooled reactor plants. Along the way, rules have been developed for gas-cooled and low pressure heavy water reactor plants. The growth in size of Section XI from its modest beginning has been largely because of recognition that the rules governing plant inspection/operation need to be considerably different from the rules provided for the component designer/manufacturer. Rules have been developed in the areas of repair/replacement technology, NDE methodology, NDE acceptance standards, and analytical evaluation methods in the absence of appropriate rules in Section III

  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. 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

  15. 78 FR 12089 - Revision of Certain Dollar Amounts in the Bankruptcy Code Prescribed Under Section 104(a) of the...

    Science.gov (United States)

    2013-02-21

    ... United States Courts, Washington, DC 20544, telephone (202) 502-1900, or by email at Bankruptcy_Judges... business debtor. time it appears). time it appears). Section 109(e)--allowable debt $360,475 (each $383,175...

  16. A computer code for the prediction of mill gases and hot air distribution between burners sections as input parameters for 3D CFD furnace calculation

    International Nuclear Information System (INIS)

    Tucakovic, Dragan; Zivanovic, Titoslav; Beloshevic, Srdjan

    2006-01-01

    Current computer technology development enables application of powerful software packages that can provide a reliable insight into real operating conditions of a steam boiler in the Thermal Power Plant. Namely, an application of CFD code to the 3D analysis of combustion and heat transfer in a furnace provides temperature, velocity and concentration fields in both cross sectional and longitudinal planes of the observed furnace. In order to obtain reliable analytical results, which corresponds to real furnace conditions, it is necessary to accurately predict a distribution of mill gases and hot air between burners' sections, because these parameters are input values for the furnace 3D calculation. Regarding these tasks, the computer code for the prediction of mill gases and hot air distribution has been developed at the Department for steam boilers of the Faculty of Mechanical Engineering in Belgrade. The code is based on simultaneous calculations of material and heat balances for fan mill and air tracts. The aim of this paper is to present a methodology of performed calculations and results obtained for the steam boiler furnace of 350 MWe Thermal Power Plant equipped with eight fan mills. Key words: mill gases, hot air, aerodynamic calculation, air tract, mill tract.

  17. ANITA-IEAF activation code package - updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

    Science.gov (United States)

    Frisoni, Manuela

    2017-09-01

    ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

  18. Application of modified REFIT code for J-PARC/MLF to evaluation of neutron capture cross section on 155,157Gd

    Science.gov (United States)

    Mizuyama, Kazuhito; Iwamoto, Nobuyuki; Iwamoto, Osamu; Hasemi, Hiroyuki; Kino, Koichi; Kimura, Atsushi; Kiyanagi, Yoshiaki

    2017-09-01

    In order to analyze the experimental data measured by the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Japan Proton Accelerator Research Complex/Materials and Life Science Experimental Facility (J-PARC/MLF), it is necessary to take into account the double-bunch structure of the neutron pulse and the energy resolution function for the operational condition of the J-PARC/MLF. The modified REFIT code has been developed to treat the double-bunch neutron pulse and the energy resolution function for J-PARC/MLF. In this study, we applied the modified REFIT code to analyze the new data of the neutron capture cross section of 155Gd and 157Gd recently measured by ANNRI in the J-PARC/MLF, and obtained the resonance parameters of two Gd isotopes. We discussed the differences between the our obtained results and the other libraries.

  19. Sampling procedures using optical-data and partial wave cross sections in a Monte Carlo code for simulating kilovolt electron and positron transport in solids

    International Nuclear Information System (INIS)

    Fernandez-Varea, J.M.; Salvat, F.; Liljequist, D.

    1994-09-01

    The details of a Monte Carlo code for computing the penetration and energy loss of electrons and positrons in solids are described. The code, intended for electrons and positrons with energies from ∼ 100 eV to ∼ 100 keV, is based on the simulation of individual elastic and inelastic collisions. Elastic collisions are simulated using differential cross sections computed by the relativistic partial wave method applied to a muffin-tin Dirac-Hartree-Fock-Slater potential. Inelastic collisions are simulated by means of a model based on optical and photoelectric data, which are extended to the non-zero momentum transfer region by means of somewhat different algorithms for valence electron excitations and inner-shell excitations. This report focuses on the description of detailed formulae and sampling methods. 10 refs, 3 figs, 8 tabs

  20. Evaluation procedure of the structural integrity of a pipe of nuclear use. Application of codes for design and service. Case study

    International Nuclear Information System (INIS)

    Sanzi, H.; Asta, E.

    2009-01-01

    In the present work, we are presenting the most important results of the local stresses occurred in the cracked pipes with a axial through-wall, under Failure Concept 0.1A, using Finite Element Method and Fracture Mechanics. As requested, the component has been verified based 3D FE plastic analysis, under the postulated failure loading, assuring with this method a high degree of accuracy in the results. Codes used by Design and Service, as ASME Section III Div. 1 and API 579, have been used in the analysis. (author)

  1. Classical collisions of protons with hydrogen atoms. [Equations of motion, cross sections, C code, FORTRAN, moments of inertia

    Energy Technology Data Exchange (ETDEWEB)

    Banks, D; Hughes, P E; Percival, I C [Queen Mary Coll., London (UK); Barnes, K S [National Health Service Operational Research Group, Royal Institute of Public Administration, Reading, Berkshire, UK; Richards, D [Open Univ., Milton Keynes (UK); Valentine, N A [Digital Equipment Corporation, Bilton House, Uxbridge Road, Ealing, London, UK; Wilson, Mc B [Glasgow Univ. (UK). Dept. of Natural Philosophy

    1977-01-01

    The program solves the equations of motion for the interaction of 3 charged particles, obtaining final states in terms of initial states, and energy transfers, angles of ejection, and final cartesian co-ordinates of relative motion. Using a Monte Carlo method on many orbits total ionization and charge transfer cross sections, integral energy transfer cross sections and moments of energy transfers are estimated. Facilities are provided for obtaining angular distributions, momentum transfer cross sections and for comparison with various approximate classical theories. The equations of motion are solved using stepwise fourth-order Runge-Kutta integration with automatic steplength change. Selection of initial conditions is determined by the user, usually as a statistical distribution determined by a pseudorandom number subroutine. Classical representation theory and transformation methods are extensively used.

  2. Awareness and reported violations of the WHO International Code and Pakistan's national breastfeeding legislation; a descriptive cross-sectional survey

    Directory of Open Access Journals (Sweden)

    Faragher Brian

    2008-10-01

    Full Text Available Abstract Background National legislation in Pakistan adopted the International Code of Marketing of Breastmilk Substitutes in 2002 to restrict the promotion of infant formula feeding. Our objectives were to assess health professionals' awareness of this law in urban government hospitals and describe their reports of violations, including receiving free samples, gifts and sponsorship. Methods Structured interviews were conducted with health staff between July and August 2006 at 12 urban government hospitals in Islamabad, Rawalpindi and Peshawar including paediatricians, obstetricians, nurses, resident doctors, midwives and lady health visitors (LHVs. Results Of the 427 health workers interviewed, the majority were not aware of the national breastfeeding law (70.5%; n = 301 or the International Code (79.6%; n = 340. Paediatricians, and staff who had been working for 10 years or more, were more likely to be aware of the law [OR = 7.00, 95% CI 3.12, 15.7 (paediatricians; OR = 2.48, 95% CI 1.45, 4.24 (10 years working]. More than one third (38.4%, n = 164 had received small gifts such as pens, pencils and calendars; 12.4% (n = 53 had received sponsorship for training or conferences; and 15.9% (n = 68 had received free samples of infant formula from the Companies. Staff who were aware of the law were also more likely to report receiving gifts (OR = 1.64, 95% CI 1.08, 2.51 and free samples (OR = 1.86, 95% CI 1.09, 3.19. Conclusion Most hospital health professionals were unaware of national breastfeeding legislation in Pakistan, and infant formula companies were continuing to flout the ban on gifts, free samples and sponsorship for health staff.

  3. Code, standard and specifications

    International Nuclear Information System (INIS)

    Abdul Nassir Ibrahim; Azali Muhammad; Ab. Razak Hamzah; Abd. Aziz Mohamed; Mohamad Pauzi Ismail

    2008-01-01

    Radiography also same as the other technique, it need standard. This standard was used widely and method of used it also regular. With that, radiography testing only practical based on regulations as mentioned and documented. These regulation or guideline documented in code, standard and specifications. In Malaysia, level one and basic radiographer can do radiography work based on instruction give by level two or three radiographer. This instruction was produced based on guideline that mention in document. Level two must follow the specifications mentioned in standard when write the instruction. From this scenario, it makes clearly that this radiography work is a type of work that everything must follow the rule. For the code, the radiography follow the code of American Society for Mechanical Engineer (ASME) and the only code that have in Malaysia for this time is rule that published by Atomic Energy Licensing Board (AELB) known as Practical code for radiation Protection in Industrial radiography. With the existence of this code, all the radiography must follow the rule or standard regulated automatically.

  4. 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.

  5. Estimation of the defect detection probability for ultrasonic tests on thick sections steel weldments. Technical report

    International Nuclear Information System (INIS)

    Johnson, D.P.; Toomay, T.L.; Davis, C.S.

    1979-02-01

    An inspection uncertainty analysis of published PVRC Specimen 201 data is reported to obtain an estimate of the probability of recording an indication as a function of imperfection height for ASME Section XI Code ultrasonic inspections of the nuclear reactor vessel plate seams and to demonstrate the advantages of inspection uncertainty analysis over conventional detection/nondetection counting analysis. This analysis found the probability of recording a significant defect with an ASME Section XI Code ultrasonic inspection to be very high, if such a defect should exist in the plate seams of a nuclear reactor vessel. For a one-inch high crack, for example, this analysis gives a best estimate recording probability of .985 and a 90% lower confidence bound recording probabilty of .937. It is also shown that inspection uncertainty analysis gives more accurate estimates and gives estimates over a much greater flaw size range than is possible with conventional analysis. There is reason to believe that the estimation procedure used is conservative, the estimation is based on data generated several years ago, on very small defects, in an environment that is different from the actual in-service inspection environment

  6. [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.

  7. 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.

  8. The Determination of Neutron-Induced Reaction Cross Section Data on Even-Even, Magic- Number Nuclide Chromium 52 Using EXIFON Code

    International Nuclear Information System (INIS)

    Jonah, S.A.

    2013-01-01

    The EXIFON code version 2.0 is a calculational tool, which is based on both many-body theory and random matrix physics. In this work, it has been used to calculate neutron induced reaction cross section data from 0 to 20 MeV on an even-even, magic number nuclide 52 Cr with neutron number, N=28. Specifically, the (n,p), (n,α) and (n,2n) reaction cross section data were calculated as functions of incident energy of neutrons. Data obtained from the experimental data in the IAEA, EXFOR data Library and recommended data libraries around the globe, JENDL, ENDF and JEFF were used to validate the calculated data. The data indicate that the calculated data without shell corrections are in good agreement with experimental data as well as the recommended data from the evaluated data libraries. The calculated results could provide useful insight into the choice of some input parameters near closed shells using the EXIFON code.

  9. Measurements of D-T neutron induced radioactivity in plasma-facing materials and their role in qualification of activation cross-section libraries and codes

    International Nuclear Information System (INIS)

    Kumar, A.; Abdou, M.A.; Kosako, K.; Oyama, Y.; Nakamura, T.; Maekawa, H.

    1995-01-01

    The D-T neutron-induced radioactivity constitutes one of the foremost issues in fusion reactor design. The validation of activation cross-sections and decay data libraries is one of the important requirements for validating ITER design from safety and waste disposal viewpoints. An elaborate, experimental program was initiated in 1988, under USDOE-JAERI collaborative program, to validate the radioactivity codes/libraries. The measurements of decay-γ spectra from irradiated, high purity samples of Al, Si, Ti, V, Cr, Mn-Cu alloy, Fe, Co, Ni, Cu, stainless steel 316 (AISI 316), Zn, Zr, Nb, Mo, In, Sn, Ta, W, and Pb, among others, were conducted under D-T neutron fluences varying from 1.6 x 10 10 ncm -2 to 6.1 x 10 13 ncm -2 . As many as 14 neutron energy spectra were covered for a number of materials. The analysis of isotopic activities of the irradiated materials using activation cross-section libraries of four leading radioactivity codes, i.e. ACT4/THIDA-2, REAC-3, DKR-ICF, and RACC, has shown large discrepancies among the calculations, on the one hand, and between the calculations and the measurements, on the other. A discussion is also presented on definition and obtention of safety cum quality factors for various activation libraries. (orig.)

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. Disposal Notifications and Quarterly Membership Updates for the Utility Solid Waste Group Members’ Risk-Based Approvals to Dispose of PCB Remediation Waste Under Title 40 of the Code of Federal Regulations Section 761.61(c)

    Science.gov (United States)

    Disposal Notifications and Quarterly Membership Updates for the Utility Solid Waste Group Members’ Risk-Based Approvals to Dispose of Polychlorinated Biphenyl (PCB) Remediation Waste Under Title 40 of the Code of Federal Regulations Section 761.61(c)

  15. 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

  16. The use of routine cell codes for evaluating the In-rod effective cross sections of resonance absorber

    International Nuclear Information System (INIS)

    Segev, M.

    1996-01-01

    The last three years have witnessed an increasing interest in the in-rod distribution of resonance absorption and of temperature. High burnup, especially beyond the 'classical' limit of 30 GWd/T, is expected to generate uneven in-rod isotope distributions with consequences for fuel rod integrity and reactor Doppler feedback. There are recent indications that, even for a freshly loaded uranium-oxide rod, proper account of the U 238 in-rod absorption rate distribution results in a doppler coefficient some 15% lower in magnitude than its routinely calculated value. Presently a special form of application is made of the Bogart approach. This approach is based on the fact that, as a fuel rod is filled in from the outside, its resonance capture rate increases monotonically, despite file decreasing effective capture cross section for the thickness annulus. Bogart used His observation to derive a differential equation for the in-rod absorption distribution. Presently we capitalize on the idea in a discrete form. (author)

  17. The use of routine cell codes for evaluating the In-rod effective cross sections of resonance absorber

    Energy Technology Data Exchange (ETDEWEB)

    Segev, M [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Nuclear Engineering

    1996-12-01

    The last three years have witnessed an increasing interest in the in-rod distribution of resonance absorption and of temperature. High burnup, especially beyond the `classical` limit of 30 GWd/T, is expected to generate uneven in-rod isotope distributions with consequences for fuel rod integrity and reactor Doppler feedback. There are recent indications that, even for a freshly loaded uranium-oxide rod, proper account of the U{sup 238} in-rod absorption rate distribution results in a doppler coefficient some 15% lower in magnitude than its routinely calculated value. Presently a special form of application is made of the Bogart approach. This approach is based on the fact that, as a fuel rod is filled in from the outside, its resonance capture rate increases monotonically, despite file decreasing effective capture cross section for the thickness annulus. Bogart used His observation to derive a differential equation for the in-rod absorption distribution. Presently we capitalize on the idea in a discrete form. (author).

  18. 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.

  19. 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.

  20. 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)

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

    International Nuclear Information System (INIS)

    Motohashi, Kazuhiko

    2009-01-01

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

  2. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  3. 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.

  4. 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.

  5. Assessment of crack-like flaws - Comparison of procedures in BS 7910, API 579-1/ASME FFS-1, RSE-M AND FITNET

    International Nuclear Information System (INIS)

    Chaudouet, A.

    2007-01-01

    Among all Fitness For Service Codes enabling to assess flaws in metallic structures and to evaluate their remaining life, new editions of the most important ones at the international level have been issued recently. The latest edition of BS 7910 in United Kingdom has been released in October 2005. In the USA, API and ASME have edited a new standard in 2007, API579-1/ASME FFS-1, dedicated to pressure equipment. In France, the rules concerning the of Light Water Reactors, RSE-M, have been updated in 2005. Finally, in Europe, the FITNET network is writing a document based on BS 7910 but extended with the most recent results in this domain. Rules given in these documents to assess crack-like flaws with respect to fracture and fatigue propagation are presented. They are compared in order to point out the most interesting aspects of each ones and to identify those which could be generalized. An example assessed with the above mentioned 'Codes' enlightens the differences in the results with respect to the 'Code' used. (author) [fr

  6. 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.

  7. 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.

  8. Recommendations for codes and standards to be used for design and fabrication of high level waste canister

    International Nuclear Information System (INIS)

    Bermingham, A.J.; Booker, R.J.; Booth, H.R.; Ruehle, W.G.; Shevekov, S.; Silvester, A.G.; Tagart, S.W.; Thomas, J.A.; West, R.G.

    1978-01-01

    This study identifies codes, standards, and regulatory requirements for developing design criteria for high-level waste (HLW) canisters for commercial operation. It has been determined that the canister should be designed as a pressure vessel without provision for any overpressure protection type devices. It is recommended that the HLW canister be designed and fabricated to the requirements of the ASME Section III Code, Division 1 rules, for Code Class 3 components. Identification of other applicable industry and regulatory guides and standards are provided in this report. Requirements for the Design Specification are found in the ASME Section III Code. It is recommended that design verification be conducted principally with prototype testing which will encompass normal and accident service conditions during all phases of the canister life. Adequacy of existing quality assurance and licensing standards for the canister was investigated. One of the recommendations derived from this study is a requirement that the canister be N stamped. In addition, acceptance standards for the HLW waste should be established and the waste qualified to those standards before the canister is sealed. A preliminary investigation of use of an overpack for the canister has been made, and it is concluded that the use of an overpack, as an integral part of overall canister design, is undesirable, both from a design and economics standpoint. However, use of shipping cask liners and overpack type containers at the Federal repository may make the canister and HLW management safer and more cost effective. There are several possible concepts for canister closure design. These concepts can be adapted to the canister with or without an overpack. A remote seal weld closure is considered to be one of the most suitable closure methods; however, mechanical seals should also be investigated

  9. CREOLE experiment study on the reactivity temperature coefficient with sensitivity and uncertainty analysis using the MCNP5 code and different neutron cross section evaluations

    International Nuclear Information System (INIS)

    Boulaich, Y.; El Bardouni, T.; Erradi, L.; Chakir, E.; Boukhal, H.; Nacir, B.; El Younoussi, C.; El Bakkari, B.; Merroun, O.; Zoubair, M.

    2011-01-01

    Highlights: → In the present work, we have analyzed the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. → Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values. → In order to specify the source of the relatively large discrepancy in the case of ENDF-BVII nuclear data evaluation, the k eff discrepancy between ENDF-BVII and JENDL3.3 was decomposed by using sensitivity and uncertainty analysis technique. - Abstract: In the present work, we analyze the CREOLE experiment on the reactivity temperature coefficient (RTC) by using the three-dimensional continuous energy code (MCNP5) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA/Cadarache, was mainly dedicated to the RTC studies for both UO 2 and UO 2 -PuO 2 PWR type lattices covering the whole temperature range from 20 deg. C to 300 deg. C. We have developed an accurate 3D model of the EOLE reactor by using the MCNP5 Monte Carlo code which guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII, JEFF3.1 and JENDL3.3 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the RTC were analyzed and the results have shown that the JENDL3.3 and JEFF3.1 evaluations give the most consistent values; the discrepancy is

  10. OM Code Requirements For MOVs -- OMN-1 and Appendix III

    Energy Technology Data Exchange (ETDEWEB)

    Kevin G. DeWall

    2011-08-01

    The purpose or scope of the ASME OM Code is to establish the requirements for pre-service and in-service testing of nuclear power plant components to assess their operational readiness. For MOVs this includes those that perform a specific function in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, and mitigating the consequences of an accident. This paper will present a brief history of industry and regulatory activities related to MOVs and the development of Code requirements to address weaknesses in earlier versions of the OM Code. The paper will discuss the MOV requirements contained in the 2009 version of ASME OM Code, specifically Mandatory Appendix III and OMN-1, Revision 1.

  11. OM Code Requirements For MOVs -- OMN-1 and Appendix III

    International Nuclear Information System (INIS)

    DeWall, Kevin G.

    2011-01-01

    The purpose or scope of the ASME OM Code is to establish the requirements for pre-service and in-service testing of nuclear power plant components to assess their operational readiness. For MOVs this includes those that perform a specific function in shutting down a reactor to the safe shutdown condition, maintaining the safe shutdown condition, and mitigating the consequences of an accident. This paper will present a brief history of industry and regulatory activities related to MOVs and the development of Code requirements to address weaknesses in earlier versions of the OM Code. The paper will discuss the MOV requirements contained in the 2009 version of ASME OM Code, specifically Mandatory Appendix III and OMN-1, Revision 1.

  12. Pipe elbow stiffness coefficients including shear and bend flexibility factors for use in direct stiffness codes

    International Nuclear Information System (INIS)

    Perry, R.F.

    1977-01-01

    Historically, developments of computer codes used for piping analysis were based upon the flexibility method of structural analysis. Because of the specialized techniques employed in this method, the codes handled systems composed of only piping elements. Over the past ten years, the direct stiffness method has gained great popularity because of its systematic solution procedure regardless of the type of structural elements composing the system. A great advantage is realized with a direct stiffness code that combines piping elements along with other structural elements such as beams, plates, and shells, in a single model. One common problem, however, has been the lack of an accurate pipe elbow element that would adequately represent the effects of transverse shear and bend flexibility factors. The purpose of the present paper is to present a systematic derivation of the required 12x12 stiffness matrix and load vectors for a three dimensional pipe elbow element which includes the effects of transverse shear and pipe bend flexibility according to the ASME Boiler and Pressure Vessel Code, Section III. The results are presented analytically and as FORTRAN subroutines to be directly incorporated into existing direct stiffness codes. (Auth.)

  13. Analytical considerations in the code qualification of piping systems

    International Nuclear Information System (INIS)

    Antaki, G.A.

    1995-01-01

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

  14. Studies on Section XI ultrasonic repeatability

    International Nuclear Information System (INIS)

    Jamison, T.D.; McDearman, W.R.

    1981-05-01

    A block representative of a nuclear component has been welded containing intentional defects. Acoustic emission data taken during the welding correlate well with ultrasonic data. Repetitive ultrasonic examinations have been performed by skilled operators using a procedure based on that desribed in ASME Section XI. These examinations were performed by different examination teams using different ultrasonic equipment in such a manner that the effects on the repeatability of the ultrasonic test method caused by the operator and by the use of different equipment could be estimated. It was tentatively concluded that when considering a large number of inspections: (1) there is no significant difference in indication sizing between operators, and (2) there is a significant difference in amplitude and defect sizing when instruments having different, Code acceptable operating characteristics are used. It was determined that the Section XI sizing parameters follow a bivariate normal distribution. Data derived from ultrasonically and physically sizing indications in nuclear components during farication show that the Section XI technique tends to overestimate the size of the reflectors

  15. Cross-sectional association between ZIP code-level gentrification and homelessness among a large community-based sample of people who inject drugs in 19 US cities.

    Science.gov (United States)

    Linton, Sabriya L; Cooper, Hannah Lf; Kelley, Mary E; Karnes, Conny C; Ross, Zev; Wolfe, Mary E; Friedman, Samuel R; Jarlais, Don Des; Semaan, Salaam; Tempalski, Barbara; Sionean, Catlainn; DiNenno, Elizabeth; Wejnert, Cyprian; Paz-Bailey, Gabriela

    2017-06-20

    Housing instability has been associated with poor health outcomes among people who inject drugs (PWID). This study investigates the associations of local-level housing and economic conditions with homelessness among a large sample of PWID, which is an underexplored topic to date. PWID in this cross-sectional study were recruited from 19 large cities in the USA as part of National HIV Behavioral Surveillance. PWID provided self-reported information on demographics, behaviours and life events. Homelessness was defined as residing on the street, in a shelter, in a single room occupancy hotel, or in a car or temporarily residing with friends or relatives any time in the past year. Data on county-level rental housing unaffordability and demand for assisted housing units, and ZIP code-level gentrification (eg, index of percent increases in non-Hispanic white residents, household income, gross rent from 1990 to 2009) and economic deprivation were collected from the US Census Bureau and Department of Housing and Urban Development. Multilevel models evaluated the associations of local economic and housing characteristics with homelessness. Sixty percent (5394/8992) of the participants reported homelessness in the past year. The multivariable model demonstrated that PWID living in ZIP codes with higher levels of gentrification had higher odds of homelessness in the past year (gentrification: adjusted OR=1.11, 95% CI=1.04 to 1.17). Additional research is needed to determine the mechanisms through which gentrification increases homelessness among PWID to develop appropriate community-level interventions. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2017. All rights reserved. No commercial use is permitted unless otherwise expressly granted.

  16. 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....

  17. 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.

  18. 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.

  19. 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.

  20. Accelerator system model (ASM): A unique tool in exploring accelerator driven transmutation technologies (ADTT) system trade space

    Energy Technology Data Exchange (ETDEWEB)

    Myers, T.J.; Favale, A.J.; Berwald, D.H.; Burger, E.C.; Paulson, C.C.; Peacock, M.A.; Piaszczyk, C.M.; Piechowiak, E.M.; Rathke, J.W. [Northrop Grumman Corp., Bethpage, NY (United States). Advanced Technology and Development Center

    1997-09-01

    To aid in the development and optimization of emerging Accelerator Driven Transmutation Technology (ADTT) concepts, the Northrop Grumman Corporation, working together with G.H. Gillespie Associates and Los Alamos National Laboratory has developed a computational tool which combines both accelerator physics layout/analysis capabilities with engineering analysis capabilities to create a standardized platform to compare and contrast accelerator system configurations. In this context, the accelerator system configuration includes not only the accelerating structures, but also the major support systems such as the vacuum, thermal control, RF power, and cryogenic subsystem (if superconducting accelerator operation is investigated) as well as estimates of the costs for enclosures (accelerating tunnel and RF halls). This paper presents an overview of the Accelerator System Model (ASM) code flow, as well as a discussion of the data and analysis upon which it is based. Also presented is material which addresses the development of the evaluation criteria employed by this code including a presentation of the economic analysis methods, and a discussion of the cost database employed. The paper concludes with examples depicting completed and planned trade studies for both normal and superconducting accelerator applications. 8 figs.

  1. 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.

  2. 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

  3. 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

  4. Comprehensive Report For Proposed Elevated Temperature Elastic Perfectly Plastic (EPP) Code Cases Representative Example Problems

    Energy Technology Data Exchange (ETDEWEB)

    Hollinger, Greg L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-06-01

    Background: The current rules in the nuclear section of the ASME Boiler and Pressure Vessel (B&PV) Code , Section III, Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200F (650C)1. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (E-PP) analysis methods and which are expected to be applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature 2, 3, and have been recently revised to incorporate comments and simplify their application. The revised code cases have been developed. Task Objectives: The goal of the Sample Problem task is to exercise these code cases through example problems to demonstrate their feasibility and, also, to identify potential corrections and improvements should problems be encountered. This will provide input to the development of technical background documents for consideration by the applicable B&PV committees considering these code cases for approval. This task has been performed by Hollinger and Pease of Becht Engineering Co., Inc., Nuclear Services Division and a report detailing the results of the E-PP analyses conducted on example problems per the procedures of the E-PP strain limits and creep-fatigue draft code cases is enclosed as Enclosure 1. Conclusions: The feasibility of the application of the E-PP code cases has been demonstrated through example problems that consist of realistic geometry (a nozzle attached to a semi-hemispheric shell with a circumferential weld) and load (pressure; pipe reaction load applied at the end of the nozzle, including axial and shear forces, bending and torsional moments; through-wall transient temperature gradient) and design and operating conditions (Levels A, B and C).

  5. 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.

  6. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  7. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  8. Coincidence: Fortran code for calculation of (e, e'x) differential cross-sections, nuclear structure functions and polarization asymmetry in self-consistent random phase approximation with Skyrme interaction

    Energy Technology Data Exchange (ETDEWEB)

    Cavinato, M.; Marangoni, M.; Saruis, A.M.

    1990-10-01

    This report describes the COINCIDENCE code written for the IBM 3090/300E computer in Fortran 77 language. The output data of this code are the (e, e'x) threefold differential cross-sections, the nuclear structure functions, the polarization asymmetry and the angular correlation coefficients. In the real photon limit, the output data are the angular distributions for plane polarized incident photons. The code reads from tape the transition matrix elements previously calculated, by in continuum self-consistent RPA (random phase approximation) theory with Skyrme interactions. This code has been used to perform a numerical analysis of coincidence (e, e'x) reactions with polarized electrons on the /sup 16/O nucleous.

  9. Use of Neuber's rule to estimate the fatigue life of notched specimens of ASME SA 106-B steel piping in 2880C air

    International Nuclear Information System (INIS)

    Terrell, J.B.

    1989-01-01

    Fatigue strain-life tests were conducted on notched specimens of ADMESA 106-B piping steel at PWR operating temperatures (288 0 C (550 0 F)), under completely reversed loading. Fatigue limits at 10 7 cycles were estimated for smooth specimens to be 185 M Pa (26.8 ksi) at 24 0 C and 232 MPa (33.7 ksi) at 288 0 C. The higher fatigue strength observed at the PWR temperature is postulated to be caused by dynamic strain aging processes. However, a reduction in fatigue strength in the low cycle fatigue regime was observed in 288 0 C air environment tests, which may indicate that the current ASME Section III design curve for carbon steels is nonconservative in its positioning. Notch strain histories were estimated for the notched specimen tests using various interpretations of Neuber's rule. It was concluded that the use of the fatigue notch concentration factor (K f ) in the Neuber relation in conjunction with the uniaxial cyclic stress-strain curve provided the best correlation of notched specimen fatigue data with results obtained from smooth specimen tests. The notched specimen strain-life results derived from the application of Neuber's rule alone proved to be conservative when compared with smooth specimen test results to such an extent that Neuber-generated notch stresses and strain amplitudes cannot accurately be compared with the mean data curves derived from the ASME Section III fatigue curves for carbon steels which are based on net section stress measurements. (author)

  10. Simulation and optimization of a coking wastewater biological treatment process by activated sludge models (ASM).

    Science.gov (United States)

    Wu, Xiaohui; Yang, Yang; Wu, Gaoming; Mao, Juan; Zhou, Tao

    2016-01-01

    Applications of activated sludge models (ASM) in simulating industrial biological wastewater treatment plants (WWTPs) are still difficult due to refractory and complex components in influents as well as diversity in activated sludges. In this study, an ASM3 modeling study was conducted to simulate and optimize a practical coking wastewater treatment plant (CWTP). First, respirometric characterizations of the coking wastewater and CWTP biomasses were conducted to determine the specific kinetic and stoichiometric model parameters for the consecutive aeration-anoxic-aeration (O-A/O) biological process. All ASM3 parameters have been further estimated and calibrated, through cross validation by the model dynamic simulation procedure. Consequently, an ASM3 model was successfully established to accurately simulate the CWTP performances in removing COD and NH4-N. An optimized CWTP operation condition could be proposed reducing the operation cost from 6.2 to 5.5 €/m(3) wastewater. This study is expected to provide a useful reference for mathematic simulations of practical industrial WWTPs. Copyright © 2015 Elsevier Ltd. All rights reserved.

  11. 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)

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

    NARCIS (Netherlands)

    Luger, T.; Leent, E.J. van; 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.; Valk, P.G.M. van der; Bos, J.D.

    2001-01-01

    BACKGROUND: 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. OBJECTIVES: This study was designed to determine

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

    NARCIS (Netherlands)

    Luger, T; Van Leent, EJM; Graeber, M; Hedgecock, S; Thurston, M; Kandra, A; Berth-Jones, J; Bjerke, J; Christophers, E; Knulst, AC; Morren, M; Morris, A; Reitamo, S; Roed-Petersen, J; Schoepf, E; Thestrup-Pedersen, K; van der Valk, P. G. M.; Bos, JD

    Background 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. Objectives This study was: designed to determine

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

    NARCIS (Netherlands)

    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.

    2001-01-01

    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

  15. Precipitation-hardening stainless steel bars, shapes, and forgings (ASME SA-564 with additional requirements)

    International Nuclear Information System (INIS)

    1975-05-01

    A standard prescribing requirements for precipitation-hardening stainless steel bars, shapes, and forgings (ASME SA-564 with additional requirements) for nuclear and associated applications is presented. This standard supersedes RDT M 7-6T, dated January 1974. (U.S.)

  16. Technical report on comparative analysis of ASME QA requirements and ISO series

    International Nuclear Information System (INIS)

    Kim, Kwan Hyun

    2000-06-01

    This technical report provides the differences on the QA requirement ASME and ISO in nuclear fields. This report applies to the quality assurance(QA) programmes of the design of two requirement. The organization having overall responsibility for the nuclear design, preservation, fabrication shall be described in this report in each stage of design project

  17. Roles of the outer membrane protein AsmA of Salmonella enterica in the control of marRAB expression and invasion of epithelial cells.

    Science.gov (United States)

    Prieto, Ana I; Hernández, Sara B; Cota, Ignacio; Pucciarelli, M Graciela; Orlov, Yuri; Ramos-Morales, Francisco; García-del Portillo, Francisco; Casadesús, Josep

    2009-06-01

    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. enterica, suggesting that lack of AsmA activates expression of the marRAB operon. Hence, asmA mutations may enhance bile resistance by inducing gene expression changes in the marRAB-controlled Mar regulon. In silico analysis of AsmA structure predicted the existence of one transmembrane domain. Biochemical analysis of subcellular fractions revealed that the asmA gene of S. enterica encodes a protein of approximately 70 kDa located in the outer membrane. Because AsmA is unrelated to known transport and/or efflux systems, we propose that activation of marRAB in asmA mutants may be a consequence of envelope reorganization. Competitive infection of BALB/c mice with asmA(+) and asmA isogenic strains indicated that lack of AsmA attenuates Salmonella virulence by the oral route but not by the intraperitoneal route. Furthermore, asmA mutants showed a reduced ability to invade epithelial cells in vitro.

  18. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Crecy, F. de; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  19. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    International Nuclear Information System (INIS)

    Ounsy, A.; Brun, B.

    1993-01-01

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs

  20. The adjoint sensitivity method. A contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  1. The adjoint sensitivity method, a contribution to the code uncertainty evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ounsy, A; Crecy, F de; Brun, B

    1994-12-31

    The application of the ASM (Adjoint Sensitivity Method) to thermohydraulic codes, is examined. The advantage of the method is to be very few CPU time consuming in comparison with usual approach requiring one complete code run per sensitivity determination. The mathematical aspects of the problem are first described, and the applicability of the method of the functional-type response of a thermalhydraulic model is demonstrated. On a simple example of non linear hyperbolic equation (Burgers equation) the problem has been analyzed. It is shown that the formalism used in the literature treating this subject is not appropriate. A new mathematical formalism circumventing the problem is proposed. For the discretized form of the problem, two methods are possible: the Continuous ASM and the Discrete ASM. The equivalence of both methods is demonstrated; nevertheless only the DASM constitutes a practical solution for thermalhydraulic codes. The application of the DASM to the thermalhydraulic safety code CATHARE is then presented for two examples. They demonstrate that ASM constitutes an efficient tool for the analysis of code sensitivity. (authors) 7 figs., 5 tabs., 8 refs.

  2. Statistical analysis of geomagnetic field intensity differences between ASM and VFM instruments onboard Swarm constellation

    Science.gov (United States)

    De Michelis, Paola; Tozzi, Roberta; Consolini, Giuseppe

    2017-02-01

    From the very first measurements made by the magnetometers onboard Swarm satellites launched by European Space Agency (ESA) in late 2013, it emerged a discrepancy between scalar and vector measurements. An accurate analysis of this phenomenon brought to build an empirical model of the disturbance, highly correlated with the Sun incidence angle, and to correct vector data accordingly. The empirical model adopted by ESA results in a significant decrease in the amplitude of the disturbance affecting VFM measurements so greatly improving the vector magnetic data quality. This study is focused on the characterization of the difference between magnetic field intensity measured by the absolute scalar magnetometer (ASM) and that reconstructed using the vector field magnetometer (VFM) installed on Swarm constellation. Applying empirical mode decomposition method, we find the intrinsic mode functions (IMFs) associated with ASM-VFM total intensity differences obtained with data both uncorrected and corrected for the disturbance correlated with the Sun incidence angle. Surprisingly, no differences are found in the nature of the IMFs embedded in the analyzed signals, being these IMFs characterized by the same dominant periodicities before and after correction. The effect of correction manifests in the decrease in the energy associated with some IMFs contributing to corrected data. Some IMFs identified by analyzing the ASM-VFM intensity discrepancy are characterized by the same dominant periodicities of those obtained by analyzing the temperature fluctuations of the VFM electronic unit. Thus, the disturbance correlated with the Sun incidence angle could be still present in the corrected magnetic data. Furthermore, the ASM-VFM total intensity difference and the VFM electronic unit temperature display a maximal shared information with a time delay that depends on local time. Taken together, these findings may help to relate the features of the observed VFM-ASM total intensity

  3. Review and comparison of WWER and LWR Codes and Standards

    International Nuclear Information System (INIS)

    Buckthorpe, D.; Tashkinov, A.; Brynda, J.; Davies, L.M.; Cueto-Felgeueroso, C.; Detroux, P.; Bieniussa, K.; Guinovart, J.

    2003-01-01

    The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)

  4. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  5. Comparison of design margin for core shroud in between design and construction code and fitness-for-service code

    International Nuclear Information System (INIS)

    Dozaki, Koji

    2007-01-01

    Structural design methods for core shroud of BWR are specified in JSME Design and Construction Code, like ASME Boiler and Pressure Vessel Code Sec. III, as a part of core support structure. Design margins are defined according to combination of the structural design method selected and service limit considered. Basically, those margins in JSME Code were determined after ASME Sec. III. Designers can select so-called twice-slope method for core shroud design among those design methods. On the other hand, flaw evaluation rules have been established for core shroud in JSME Fitness-for-Service Code. Twice-slope method is also adopted for fracture evaluation in that code even when the core shroud contains a flaw. Design margin was determined as structural factors separately from Design and Construction Code. As a natural consequence, there is a difference in those design margins between the two codes. In this paper, it is shown that the design margin in Fitness-for-Service Code is conservative by experimental evidences. Comparison of design margins between the two codes is discussed. (author)

  6. Aquelarre. A computer code for fast neutron cross sections from the statistical model; AQUELARRE. Un programa numerico para el calculo de secciones eficaces neutronicas mediante el modelo de nucleo compuesto

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J.

    1974-07-01

    A Fortran V computer code for Univac 1108/6 using the partial statistical (or compound nucleus) model is described. The code calculates fast neutron cross sections for the (n, n'), (n, p), (n, d) and (n, {alpha}) reactions and the angular distributions and Legendre moments for the (n, n) and (n, n') processes in heavy and intermediate spherical nuclei. A local Optical Model with spin-orbit interaction for each level is employed, allowing for the width fluctuation and Moldauer corrections, as well as the inclusion of discrete and continuous levels. (Author) 67 refs.

  7. Three-Dimensional (X,Y,Z) Deterministic Analysis of the PCA-Replica Neutron Shielding Benchmark Experiment using the TORT-3.2 Code and Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    OpenAIRE

    Pescarini Massimo; Orsi Roberto; Frisoni Manuela

    2016-01-01

    The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...

  8. Aquelarre. A computer code for fast neutron cross sections from the statistical model; AQUELARRE. Un programa numerico para el calculo de secciones eficaces neutronicas mediante el modelo de nucleo compuesto

    Energy Technology Data Exchange (ETDEWEB)

    Guasp, J

    1974-07-01

    A Fortran V computer code for Univac 1108/6 using the partial statistical (or compound nucleus) model is described. The code calculates fast neutron cross sections for the (n, n'), (n, p), (n, d) and (n, {alpha}) reactions and the angular distributions and Legendre moments for the (n, n) and (n, n') processes in heavy and intermediate spherical nuclei. A local Optical Model with spin-orbit interaction for each level is employed, allowing for the width fluctuation and Moldauer corrections, as well as the inclusion of discrete and continuous levels. (Author) 67 refs.

  9. Application of nuclear air cleaning and treatment codes

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1995-01-01

    All modifications to existing ventilation systems, as well as any new ventilation systems used on the Hanford Site are required to meet both American Society of Mechanical Engineers (ASME) codes N509 and N510. Difficulties encountered when applying code N509 at the Hanford Site include the composition of the ventilation air stream and requirements related to ventilation equipment procurement. Also, the existing ventilation systems for the waste tanks at the Hanford Site cannot be tested in accordance with code N510 because of the current configuration of these systems

  10. Application of nuclear air cleaning and treatment codes

    Energy Technology Data Exchange (ETDEWEB)

    Kriskovich, J.R. [Westinghouse Hanford Company, Richland, WA (United States)

    1995-02-01

    All modifications to existing ventilation systems, as well as any new ventilation systems used on the Hanford Site are required to meet both American Society of Mechanical Engineers (ASME) codes N509 and N510. Difficulties encountered when applying code N509 at the Hanford Site include the composition of the ventilation air stream and requirements related to ventilation equipment procurement. Also, the existing ventilation systems for the waste tanks at the Hanford Site cannot be tested in accordance with code N510 because of the current configuration of these systems.

  11. Codification of LMFBR rules and comparison of codes

    International Nuclear Information System (INIS)

    Faure, O.; Debaene, J.P.

    1993-01-01

    The first part of this report presents the basic RCC-MR (regles de conception et de construction des materiels mecaniques des ilots nucleaires, reacteurs a neutrons rapides) design rules and their purpose. The second part is a qualitative comparison between RCC-MR, Code case N47 (ASME) and ETSDG Guide (MONJU Guide), made on the following topics: negligible creep test, ratcheting, creep fatigue, buckling, piping rules. An outline is given on improvements to RCC-MR rules now in progress

  12. PROSPERE (SPM 211) - A code for spectrum and effective cross-section calculations in a cell with several media and with one or two moderators

    International Nuclear Information System (INIS)

    Tran Tuc Vi

    1969-12-01

    The PROSPERE code uses the CADILHAC model for neutron thermalization and the so-called 'zone-source' method for space treatment. First flight collision probabilities are calculated rapidly enough to allow fuel to be divided in concentric annuli. The ABH method is still used in the moderator (except in the case of thin moderators). Two moderators can be treated: one of them can be introduced in any media, eventually with variable densities. The PROSPERE code simplifies energy and space treatment and, as such, brings considerable computer-time savings with respect to THERMOS, in most cases with an accuracy of the same order. (author) [fr

  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. 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)

  15. Meeting report on the ASM Conference on Mechanisms of Interbacterial Cooperation and Competition.

    Science.gov (United States)

    Lories, Bram; Parijs, Ilse; Foster, Kevin R; Steenackers, Hans P

    2017-08-14

    The ASM Conference on Mechanisms of Interbacterial Cooperation and Competition was held in Washington DC, from 1 to 4 March 2017. The conference provided an international forum for sociomicrobiologists from different disciplines to present and discuss new findings. The meeting covered a wide range of topics, spanning molecular mechanisms, ecology, evolution, computation and manipulation of interbacterial interactions, and encompassed social communities in medicine, the natural environment, and industry. This report summarizes the presentations and emerging themes. Copyright © 2017 American Society for Microbiology.

  16. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

  17. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

  18. Conservatism of ASME KIR-reference curve with respect to crack arrest

    International Nuclear Information System (INIS)

    Wallin, K.; Rintamaa, R.; Nagel, G.

    1999-01-01

    The conservatism of the RT NDT temperature indexing parameter and the ASME K IR -reference curve with respect to crack arrest toughness, has been evaluated. Based on an analysis of the original ASME K Ia data, it was established that inherently, the ASME K IR -reference curve corresponds to an overall 5% lower bound curve with respect to crack arrest. It was shown that the scatter of crack arrest toughness is essentially material independent and has a standard deviation of 18% and the temperature dependence of K Ia has the same form as predicted by the master curve for crack initiation toughness. The 'built in' offset between the mean 100 MPa√(m) crack arrest temperature, TK Ia , and RT NDT is 38 C (TK Ia =RT NDT +38 C) and the experimental relation between TK Ia and NDT is, TK Ia =NDT+28 C. The K IR -reference curve using NDT as reference temperature will be conservative with respect to the general 5% lower bound K Ia(5%) -curve, with a 75% confidence. The use of RT NDT , instead of NDT, will generally increase the degree of conservatism, both for non-irradiated as well as irradiated materials, close to a 95% confidence level. This trend is pronounced for materials with Charpy-V upper shelf energies below 100 J. It is shown that the K IR -curve effectively constitutes a deterministic lower bound curve for crack arrest. The findings are valid both for nuclear pressure vessel plates, forgings and welds. (orig.)

  19. Statistical re-evaluation of the ASME KIC and KIR fracture toughness reference curves

    International Nuclear Information System (INIS)

    Wallin, K.

    1999-01-01

    Historically the ASME reference curves have been treated as representing absolute deterministic lower bound curves of fracture toughness. In reality, this is not the case. They represent only deterministic lower bound curves to a specific set of data, which represent a certain probability range. A recently developed statistical lower bound estimation method called the 'master curve', has been proposed as a candidate for a new lower bound reference curve concept. From a regulatory point of view, the master curve is somewhat problematic in that it does not claim to be an absolute deterministic lower bound, but corresponds to a specific theoretical failure probability that can be chosen freely based on application. In order to be able to substitute the old ASME reference curves with lower bound curves based on the master curve concept, the inherent statistical nature (and confidence level) of the ASME reference curves must be revealed. In order to estimate the true inherent level of safety, represented by the reference curves, the original database was re-evaluated with statistical methods and compared to an analysis based on the master curve concept. The analysis reveals that the 5% lower bound master curve has the same inherent degree of safety as originally intended for the K IC -reference curve. Similarly, the 1% lower bound master curve corresponds to the K IR -reference curve. (orig.)

  20. Statistical re-evaluation of the ASME KIC and KIR fracture toughness reference curves

    International Nuclear Information System (INIS)

    Wallin, K.; Rintamaa, R.

    1998-01-01

    Historically the ASME reference curves have been treated as representing absolute deterministic lower bound curves of fracture toughness. In reality, this is not the case. They represent only deterministic lower bound curves to a specific set of data, which represent a certain probability range. A recently developed statistical lower bound estimation method called the 'Master curve', has been proposed as a candidate for a new lower bound reference curve concept. From a regulatory point of view, the Master curve is somewhat problematic in that it does not claim to be an absolute deterministic lower bound, but corresponds to a specific theoretical failure probability that can be chosen freely based on application. In order to be able to substitute the old ASME reference curves with lower bound curves based on the master curve concept, the inherent statistical nature (and confidence level) of the ASME reference curves must be revealed. In order to estimate the true inherent level of safety, represented by the reference curves, the original data base was re-evaluated with statistical methods and compared to an analysis based on the master curve concept. The analysis reveals that the 5% lower bound Master curve has the same inherent degree of safety as originally intended for the K IC -reference curve. Similarly, the 1% lower bound Master curve corresponds to the K IR -reference curve. (orig.)

  1. Pressure vessel code construction capabilities for a nickel-chromium-tungsten-molybdenum alloy

    International Nuclear Information System (INIS)

    Rothman, M.F.

    1990-01-01

    HAYNES alloy 230 (UNS NO6230) has achieved wide usage in a variety of high-temperature aerospace, chemical process industry and industrial heating applications since its introduction in 1981. Combining high elevated temperature strength with excellent metallurgical stability, environment-resistance and relatively straight forward fabrication characteristics, this Ni-Cr-W-Mo alloy was an excellent candidate for ASME Pressure vessel Code applications. Coverage under case No. 2063 was granted in July, 1989, for both Section I and Section VIII Division 1 construction. In this paper, the metallurgy of 230 alloy will be described, and its design strength capabilities contrasted with those for more established code materials. Other important performance capabilities, such as long-term thermal stability, oxidation-resistance, fatigue-resistance, and resistance to other forms of environmental degradation will be discussed. It will be shown that the combined properties of 230 alloy offer some significant advantages over other materials for applications such as expansion bellows, heat-exchangers, valves and other components in the fossil energy, nuclear energy and chemical process industries, among others

  2. 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

  3. Comparison of SKIFS 2004:1 and Tillsynshandbok PSA against the ASME PRA Standard and European requirements on PSA

    International Nuclear Information System (INIS)

    Hellstroem, Per

    2005-04-01

    Requirements on PSA for risk informed applications are expressed in different international documents. The ASME PRA standard published in spring 2002 is one such document, PSA requirements are also expressed in the European Utility Requirements (EUR) for new reactors. The Swedish PSA requirements are provided in the Swedish regulators (SKI) statutes SKIFS 2004:1. SKI also has a review handbook for PSA activities (SKI report 2003:48). The review handbook is a support during review of the utilities PSA activities and the PSAs themselves. The review handbook expresses SKIs expectations by providing so called important aspects for both the PSA work and the PSAs, A comparison of SKIFS requirements and the important aspects in the Review handbook, on one side, and the requirements on PSA in EUR and ASME on the other side, is presented. The comparison shows a large difference in the level of detail in the different documents, where ASME is most detailed and specific. This is expected since the SKI review handbook not is a 'PSA guide' in the same way as the ASME PRA standard. A direct comparison of the ASME PRA standard requirements with the important aspects in the review handbook cannot answer the question which ASME capacity level that is achieved by a PSA meeting all important aspects. The conclusion is that it is not likely to achieve capacity level 2 and 3, since very few ASME level 3 attributes are explicitly expressed as important aspects, though many are expressed in general terms. The review handbook important aspects that are most similar to the ASME capacity level 1 attributes are initiating events, sequence analysis, and system analysis while less similarity is found for analysis of operator actions data analysis, quantification and containment analysis (level 2). Less similarity is found for capacity level 2 and 3. However, the number of additional ASME attributes on capacity level 2 and 3 are few. There are also important aspects in the review handbook that

  4. ASM-3 acid sphingomyelinase functions as a positive regulator of the DAF-2/AGE-1 signaling pathway and serves as a novel anti-aging target.

    Science.gov (United States)

    Kim, Yongsoon; Sun, Hong

    2012-01-01

    In C. elegans, the highly conserved DAF-2/insulin/insulin-like growth factor 1 receptor signaling (IIS) pathway regulates longevity, metabolism, reproduction and development. In mammals, acid sphingomyelinase (ASM) is an enzyme that hydrolyzes sphingomyelin to produce ceramide. ASM has been implicated in CD95 death receptor signaling under certain stress conditions. However, the involvement of ASM in growth factor receptor signaling under physiological conditions is not known. Here, we report that in vivo ASM functions as a positive regulator of the DAF-2/IIS pathway in C. elegans. We have shown that inactivation of asm-3 extends animal lifespan and promotes dauer arrest, an alternative developmental process. A significant cooperative effect on lifespan is observed between asm-3 deficiency and loss-of-function alleles of the age-1/PI 3-kinase, with the asm-3; age-1 double mutant animals having a mean lifespan 259% greater than that of the wild-type animals. The lifespan extension phenotypes caused by the loss of asm-3 are dependent on the functions of daf-16/FOXO and daf-18/PTEN. We have demonstrated that inactivation of asm-3 causes nuclear translocation of DAF-16::GFP protein, up-regulates endogenous DAF-16 protein levels and activates the downstream targeting genes of DAF-16. Together, our findings reveal a novel role of asm-3 in regulation of lifespan and diapause by modulating IIS pathway. Importantly, we have found that two drugs known to inhibit mammalian ASM activities, desipramine and clomipramine, markedly extend the lifespan of wild-type animals, in a manner similar to that achieved by genetic inactivation of the asm genes. Our studies illustrate a novel strategy of anti-aging by targeting ASM, which may potentially be extended to mammals.

  5. ASM-3 acid sphingomyelinase functions as a positive regulator of the DAF-2/AGE-1 signaling pathway and serves as a novel anti-aging target.

    Directory of Open Access Journals (Sweden)

    Yongsoon Kim

    Full Text Available In C. elegans, the highly conserved DAF-2/insulin/insulin-like growth factor 1 receptor signaling (IIS pathway regulates longevity, metabolism, reproduction and development. In mammals, acid sphingomyelinase (ASM is an enzyme that hydrolyzes sphingomyelin to produce ceramide. ASM has been implicated in CD95 death receptor signaling under certain stress conditions. However, the involvement of ASM in growth factor receptor signaling under physiological conditions is not known. Here, we report that in vivo ASM functions as a positive regulator of the DAF-2/IIS pathway in C. elegans. We have shown that inactivation of asm-3 extends animal lifespan and promotes dauer arrest, an alternative developmental process. A significant cooperative effect on lifespan is observed between asm-3 deficiency and loss-of-function alleles of the age-1/PI 3-kinase, with the asm-3; age-1 double mutant animals having a mean lifespan 259% greater than that of the wild-type animals. The lifespan extension phenotypes caused by the loss of asm-3 are dependent on the functions of daf-16/FOXO and daf-18/PTEN. We have demonstrated that inactivation of asm-3 causes nuclear translocation of DAF-16::GFP protein, up-regulates endogenous DAF-16 protein levels and activates the downstream targeting genes of DAF-16. Together, our findings reveal a novel role of asm-3 in regulation of lifespan and diapause by modulating IIS pathway. Importantly, we have found that two drugs known to inhibit mammalian ASM activities, desipramine and clomipramine, markedly extend the lifespan of wild-type animals, in a manner similar to that achieved by genetic inactivation of the asm genes. Our studies illustrate a novel strategy of anti-aging by targeting ASM, which may potentially be extended to mammals.

  6. ASM-3 Acid Sphingomyelinase Functions as a Positive Regulator of the DAF-2/AGE-1 Signaling Pathway and Serves as a Novel Anti-Aging Target

    Science.gov (United States)

    Kim, Yongsoon; Sun, Hong

    2012-01-01

    In C. elegans, the highly conserved DAF-2/insulin/insulin-like growth factor 1 receptor signaling (IIS) pathway regulates longevity, metabolism, reproduction and development. In mammals, acid sphingomyelinase (ASM) is an enzyme that hydrolyzes sphingomyelin to produce ceramide. ASM has been implicated in CD95 death receptor signaling under certain stress conditions. However, the involvement of ASM in growth factor receptor signaling under physiological conditions is not known. Here, we report that in vivo ASM functions as a positive regulator of the DAF-2/IIS pathway in C. elegans. We have shown that inactivation of asm-3 extends animal lifespan and promotes dauer arrest, an alternative developmental process. A significant cooperative effect on lifespan is observed between asm-3 deficiency and loss-of-function alleles of the age-1/PI 3-kinase, with the asm-3; age-1 double mutant animals having a mean lifespan 259% greater than that of the wild-type animals. The lifespan extension phenotypes caused by the loss of asm-3 are dependent on the functions of daf-16/FOXO and daf-18/PTEN. We have demonstrated that inactivation of asm-3 causes nuclear translocation of DAF-16::GFP protein, up-regulates endogenous DAF-16 protein levels and activates the downstream targeting genes of DAF-16. Together, our findings reveal a novel role of asm-3 in regulation of lifespan and diapause by modulating IIS pathway. Importantly, we have found that two drugs known to inhibit mammalian ASM activities, desipramine and clomipramine, markedly extend the lifespan of wild-type animals, in a manner similar to that achieved by genetic inactivation of the asm genes. Our studies illustrate a novel strategy of anti-aging by targeting ASM, which may potentially be extended to mammals. PMID:23049887

  7. Evaluation of clinical coding data to determine causes of critical bleeding in patients receiving massive transfusion: a bi-national, multicentre, cross-sectional study.

    Science.gov (United States)

    McQuilten, Z K; Zatta, A J; Andrianopoulos, N; Aoki, N; Stevenson, L; Badami, K G; Bird, R; Cole-Sinclair, M F; Hurn, C; Cameron, P A; Isbister, J P; Phillips, L E; Wood, E M

    2017-04-01

    To evaluate the use of routinely collected data to determine the cause(s) of critical bleeding in patients who receive massive transfusion (MT). Routinely collected data are increasingly being used to describe and evaluate transfusion practice. Chart reviews were undertaken on 10 randomly selected MT patients at 48 hospitals across Australia and New Zealand to determine the cause(s) of critical bleeding. Diagnosis-related group (DRG) and International Classification of Diseases (ICD) codes were extracted separately and used to assign each patient a cause of critical bleeding. These were compared against chart review using percentage agreement and kappa statistics. A total of 427 MT patients were included with complete ICD and DRG data for 427 (100%) and 396 (93%), respectively. Good overall agreement was found between chart review and ICD codes (78·3%; κ = 0·74, 95% CI 0·70-0·79) and only fair overall agreement with DRG (51%; κ = 0·45, 95% CI 0·40-0·50). Both ICD and DRG were sensitive and accurate for classifying obstetric haemorrhage patients (98% sensitivity and κ > 0·94). However, compared with the ICD algorithm, DRGs were less sensitive and accurate in classifying bleeding as a result of gastrointestinal haemorrhage (74% vs 8%; κ = 0·75 vs 0·1), trauma (92% vs 62%; κ = 0·78 vs 0·67), cardiac (80% vs 57%; κ = 0·79 vs 0·60) and vascular surgery (64% vs 56%; κ = 0·69 vs 0·65). Algorithms using ICD codes can determine the cause of critical bleeding in patients requiring MT with good to excellent agreement with clinical history. DRG are less suitable to determine critical bleeding causes. © 2016 British Blood Transfusion Society.

  8. Neurospora crassa ASM-1 complements the conidiation defect in a stuA mutant of Aspergillus nidulans.

    Science.gov (United States)

    Chung, Dawoon; Upadhyay, Srijana; Bomer, Brigitte; Wilkinson, Heather H; Ebbole, Daniel J; Shaw, Brian D

    2015-01-01

    Aspergillus nidulans StuA and Neurospora crassa ASM-1 are orthologous APSES (ASM-1, PHD1, SOK2, Efg1, StuA) transcription factors conserved across a diverse group of fungi. StuA and ASM-1 have roles in asexual (conidiation) and sexual (ascospore formation) development in both organisms. To address the hypothesis that the last common ancestor of these diverse fungi regulated conidiation with similar genes, asm-1 was introduced into the stuA1 mutant of A. nidulans. Expression of asm-1 complemented defective conidiophore morphology and restored conidia production to wild type levels in stuA1. Expression of asm-1 in the stuA1 strain did not rescue the defect in sexual development. When the conidiation regulator AbaA was tagged at its C-terminus with GFP in A. nidulans, it localized to nuclei in phialides. When expressed in the stuA1 mutant, AbaA::GFP localized to nuclei in conidiophores but no longer was confined to phialides, suggesting that expression of AbaA in specific cell types of the conidiophore was conditioned by StuA. Our data suggest that the function in conidiation of StuA and ASM-1 is conserved and support the view that, despite the great morphological and ontogenic diversity of their condiphores, the last common ancestor of A. nidulans and N. crassa produced an ortholog of StuA that was involved in conidiophore development. © 2015 by The Mycological Society of America.

  9. Bronchodilatory and anti-inflammatory effects of ASM-024, a nicotinic receptor ligand, developed for the treatment of asthma.

    Science.gov (United States)

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

    2014-01-01

    Conventional asthma and COPD treatments include the use of bronchodilators, mainly β2-adrenergic agonists, muscarinic receptor antagonists and corticosteroids or leukotriene antagonists as anti-inflammatory agents. These active drugs are administered either separately or given as a fixed-dose combination medication into a single inhaler. ASM-024, a homopiperazinium compound, derived from the structural modification of diphenylmethylpiperazinium (DMPP), has been developed to offer an alternative mechanism of action that could provide symptomatic control through combined anti-inflammatory and bronchodilator properties in a single entity. A dose-dependent inhibition of cellular inflammation in bronchoalveolar lavage fluid was observed in ovalbumin-sensitized mice, subsequently treated for 3 days by nose-only exposure with aerosolized ASM-024 at doses up to 3.8 mg/kg (ED50 = 0.03 mg/kg). The methacholine ED250 values indicated that airway hyperresponsivenness (AHR) to methacholine decreased following ASM-024 administration by inhalation at a dose of 1.5 mg/kg, with a value of 0.145 ± 0.032 mg/kg for ASM 024-treated group as compared to 0.088 ± 0.023 mg/kg for untreated mice. In in vitro isometric studies, ASM-024 elicited dose-dependent relaxation of isolated mouse tracheal, human, and dog bronchial preparations contracted with methacholine and guinea pig tracheas contracted with histamine. ASM-024 showed also a dose and time dependant protective effect on methacholine-induced contraction. Overall, with its combined anti-inflammatory, bronchodilating and bronchoprotective properties, ASM-024 may represent a new class of drugs with a novel pharmacological approach that could prove useful for the chronic maintenance treatment of asthma and, possibly, COPD.

  10. Decree No. 1900 of 24 August 1989 authorizing civil divorce before a notary for reasons provided in Section 8 of Article 154 of the Civil Code.

    Science.gov (United States)

    1989-01-01

    This Colombian Decree authorizes civil divorce to take place before a notary by means of public document when the spouses are in mutual agreement; the ground for the divorce is separation, either decreed by a court or formalized before a notary; and the separation has lasted more than 2 years. Such a divorce produces the same legal effects as a divorce decreed by a court. The document is to set forth the duties of both spouses with respect to the provisions of sentence 3 of Article 166 of the Civil Code. If there are minors, the document must receive the approval of the municipal or district representative. Decree No. 2275 of 7 October 1989 (Diario Oficial, 7 October 1989, pp. 75-76) amends this Decree to provide that the judgment of the municipal or district representative must be made within 10 days after the document is received. Sentence 3 of Article 166 of the Civil Code requires that spouses who are separating by mutual consent send to a judge their agreement with respect to the care and support of children, which the judge may reject if necessary in the interests of the children.

  11. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    Energy Technology Data Exchange (ETDEWEB)

    Beckert, C.

    2007-12-19

    Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer

  12. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    International Nuclear Information System (INIS)

    White, Morgan C.

    2000-01-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

  13. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

    2000-07-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

  14. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  15. The ASM-NSF Biology Scholars Program: An Evidence-Based Model for Faculty Development.

    Science.gov (United States)

    Chang, Amy L; Pribbenow, Christine M

    2016-05-01

    The American Society for Microbiology (ASM) established its ASM-NSF (National Science Foundation) Biology Scholars Program (BSP) to promote undergraduate education reform by 1) supporting biologists to implement evidence-based teaching practices, 2) engaging life science professional societies to facilitate biologists' leadership in scholarly teaching within the discipline, and 3) participating in a teaching community that fosters disciplinary-level science, technology, engineering, and mathematics (STEM) reform. Since 2005, the program has utilized year-long residency training to provide a continuum of learning and practice centered on principles from the scholarship of teaching and learning (SoTL) to more than 270 participants ("scholars") from biology and multiple other disciplines. Additionally, the program has recruited 11 life science professional societies to support faculty development in SoTL and discipline-based education research (DBER). To identify the BSP's long-term outcomes and impacts, ASM engaged an external evaluator to conduct a study of the program's 2010-2014 scholars (n = 127) and society partners. The study methods included online surveys, focus groups, participant observation, and analysis of various documents. Study participants indicate that the program achieved its proposed goals relative to scholarship, professional society impact, leadership, community, and faculty professional development. Although participants also identified barriers that hindered elements of their BSP participation, findings suggest that the program was essential to their development as faculty and provides evidence of the BSP as a model for other societies seeking to advance undergraduate science education reform. The BSP is the longest-standing faculty development program sponsored by a collective group of life science societies. This collaboration promotes success across a fragmented system of more than 80 societies representing the life sciences and helps

  16. Conservatism of ASME KIR-reference curve with respect to crack arrest

    International Nuclear Information System (INIS)

    Wallin, K.; Rintamaa, R.; Nagel, G.

    2001-01-01

    The conservatism of the RT NDT temperature indexing parameter and the ASME K IR -reference curve with respect to crack arrest toughness, has been evaluated. Based on an analysis of the original ASME K Ia data, it was established that inherently, the ASME K IR -reference curve corresponds to an overall 5% lower bound curve with respect to crack arrest. It was shown that the scatter of crack arrest toughness is essentially material independent and has a standard deviation (S.D.) of 18% and the temperature dependence of K Ia has the same form as predicted by the master curve for crack initiation toughness. The 'built in' offset between the mean 100 MPa√m crack arrest temperature, TK Ia , and RT NDT is 38 deg. C (TK Ia =RT NDT +38 deg. C) and the experimental relation between TK Ia and NDT is, TK Ia =NDT+28 deg. C. The K IR -reference curve using NDT as reference temperature will be conservative with respect to the general 5% lower bound K Ia(5%) -curve, with a 75% confidence. The use of RT NDT , instead of NDT, will generally increase the degree of conservatism, both for non-irradiated as well as irradiated materials, close to a 95% confidence level. This trend is pronounced for materials with Charpy-V upper shelf energies below 100 J. It is shown that the K IR -curve effectively constitutes a deterministic lower bound curve for crack arrest The findings are valid both for nuclear pressure vessel plates, forgings and welds

  17. The ASM-NSF Biology Scholars Program: An Evidence-Based Model for Faculty Development

    Directory of Open Access Journals (Sweden)

    Amy L. Chang

    2016-05-01

    Full Text Available The American Society for Microbiology (ASM established its ASM-NSF (National Science Foundation Biology Scholars Program (BSP to promote undergraduate education reform by 1 supporting biologists to implement evidence-based teaching practices, 2 engaging life science professional societies to facilitate biologists’ leadership in scholarly teaching within the discipline, and 3 participating in a teaching community that fosters disciplinary-level science, technology, engineering, and mathematics (STEM reform. Since 2005, the program has utilized year-long residency training to provide a continuum of learning and practice centered on principles from the scholarship of teaching and learning (SoTL to more than 270 participants (“scholars” from biology and multiple other disciplines. Additionally, the program has recruited 11 life science professional societies to support faculty development in SoTL and discipline-based education research (DBER. To identify the BSP’s long-term outcomes and impacts, ASM engaged an external evaluator to conduct a study of the program’s 2010­–2014 scholars (n = 127 and society partners. The study methods included online surveys, focus groups, participant observation, and analysis of various documents. Study participants indicate that the program achieved its proposed goals relative to scholarship, professional society impact, leadership, community, and faculty professional development. Although participants also identified barriers that hindered elements of their BSP participation, findings suggest that the program was essential to their development as faculty and provides evidence of the BSP as a model for other societies seeking to advance undergraduate science education reform. The BSP is the longest-standing faculty development program sponsored by a collective group of life science societies. This collaboration promotes success across a fragmented system of more than 80 societies representing the life

  18. Batse/Sax and Batse/RXTE-ASM Joint Spectral Studies of GRBs

    Science.gov (United States)

    Paciesas, William S.

    2002-01-01

    We proposed to make joint spectral analysis of gamma-ray bursts (GRBs) in the BATSE data base that are located within the fields of view of either the BeppoSAX wide field cameras (WFCs) or the RXTE all-sky monitor (ASM). The very broad-band coverage obtained in this way would facilitate various studies of GRB spectra that are difficult to perform with BATSE data alone. Unfortunately, the termination of the CGRO mission in June 2000 was not anticipated at the time of the proposal, and the sample of common events turned out to be smaller than we would have liked.

  19. Robust boundary detection of left ventricles on ultrasound images using ASM-level set method.

    Science.gov (United States)

    Zhang, Yaonan; Gao, Yuan; Li, Hong; Teng, Yueyang; Kang, Yan

    2015-01-01

    Level set method has been widely used in medical image analysis, but it has difficulties when being used in the segmentation of left ventricular (LV) boundaries on echocardiography images because the boundaries are not very distinguish, and the signal-to-noise ratio of echocardiography images is not very high. In this paper, we introduce the Active Shape Model (ASM) into the traditional level set method to enforce shape constraints. It improves the accuracy of boundary detection and makes the evolution more efficient. The experiments conducted on the real cardiac ultrasound image sequences show a positive and promising result.

  20. 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.

  1. Computer codes KASCO and KARDIF for processing cross-sections data; Computerprogramme zur Bearbeitung der Wirkungsquerschnitts-Dateien KASCO und KARDIF

    Energy Technology Data Exchange (ETDEWEB)

    Muenzel, H; Neumann, B; Klewe-Nebenius, H; Pfennig, G

    1981-12-01

    Document of internal interest, not to be sent out without permission of authors. Summary of the computer program developed at the Karlsruhe Charged Group (Kachapag) for producing from EXFOR the handbook series `Physik Daten/Physics Data Nr. 15` of the Fachinformationszentrum Karlsruhe. (author) 6 figs. The full text is available from IAEA Nuclear Data Section

  2. 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.

  3. Comparison of Crack Growth Test Results at Elevated Temperature and Design Code Material Properties for Grade 91 Steel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyeong-Yeon; Kim, Woo-Gon; Kim, Nak-Hyun [Korea Atomic Energy Reserach Institute, Daejeon (Korea, Republic of)

    2015-01-15

    The material properties of crack growth models at an elevated temperature were derived from the results of numerous crack growth tests for Mod.9Cr-1Mo (ASME Grade 91) steel specimens under fatigue loading and creep loading at an elevated temperature. These crack growth models were needed for defect assessment under creep-fatigue loading. The mathematical crack growth rate models for fatigue crack growth (FCG) and creep crack growth (CCG) were determined based on the test results, and the models were compared with those of the French design code RCCMRx to investigate the conservatism of the code. The French design code RCC-MRx provides an FCG model and a CCG model for Grade 91 steel in Section III Tome 6. It was shown that the FCG model of RCC-MRx is conservative, while the CCG model is non-conservative compared with the present test data. Thus, it was shown that further validation of the property was required. Mechanical strength tests and creep tests were also conducted, and the test results were compared with those of RCC-MRx.

  4. Low back pain status in elite and semi-elite Australian football codes: a cross-sectional survey of football (soccer, Australian rules, rugby league, rugby union and non-athletic controls

    Directory of Open Access Journals (Sweden)

    McHardy Andrew

    2009-04-01

    Full Text Available Abstract Background Our understanding of the effects of football code participation on low back pain (LBP is limited. It is unclear whether LBP is more prevalent in athletic populations or differs between levels of competition. Thus it was the aim of this study to document and compare the prevalence, intensity, quality and frequency of LBP between elite and semi-elite male Australian football code participants and a non-athletic group. Methods A cross-sectional survey of elite and semi-elite male Australian football code participants and a non-athletic group was performed. Participants completed a self-reported questionnaire incorporating the Quadruple Visual Analogue Scale (QVAS and McGill Pain Questionnaire (short form (MPQ-SF, along with additional questions adapted from an Australian epidemiological study. Respondents were 271 elite players (mean age 23.3, range 17–39, 360 semi-elite players (mean age 23.8, range 16–46 and 148 non-athletic controls (mean age 23.9, range 18–39. Results Groups were matched for age (p = 0.42 and experienced the same age of first onset LBP (p = 0.40. A significant linear increase in LBP from the non-athletic group, to the semi-elite and elite groups for the QVAS and the MPQ-SF was evident (p Conclusion Foolers in Australia have significantly more severe and frequent LBP than a non-athletic group and this escalates with level of competition.

  5. When a Plant Resistance Inducer Leaves the Lab for the Field: Integrating ASM into Routine Apple Protection Practices.

    Science.gov (United States)

    Marolleau, Brice; Gaucher, Matthieu; Heintz, Christelle; Degrave, Alexandre; Warneys, Romain; Orain, Gilles; Lemarquand, Arnaud; Brisset, Marie-Noëlle

    2017-01-01

    Plant resistance inducers, also called elicitors, could be useful to reduce the use of pesticides. However, their performance in controlling diseases in the field remains unsatisfactory due to lack of specific knowledge of how they can integrate crop protection practices. In this work, we focused on apple crop and acibenzolar- S -methyl (ASM), a well-known SAR (systemic acquired resistance) inducer of numerous plant species. We provide a protocol for orchard-effective control of apple scab due to the ascomycete fungus Venturia inaequalis , by applying ASM in combination with a light integrated pest management program. Besides we pave the way for future optimization levers by demonstrating in controlled conditions (i) the high influence of apple genotypes, (ii) the ability of ASM to prime defenses in newly formed leaves, (iii) the positive effect of repeated elicitor applications, (iv) the additive effect of a thinning fruit agent.

  6. 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

  7. 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

  8. Dietary assessment of British police force employees: a description of diet record coding procedures and cross-sectional evaluation of dietary energy intake reporting (The Airwave Health Monitoring Study).

    Science.gov (United States)

    Gibson, Rachel; Eriksen, Rebeca; Lamb, Kathryn; McMeel, Yvonne; Vergnaud, Anne-Claire; Spear, Jeanette; Aresu, Maria; Chan, Queenie; Elliott, Paul; Frost, Gary

    2017-04-04

    Dietary intake is a key aspect of occupational health. To capture the characteristics of dietary behaviour that is affected by occupational environment that may affect disease risk, a collection of prospective multiday dietary records is required. The aims of this paper are to: (1) collect multiday dietary data in the Airwave Health Monitoring Study, (2) describe the dietary coding procedures applied and (3) investigate the plausibility of dietary reporting in this occupational cohort. A dietary coding protocol for this large-scale study was developed to minimise coding error rate. Participants (n 4412) who completed 7-day food records were included for cross-sectional analyses. Energy intake (EI) misreporting was estimated using the Goldberg method. Multivariate logistic regression models were applied to determine participant characteristics associated with EI misreporting. British police force employees enrolled (2007-2012) into the Airwave Health Monitoring Study. The mean code error rate per food diary was 3.7% (SD 3.2%). The strongest predictors of EI under-reporting were body mass index (BMI) and physical activity. Compared with participants with BMI30 kg/m 2 had increased odds of being classified as under-reporting EI (men OR 5.20 95% CI 3.92 to 6.89; women OR 2.66 95% CI 1.85 to 3.83). Men and women in the highest physical activity category compared with the lowest were also more likely to be classified as under-reporting (men OR 3.33 95% CI 2.46 to 4.50; women OR 4.34 95% CI 2.91 to 6.55). A reproducible dietary record coding procedure has been developed to minimise coding error in complex 7-day diet diaries. The prevalence of EI under-reporting is comparable with existing national UK cohorts and, in agreement with previous studies, classification of under-reporting was biased towards specific subgroups of participants. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/.

  9. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  10. Towards a consensus-based biokinetic model for green microalgae – The ASM-A

    DEFF Research Database (Denmark)

    Wágner, Dorottya Sarolta; Valverde Pérez, Borja; Sæbø, Mariann

    2016-01-01

    developed to predict microalgal growth. However, none of these models can effectively describe all the relevant processes when microalgal growth is coupled with nutrient removal and recovery from wastewaters. Here, we present a mathematical model developed to simulate green microalgal growth (ASM-A) using...... and substrate availability can introduce significant variability on parameter values for predicting the reaction rates for bulk nitrate and the intracellularly stored nitrogen state-variables, thereby requiring scenario specific model calibration. ASM-A was identified using standard cultivation medium...

  11. Production of neutron cross section library based on JENDL-4.0 to continuous-energy Monte Carlo code MVP and its application to criticality analysis of benchmark problems in the ICSBEP handbook

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Nagaya, Yasunobu

    2011-09-01

    In May 2010, JENDL-4.0 was released from Japan Atomic Energy Agency as the updated Japanese Nuclear Data Library. It was processed by the nuclear data processing system LICEM and an arbitrary-temperature neutron cross section library MVPlib - nJ40 was produced for the neutron and photon transport calculation code MVP based on the continuous-energy Monte Carlo method. The library contains neutron cross sections for 406 nuclides on the free gas model, thermal scattering cross sections, and cross sections of pseudo fission products for burn-up calculations with MVP. Criticality benchmark calculations were carried out with MVP and MVPlib - nJ40 for about 1,000 cases of critical experiments stored in the hand book of International Criticality Safety Benchmark Evaluation Project (ICSBEP), which covers a wide variety of fuel materials, fuel forms, and neutron spectra. We report all comparison results (C/E values) of effective neutron multiplication factors between calculations and experiments to give a validation data for the prediction accuracy of JENDL-4.0 for criticalities. (author)

  12. Transport theory and codes

    International Nuclear Information System (INIS)

    Clancy, B.E.

    1986-01-01

    This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on

  13. Interaction between endoplasmic/sarcoplasmic reticulum stress (ER/SR stress), mitochondrial signaling and Ca(2+) regulation in airway smooth muscle (ASM).

    Science.gov (United States)

    Delmotte, Philippe; Sieck, Gary C

    2015-02-01

    Airway inflammation is a key aspect of diseases such as asthma. Several inflammatory cytokines (e.g., TNFα and IL-13) increase cytosolic Ca(2+) ([Ca(2+)]cyt) responses to agonist stimulation and Ca(2+) sensitivity of force generation, thereby enhancing airway smooth muscle (ASM) contractility (hyper-reactive state). Inflammation also induces ASM proliferation and remodeling (synthetic state). In normal ASM, the transient elevation of [Ca(2+)]cyt induced by agonists leads to a transient increase in mitochondrial Ca(2+) ([Ca(2+)]mito) that may be important in matching ATP production with ATP consumption. In human ASM (hASM) exposed to TNFα and IL-13, the transient increase in [Ca(2+)]mito is blunted despite enhanced [Ca(2+)]cyt responses. We also found that TNFα and IL-13 induce reactive oxidant species (ROS) formation and endoplasmic/sarcoplasmic reticulum (ER/SR) stress (unfolded protein response) in hASM. ER/SR stress in hASM is associated with disruption of mitochondrial coupling with the ER/SR membrane, which relates to reduced mitofusin 2 (Mfn2) expression. Thus, in hASM it appears that TNFα and IL-13 result in ROS formation leading to ER/SR stress, reduced Mfn2 expression, disruption of mitochondrion-ER/SR coupling, decreased mitochondrial Ca(2+) buffering, mitochondrial fragmentation, and increased cell proliferation.

  14. A Retrospective Look at 20 Years of ASM Education Programs (1990-2010 and a Prospective Look at the Next 20 Years (2011-2030

    Directory of Open Access Journals (Sweden)

    Amy Chang

    2011-03-01

    Full Text Available The Education Board of the American Society for Microbiology (ASM was established in the mid-1970s to address the graduate and medical education needs of ASM members. Since then, I have watched our offerings evolve from a small, graduate-level travel grant program for ASM meetings to a growing suite of professional development and networking opportunities including fellowships, publications, and conferences. Along the way, our audience has expanded from  graduate students to undergraduate biology and K-12 teachers, students of all ages, researchers, and the public.I have been fortunate enough to watch several pivotal programs and projects support our growth and change the status quo by providing opportunities for biology educators to flourish. These include the: (i Coalition for Education in the Life Sciences, (ii ASM Division on Microbiology Education, (iii ASM Conference for Undergraduate Educators, (iv ASM Journal of Microbiology & Biology Education, and (v ASM Fellowship Fund. In this review, the background and details I offer on each initiative help explain ASM Education offerings, how our growth has been supported, and where are we headed.

  15. IM (Integrity Management) software must show flexibility to local codes

    Energy Technology Data Exchange (ETDEWEB)

    Brors, Markus [ROSEN Technology and Research Center GmbH (Germany); Diggory, Ian [Macaw Engineering Ltd., Northumberland (United Kingdom)

    2009-07-01

    There are many internationally recognized codes and standards, such as API 1160 and ASME B31.8S, which help pipeline operators to manage and maintain the integrity of their pipeline networks. However, operators in many countries still use local codes that often reflect the history of pipeline developments in their region and are based on direct experience and research on their pipelines. As pipeline companies come under increasing regulatory and financial pressures to maintain the integrity of their networks, it is important that operators using regional codes are able to benchmark their integrity management schemes against these international standards. Any comprehensive Pipeline Integrity Management System (PIMS) software package should therefore not only incorporate industry standards for pipeline integrity assessment but also be capable of implementing regional codes for comparison purposes. This paper describes the challenges and benefits of incorporating one such set of regional pipeline standards into ROSEN Asset Integrity Management Software (ROAIMS). (author)

  16. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...

  17. Simulation of municipal-industrial full scale WWTP in an arid climate by application of ASM3

    Directory of Open Access Journals (Sweden)

    Abdelsalam Elawwad

    2017-03-01

    Full Text Available In developing countries, and due to the high cost of treatment of industrial wastewater, municipal wastewater treatment facilities usually receive a mixture of municipal wastewater and partially treated industrial wastewater. As a result, an increased potential for shock loads with high pollutant concentrations is expected. The use of mathematical modelling of wastewater treatment is highly efficient in such cases. A dynamic model based on activated sludge model no. 3 (ASM3 describing the performance of the activated sludge process at a full scale wastewater treatment plant (WWTP receiving mixed domestic–industrial wastewater located in an arid area is presented. ASM3 was extended by adding the Arrhenius equation to respond to changes in temperature. BioWin software V.4 was used as the model platform. The model was calibrated under steady-state conditions, adjusting only three kinetic and stoichiometric parameters: maximum heterotrophic growth rate (μH = 8 d−1, heterotrophic aerobic decay rate (bH, O2 = 0.18 d−1, and aerobic heterotrophic yield (YH,O2 = 0.4 (gCOD/gCOD. ASM3 was successful in predicting the WWTP performance, as the model was validated with 10 months of routine daily measurements. ASM3 extended with the Arrhenius equation could be helpful in the design and operation of WWTPs with mixed municipal–industrial influent in arid areas.

  18. Using ASM Podcasts to Excite Undergraduate Students about Current Microbiological Research

    Directory of Open Access Journals (Sweden)

    Stacey E. Lettini

    2014-08-01

    Full Text Available Innovative technology is often used as a mechanism to engage students in and out of the classroom and can be used to increase critical thinking skills. Podcasts are an excellent way to introduce students to current topics and research in microbiology. The American Society for Microbiology (ASM produces three podcasts that are microbiologically focused: This Week in Microbiology (TWiM, This Week in Parasitology (TWiP, and This Week in Virology (TWiV. These podcasts are usually presented in a manner similar to a journal club, as the presenters regularly invite guests to discuss current research papers. Since students often find reading scientific literature difficult and get bogged down in the details rather than seeing the over-arching purpose of a paper, these podcasts have been used in a General Microbiology course to introduce recent research articles. The students were first assigned an original research article to read and review, and they were asked to generate questions pertaining to things they did not understand. Next, students listened to the corresponding podcast that discussed the article and used it to answer their questions. This was followed by a classroom discussion of the article and the podcast. The ASM podcast helped to demystify original research by providing details of the experimental design and presentation of the results in a language that is more casual and relatable. Students demonstrated greater critical thinking and comprehension of microbiology literature after listening to the podcast. This activity can be used in a variety of courses in the biology curriculum.

  19. User-inspired design methodology using Affordance Structure Matrix (ASM for construction projects

    Directory of Open Access Journals (Sweden)

    Maheswari J. Uma

    2017-01-01

    Full Text Available Traditionally, design phase of construction projects is often performed with incomplete and inaccurate user preferences. This is due to inefficiencies in the methodologies used for capturing the user requirements that can subsequently lead to inconsistencies and result in non-optimised end-result. Iterations and subsequent reworks due to such design inefficiencies is one of the major reasons for unsuccessful project delivery as they impact project performance measures such as time and cost among others. The existing design theories and practice are primarily based on functional requirements. Function-based design deals with design of artifact alone, which may yield favourable or unfavourable consequences with the design artifact. However, incorporating other interactions such as interactions between user & designer is necessary for optimised end-result. Hence, the objective of this research work is to devise a systematic design methodology considering all the three interactions among users, designers and artefacts for improved design efficiency. In this study, it has been attempted to apply the theory of affordances in a case project that involves the design of an offshore facility. A step-by-step methodology for developing Affordance Structure Matrix (ASM, which integrates House of Quality (HOQ and Design Structure Matrix (DSM, is proposed that can effectively capture the user requirements. HOQ is a popular quality management tool for capturing client requirements and DSM is a matrix-based tool that can capture the interdependency among the design entities. The proposed methodology utilises the strengths of both the tools, as DSM compliments HOQ in the process. In this methodology, different affordances such as AUA (Artifact-User-Affordance, AAA (Artifact-Artifact-Affordance and DDA (Designer-Designer-Affordance are captured systematically. Affordance is considered to be user-driven in this context that is in contrast to prevailing design

  20. Low back pain status in elite and semi-elite Australian football codes: a cross-sectional survey of football (soccer), Australian rules, rugby league, rugby union and non-athletic controls.

    Science.gov (United States)

    Hoskins, Wayne; Pollard, Henry; Daff, Chris; Odell, Andrew; Garbutt, Peter; McHardy, Andrew; Hardy, Kate; Dragasevic, George

    2009-04-17

    Our understanding of the effects of football code participation on low back pain (LBP) is limited. It is unclear whether LBP is more prevalent in athletic populations or differs between levels of competition. Thus it was the aim of this study to document and compare the prevalence, intensity, quality and frequency of LBP between elite and semi-elite male Australian football code participants and a non-athletic group. A cross-sectional survey of elite and semi-elite male Australian football code participants and a non-athletic group was performed. Participants completed a self-reported questionnaire incorporating the Quadruple Visual Analogue Scale (QVAS) and McGill Pain Questionnaire (short form) (MPQ-SF), along with additional questions adapted from an Australian epidemiological study. Respondents were 271 elite players (mean age 23.3, range 17-39), 360 semi-elite players (mean age 23.8, range 16-46) and 148 non-athletic controls (mean age 23.9, range 18-39). Groups were matched for age (p = 0.42) and experienced the same age of first onset LBP (p = 0.40). A significant linear increase in LBP from the non-athletic group, to the semi-elite and elite groups for the QVAS and the MPQ-SF was evident (p < 0.001). Elite subjects were more likely to experience more frequent (daily or weekly OR 1.77, 95% CI 1.29-2.42) and severe LBP (discomforting and greater OR 1.75, 95% CI 1.29-2.38). Foolers in Australia have significantly more severe and frequent LBP than a non-athletic group and this escalates with level of competition.

  1. Structural integrity evaluation of nuclear piping cracket

    International Nuclear Information System (INIS)

    Cadiz Deleito, J.C.

    1985-01-01

    The methodology to evaluation of cracks in nuclear piping is exposed. Linear elastic fracture mechanic is used to prediction of growing crack and the net section collapse theory compared with acceptation criteria of both ASME III and ASME XI code. A case allowable under ASME XI criteria is analysed under ASME III requirements. Consideration must be given to local phenomenon in crack area and local stress evaluated and compared with ASME III acceptation criteria. (author)

  2. KENO-V code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The KENO-V code is the current release of the Oak Ridge multigroup Monte Carlo criticality code development. The original KENO, with 16 group Hansen-Roach cross sections and P 1 scattering, was one ot the first multigroup Monte Carlo codes and it and its successors have always been a much-used research tool for criticality studies. KENO-V is able to accept large neutron cross section libraries (a 218 group set is distributed with the code) and has a general P/sub N/ scattering capability. A supergroup feature allows execution of large problems on small computers, but at the expense of increased calculation time and system input/output operations. This supergroup feature is activated automatically by the code in a manner which utilizes as much computer memory as is available. The primary purpose of KENO-V is to calculate the system k/sub eff/, from small bare critical assemblies to large reflected arrays of differing fissile and moderator elements. In this respect KENO-V neither has nor requires the many options and sophisticated biasing techniques of general Monte Carlo codes

  3. Corticosteroid-Induced MKP-1 Represses Pro-Inflammatory Cytokine Secretion by Enhancing Activity of Tristetraprolin (TTP) in ASM Cells.

    Science.gov (United States)

    Prabhala, Pavan; Bunge, Kristin; Ge, Qi; Ammit, Alaina J

    2016-10-01

    Exaggerated cytokine secretion drives pathogenesis of a number of chronic inflammatory diseases, including asthma. Anti-inflammatory pharmacotherapies, including corticosteroids, are front-line therapies and although they have proven clinical utility, the molecular mechanisms responsible for their actions are not fully understood. The corticosteroid-inducible gene, mitogen-activated protein kinase (MAPK) phosphatase 1 (MKP-1, DUSP1) has emerged as a key molecule responsible for the repressive effects of steroids. MKP-1 is known to deactivate p38 MAPK phosphorylation and can control the expression and activity of the mRNA destabilizing protein-tristetraprolin (TTP). But whether corticosteroid-induced MKP-1 acts via p38 MAPK-mediated modulation of TTP function in a pivotal airway cell type, airway smooth muscle (ASM), was unknown. While pretreatment of ASM cells with the corticosteroid dexamethasone (preventative protocol) is known to reduce ASM synthetic function in vitro, the impact of adding dexamethasone after stimulation (therapeutic protocol) had not been explored. Whether dexamethasone modulates TTP in a p38 MAPK-dependent manner in this cell type was also unknown. We address this herein and utilize an in vitro model of asthmatic inflammation where ASM cells were stimulated with the pro-asthmatic cytokine tumor necrosis factor (TNF) and the impact of adding dexamethasone 1 h after stimulation assessed. IL-6 mRNA expression and protein secretion was significantly repressed by dexamethasone acting in a temporally distinct manner to increase MKP-1, deactivate p38 MAPK, and modulate TTP phosphorylation status. In this way, dexamethasone-induced MKP-1 acts via p38 MAPK to switch on the mRNA destabilizing function of TTP to repress pro-inflammatory cytokine secretion from ASM cells. J. Cell. Physiol. 231: 2153-2158, 2016. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  4. 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.

  5. A novel anti-inflammatory drug, SDZ ASM 981, for the treatment of skin diseases: in vitro pharmacology.

    Science.gov (United States)

    Grassberger, M; Baumruker, T; Enz, A; Hiestand, P; Hultsch, T; Kalthoff, F; Schuler, W; Schulz, M; Werner, F J; Winiski, A; Wolff, B; Zenke, G

    1999-08-01

    SDZ ASM 981, a novel ascomycin macrolactam derivative, has high anti-inflammatory activity in animal models of allergic contact dermatitis and shows clinical efficacy in atopic dermatitis, allergic contact dermatitis and psoriasis, after topical application. Here we report on the in vitro activities of this promising new drug. SDZ ASM 981 inhibits the proliferation of human T cells after antigen-specific or non-specific stimulation. It downregulates the production of Th1 [interleukin (IL)-2, interferon-gamma] and Th2 (IL-4, IL-10) type cytokines after antigen-specific stimulation of a human T-helper cell clone isolated from the skin of an atopic dermatitis patient. SDZ ASM 981 inhibits the phorbol myristate acetate/phytohaemagglutinin-stimulated transcription of a reporter gene coupled to the human IL-2 promoter in the human T-cell line Jurkat and the IgE/antigen-mediated transcription of a reporter gene coupled to the human tumour necrosis factor (TNF)-alpha promoter in the murine mast-cell line CPII. It does not, however, affect the human TNF-alpha promoter controlled transcription of a reporter gene in a murine dendritic cell line (DC18 RGA) after stimulation via the FcgammaRIII receptor. SDZ ASM 981 also prevents the release of preformed pro-inflammatory mediators from mast cells, as shown in the murine cell line CPII after stimulation with IgE/antigen. In summary, these results demonstrate that SDZ ASM 981 is a specific inhibitor of the production of pro-inflammatory cytokines from T cells and mast cells in vitro.

  6. Validation and implementation of sandwich structure bottom plate to rib weld joint in the base section of ITER Cryostat

    Energy Technology Data Exchange (ETDEWEB)

    Prajapati, Rajnikant, E-mail: rajnikant@iter-india.org [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Bhardwaj, Anil K.; Gupta, Girish; Joshi, Vaibhav; Patel, Mitul; Bhavsar, Jagrut; More, Vipul; Jindal, Mukesh; Bhattacharya, Avik; Jogi, Gaurav; Palaliya, Amit; Jha, Saroj; Pandey, Manish [ITER-India, Institute For Plasma Research, A-29, GIDC Electronics Estate, Sector-25, Gandhinagar 382016 (India); Jadhav, Pandurang; Desai, Hemal [Larsen & Toubro Limited, Heavy Engineering, Hazira Manufacturing Complex, Gujarat (India)

    2016-11-01

    Highlights: • ITER Cryostat base section sandwich structure bottom plate to rib weld joint is qualified through mock-up. • Established welding sequence was successfully implemented on all six sectors of cryostat base section. • Each layer liquid penetrant examination has been carried out for these weld joints and found satisfactory. - Abstract: Cryostat is a large stainless steel vacuum vessel providing vacuum environment to ITER machine components. The cryostat is ∼30 m in diameter and ∼30 m in height having variable thickness from 25 mm to 180 mm. Sandwich structure of cryostat base section withstands vacuum loading and limits the deformation under service conditions. Sandwich structure consists of top and bottom plates internally strengthened with radial and circular ribs. In current work, sandwich structure bottom plate to rib weld joint has been designed with full penetration joint as per ITER Vacuum Handbook requirement considering nondestructive examinations and welding feasibility. Since this joint was outside the scope of ASME Section VIII Div. 2, it was decided to validate through mock-up of bottom plate to rib joint. Welding sequence was established to control the distortion. Tensile test, macro-structural examination and layer by layer LPE were carried out for validation of this weld joint. However possibility of ultrasonic examination method was also investigated. The test results from the welded joint mock-up were found to confirm all code and specification requirements. The same was implemented in first sector (0–60°) of base section sandwich structure.

  7. 21 CFR 106.90 - Coding.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Coding. 106.90 Section 106.90 Food and Drugs FOOD... of Infant Formulas § 106.90 Coding. The manufacturer shall code all infant formulas in conformity with the coding requirements that are applicable to thermally processed low-acid foods packaged in...

  8. Thermal-hydraulic feedback model to calculate the neutronic cross-section in PWR reactions

    International Nuclear Information System (INIS)

    Santiago, Daniela Maiolino Norberto

    2011-01-01

    In neutronic codes,it is important to have a thermal-hydraulic feedback module. This module calculates the thermal-hydraulic feedback of the fuel, that feeds the neutronic cross sections. In the neutronic co de developed at PEN / COPPE / UFRJ, the fuel temperature is obtained through an empirical model. This work presents a physical model to calculate this temperature. We used the finite volume technique of discretized the equation of temperature distribution, while calculation the moderator coefficient of heat transfer, was carried out using the ASME table, and using some of their routines to our program. The model allows one to calculate an average radial temperature per node, since the thermal-hydraulic feedback must follow the conditions imposed by the neutronic code. The results were compared with to the empirical model. Our results show that for the fuel elements near periphery, the empirical model overestimates the temperature in the fuel, as compared to our model, which may indicate that the physical model is more appropriate to calculate the thermal-hydraulic feedback temperatures. The proposed model was validated by the neutronic simulator developed in the PEN / COPPE / UFRJ for analysis of PWR reactors. (author)

  9. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  10. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  11. Aztheca Code

    International Nuclear Information System (INIS)

    Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

    2017-09-01

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  12. Survey of coded aperture imaging

    International Nuclear Information System (INIS)

    Barrett, H.H.

    1975-01-01

    The basic principle and limitations of coded aperture imaging for x-ray and gamma cameras are discussed. Current trends include (1) use of time varying apertures, (2) use of ''dilute'' apertures with transmission much less than 50%, and (3) attempts to derive transverse tomographic sections, unblurred by other planes, from coded images

  13. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

    a computational and poetic composition for two screens: on one of these, texts and voices are repeated and disrupted by mathematical chaos, together exploring the performativity of code and language; on the other, is a mix of a computer programming syntax and human language. In this sense queer code can...... be understood as both an object and subject of study that intervenes in the world’s ‘becoming' and how material bodies are produced via human and nonhuman practices. Through mixing the natural and computer language, this article presents a script in six parts from a performative lecture for two persons...

  14. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

    NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases

  15. 2nd ASME-JSME international conference on nuclear engineering -- 1993

    International Nuclear Information System (INIS)

    Peterson, P.F.

    1993-01-01

    This volume in divided into the following sections: (1) knowledge-based systems for plant operations and maintenance; (2) expert systems and neural network applications; (3) material considerations for plant life extension; (4) materials performance-operations and advanced plants; (5) operating plant O and M simplification designs and features; (6) robotics programs for the nuclear industry; (7) experience with power monitoring; (8) radwaste reduction experience in operating plants; (9) probabilistic risk assessment in design and plant operation; (10) application of individual plant examination results; (11) seismic analysis of nuclear plants/operating plant technology and experience; (12) structural evaluation of nuclear plants; (13) risk-based regulation; (14) regulatory issues for advanced plants; (15) plant standardization and licensing; (16) status of advanced plant designs; (17) reactor concepts for the 21st century; (18) enhanced reliability and improved operations of advanced plants; (19) man-machine interface and enhanced control systems; (20) plant control and instrumentation; (21) decommissioning technology; and (22) structural analysis and design code applications for nuclear facilities. Separate abstracts were prepared for 123 papers in this volume

  16. Covariance data processing code. ERRORJ

    International Nuclear Information System (INIS)

    Kosako, Kazuaki

    2001-01-01

    The covariance data processing code, ERRORJ, was developed to process the covariance data of JENDL-3.2. ERRORJ has the processing functions of covariance data for cross sections including resonance parameters, angular distribution and energy distribution. (author)

  17. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  18. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  19. Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the objectives, meeting goals and overall NASA goals for the NASA Data Standards Working Group. The presentation includes information on the technical progress surrounding the objective, short LDPC codes, and the general results on the Pu-Pw tradeoff.

  20. ANIMAL code

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1979-01-01

    This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables