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)

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

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?

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.

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

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

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

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)

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

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)

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

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)

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

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)

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

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.

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.

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

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

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)

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

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

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

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

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

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

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.

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

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

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)

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.)

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

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)

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

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)

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

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)

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

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

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

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

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)

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.

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

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)

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)

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.

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

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

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

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

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

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)

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)

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)

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

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

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.

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

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.

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

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

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

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

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

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.

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

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.)

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.

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.

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

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

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.

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)

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)

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.

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

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

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)

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

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

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

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

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.

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)

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

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

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

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

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...

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

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

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

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

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

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.

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

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)

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)

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)

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

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)

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

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.

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

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

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

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)

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

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

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 ...

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

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

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

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.

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.

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

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)

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

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.)

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.

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

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

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

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

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)

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

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

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

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.)

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)

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

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)

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.

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

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

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.

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

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

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)

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.

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

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)

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

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.))

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)

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

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

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

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.)

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)

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)

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

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

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

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)

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)

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.

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

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

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.

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.

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.

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

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.

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

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

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.

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)

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

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)

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

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.

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

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)

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....

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)

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).

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.

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.

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.

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

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...

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

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

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.)

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.

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

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...

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

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.

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

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.

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

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.

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)

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)

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

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

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

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

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

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.

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

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

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.)

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.

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)

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

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.

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.

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

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)

Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

International Nuclear Information System (INIS)

Zado, V.

1998-01-01

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

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)

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.

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

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.

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

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

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.

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.

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

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

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.

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)

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

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

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

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

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

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.

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.

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)

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

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

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.

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

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

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.

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)

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.

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

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.

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.

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

[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.

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

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

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.

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

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)

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...

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

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.

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.

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.

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

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.

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.

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

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)

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.

Evaluation of inservice inspection examinatiions

International Nuclear Information System (INIS)

Aldrich, D.A.; Cook, J.F.

1990-05-01

In order to evaluate the effectiveness of Section 11, Division 1, ''Rules for Inservice Inspection of Nuclear Power Plant Components,'' of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code examinations, 26 inservice inspection (ISI) summary reports from 24 facilities were reviewed. It was found that these ASME Code examinations and tests are instrumental in revealing indications and defects in welds and plant components. In addition, this study uncovered that fact that some of the Section 11 requirements are apparently not clear and are misunderstood by some of the facilities. Also, the need for more stringent requirements was evaluated and some Code changes are recommended

Effect of stress relief parameters on the mechanical properties of pressure vessel steels and weldments

International Nuclear Information System (INIS)

Canonico, D.A.; Stelzman, W.J.

1976-01-01

Post weld heat treatments of thick-section A533B steel for nuclear pressure vessels are discussed with reference to the ASME code. The discussion is in the form of a lecture and summarized by noting that the ASME code, in particular Section III, Division 1, imposes a post weld heat treatment requirement on pressure vessels fabricated from low alloy high strength steels. The Code permits a holding temperature range, the high side of which could result in poorer toughness properties. Long times in excess of 100 hours and/or high temperatures, 649 0 C can result in an increase in the NDT and a decrease in the upper shelf energy

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.

Changing priorities of codes and standards -- quality engineering: Experiences in plant construction, maintenance, and operation

International Nuclear Information System (INIS)

Antony, D.D.; Suleski, P.F.; Meier, J.C.

1994-01-01

Application of the ASME Code across various fossil and nuclear plants necessitates a Company approach adapted by unique status of each plant. This arises from State Statutes, Federal Regulations and consideration of each plant's as-built history over a broad time frame of design, construction and operation. Additionally, the National Board Inspection Code accompanies Minnesota Statutes for plants owned by Northern States Power Company. This paper addresses some key points on NSP's use of ASME Code as a principal mechanical standard in plant design, construction and operation. A primary resource facilitating review of Code provisions is accurate status on current plant configuration. As plant design changes arise, the Code Edition/Addenda of original construction and installed upgrades or replacements are considered against available options allowed by current standards and dialog with the Jurisdictional Authority. Consistent with the overall goal of safe and reliable plant operation, there are numerous Code details and future needs to be addressed in concert with expected plant economics and planned outages for implementation. The discussion begins in the late 60's with new construction of Monticello and Prairie Island (both nuclear), through Sherburne County Units 1 through 3 (fossil), and their changes, replacements or repairs as operating plants

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

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

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)

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.

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.

Ranking of risk significant components for the Davis-Besse Component Cooling Water System

International Nuclear Information System (INIS)

Seniuk, P.J.

1994-01-01

Utilities that run nuclear power plants are responsible for testing pumps and valves, as specified by the American Society of Mechanical Engineers (ASME) that are required for safe shutdown, mitigating the consequences of an accident, and maintaining the plant in a safe condition. These inservice components are tested according to ASME Codes, either the earlier requirements of the ASME Boiler and Pressure Vessel Code, Section XI, or the more recent requirements of the ASME Operation and Maintenance Code, Section IST. These codes dictate test techniques and frequencies regardless of the component failure rate or significance of failure consequences. A probabilistic risk assessment or probabilistic safety assessment may be used to evaluate the component importance for inservice test (IST) risk ranking, which is a combination of failure rate and failure consequences. Resources for component testing during the normal quarterly verification test or postmaintenance test are expensive. Normal quarterly testing may cause component unavailability. Outage testing may increase outage cost with no real benefit. This paper identifies the importance ranking of risk significant components in the Davis-Besse component cooling water system. Identifying the ranking of these risk significant IST components adds technical insight for developing the appropriate test technique and test frequency

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)

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

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.)

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

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

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.

Development of reliability-based load and resistance factor design methods for piping

International Nuclear Information System (INIS)

Ayyub, Bilal M.; Hill, Ralph S. III; Balkey, Kenneth R.

2003-01-01

Current American Society of Mechanical Engineers (ASME) nuclear codes and standards rely primarily on deterministic and mechanistic approaches to design. The American Institute of Steel Construction and the American Concrete Institute, among other organizations, have incorporated probabilistic methodologies into their design codes. ASME nuclear codes and standards could benefit from developing a probabilistic, reliability-based, design methodology. This paper provides a plan to develop the technical basis for reliability-based, load and resistance factor design of ASME Section III, Class 2/3 piping for primary loading, i.e., pressure, deadweight and seismic. The plan provides a proof of concept in that LRFD can be used in the design of piping, and could achieve consistent reliability levels. Also, the results from future projects in this area could form the basis for code cases, and additional research for piping secondary loads. (author)

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....

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

PRA and Risk Informed Analysis

International Nuclear Information System (INIS)

Bernsen, Sidney A.; Simonen, Fredric A.; Balkey, Kenneth R.

2006-01-01

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. The approach allows calculations of conditional core damage probabilities both for component failures that cause initiating events and failures in standby systems that decrease the availability of these systems. The paper relates the explicit risk based methods of the new Section XI code cases to the implicit consideration of risk used in the development of Section XI. Other topics include the needed interactions of ISI engineers, plant operating staff, PRA specialists, and members of expert panels that review the risk based programs

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.

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.

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

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)

Función adrenal y metabolismo lipídico en niños asmáticos tratados con budesonida

Directory of Open Access Journals (Sweden)

Paoli-de Valeri Mariela

1999-01-01

Full Text Available Objetivo. Evaluar el efecto de bajas dosis de budesonida inhalado sobre la función adrenal y el metabolismo lipídico en niños asmáticos. Material y métodos. Se estudiaron: 10 niños asmáticos (edad promedio, 8.6 años tratados con budesonida inhalado (200-300 µg/día por un lapso mayor a tres meses (grupo A; 15 niños asmáticos (edad promedio, 7.8 años sin tratamiento esteroideo (grupo B, y 10 niños no asmáticos (grupo C. Se determinaron los niveles de cortisol basal y postestímulo con ACTH, andrógenos adrenales, lípidos y cortisol urinario. Resultados. Entre los grupos A y B no hubo diferencias significativas en las variables estudiadas. En los niños asmáticos (grupo A-B el cortisol urinario fue significativamente mayor en relación con el grupo C. Los niveles de triglicéridos, colesterol total, colesterol de la lipoproteína de baja densidad e índices aterogénicos fueron mayores en el grupo de niños asmáticos, con y sin budesonida, comparados con el grupo C. Conclusiones. El tratamiento con dosis bajas de budesonida inhalado en niños asmáticos no modificó la función del eje adrenal ni el metabolismo lipídico. Los pacientes asmáticos presentaron un perfil lipídico aterogénico que podría incrementar el riesgo de enfermedad cardiovascular.

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.

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)

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.

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

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.

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

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

Performance study of K{sub e} factors in simplified elastic plastic fatigue analyses with emphasis on thermal cyclic loading

Energy Technology Data Exchange (ETDEWEB)

Lang, Hermann, E-mail: hermann.lang@areva.com [AREVA NP GmbH, PEEA-G, Henri-Dunant-Strasse 50, 91058 Erlangen (Germany); Rudolph, Juergen; Ziegler, Rainer [AREVA NP GmbH, PEEA-G, Henri-Dunant-Strasse 50, 91058 Erlangen (Germany)

2011-08-15

As code-based fully elastic plastic code conforming fatigue analyses are still time consuming, simplified elastic plastic analysis is often applied. This procedure is known to be overly conservative for some conditions due to the applied plastification (penalty) factor K{sub e}. As a consequence, less conservative fully elastic plastic fatigue analyses based on non-linear finite element analyses (FEA) or simplified elastic plastic analysis based on more realistic K{sub e} factors have to be used for fatigue design. The demand for more realistic K{sub e} factors is covered as a requirement of practical fatigue analysis. Different code-based K{sub e} procedures are reviewed in this paper with special regard to performance under thermal cyclic loading conditions. Other approximation formulae such as those by Neuber, Seeger/Beste or Kuehnapfel are not evaluated in this context because of their applicability to mechanical loading excluding thermal cyclic loading conditions typical for power plant operation. Besides the current code-based K{sub e} corrections, the ASME Code Case N-779 (e.g. Adam's proposal) and its modification in ASME Section VIII is considered. Comparison of elastic plastic results and results from the Rules for Nuclear Facility Components and Rules for Pressure Vessels reveals a considerable overestimation of usage factor in the case of ASME III and KTA 3201.2 for the examined examples. Usage factors according to RCC-M, Adams (ASME Code Case N-779), ASME VIII (alternative) and EN 13445-3 are essentially comparable and less conservative for these examples. The K{sub v} correction as well as the applied yield criterion (Tresca or von Mises) essentially influence the quality of the more advanced plasticity corrections (e.g. ASME Code Case N-779 and RCC-M). Hence, new proposals are based on a refined K{sub v} correction.

Performance study of Ke factors in simplified elastic plastic fatigue analyses with emphasis on thermal cyclic loading

International Nuclear Information System (INIS)

Lang, Hermann; Rudolph, Juergen; Ziegler, Rainer

2011-01-01

As code-based fully elastic plastic code conforming fatigue analyses are still time consuming, simplified elastic plastic analysis is often applied. This procedure is known to be overly conservative for some conditions due to the applied plastification (penalty) factor K e . As a consequence, less conservative fully elastic plastic fatigue analyses based on non-linear finite element analyses (FEA) or simplified elastic plastic analysis based on more realistic K e factors have to be used for fatigue design. The demand for more realistic K e factors is covered as a requirement of practical fatigue analysis. Different code-based K e procedures are reviewed in this paper with special regard to performance under thermal cyclic loading conditions. Other approximation formulae such as those by Neuber, Seeger/Beste or Kuehnapfel are not evaluated in this context because of their applicability to mechanical loading excluding thermal cyclic loading conditions typical for power plant operation. Besides the current code-based K e corrections, the ASME Code Case N-779 (e.g. Adam's proposal) and its modification in ASME Section VIII is considered. Comparison of elastic plastic results and results from the Rules for Nuclear Facility Components and Rules for Pressure Vessels reveals a considerable overestimation of usage factor in the case of ASME III and KTA 3201.2 for the examined examples. Usage factors according to RCC-M, Adams (ASME Code Case N-779), ASME VIII (alternative) and EN 13445-3 are essentially comparable and less conservative for these examples. The K v correction as well as the applied yield criterion (Tresca or von Mises) essentially influence the quality of the more advanced plasticity corrections (e.g. ASME Code Case N-779 and RCC-M). Hence, new proposals are based on a refined K v correction.

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)

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.

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.

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.

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

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 ...

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).

Applying reliability centered maintenance analysis principles to inservice testing

International Nuclear Information System (INIS)

Flude, J.W.

1994-01-01

Federal regulations require nuclear power plants to use inservice test (IST) programs to ensure the operability of safety-related equipment. IST programs are based on American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements. Many of these plants also use Reliability Centered Maintenance (RCM) to optimize system maintenance. ASME Code requirements are hard to change. The process for requesting authority to use an alternate strategy is long and expensive. The difficulties of obtaining this authority make the use of RCM method on safety-related systems not cost effective. An ASME research task force on Risk Based Inservice Testing is investigating changing the Code. The change will allow plants to apply RCM methods to the problem of maintenance strategy selection for safety-related systems. The research task force is working closely with the Codes and Standards sections to develop a process related to the RCM process. Some day plants will be able to use this process to develop more efficient and safer maintenance strategies

Essential Specification Elements for Heat Exchanger Replacement

Energy Technology Data Exchange (ETDEWEB)

Bower, L.

2015-07-01

Performance upgrade and equipment degradation are the primary impetuses for a nuclear power plant to engage in the large capital cost project of heat exchanger replacement. Along with attention to these issues, consideration of heat exchanger Codes and Standards, material improvements, thermal redesign, and configuration are essential for developing User’s Design Specifications for successful replacement projects. The User’s Design Specification is the central document in procuring ASME heat exchangers. Properly stated objectives for the heat exchanger replacement are essential for obtaining the materials, configurations and thermal designs best suited for the nuclear power plant. Additionally, the code of construction required and the applied manufacturing standard (TEMA or HEI) affects how the heat exchanger may be designed or configured to meet the replacement goals. Knowledge of how Codes and Standards affect design and configuration details will aid in writing the User’s Design Specification. Joseph Oat Corporation has designed and fabricated many replacement heat exchangers for the nuclear power industry. These heat exchangers have been constructed per ASME Section III to various Code-Years or ASME Section VIII-1 to the current Code-Year also in accordance with TEMA and HEI. These heat exchangers have been a range of like-for-like replacement to complete thermal, material and configuration redesigns. Several examples of these heat exchangers with their Code, Standard and specification implications are presented. (Author.

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)

LOFT blowdown loop piping thermal analysis Class I review

International Nuclear Information System (INIS)

Kinnaman, T.L.

1978-01-01

In accordance with ASME Code, Section III requirements, all analyses of Class I components must be independently reviewed. Since the LOFT blowdown loop piping up through the blowdown valve is a Class I piping system, the thermal analyses are reviewed. The Thermal Analysis Branch comments to this review are also included. It is the opinion of the Thermal Analysis Branch that these comments satisfy all of the reviewers questions and that the analyses should stand as is, without additional considerations in meeting the ASME Code requirements and ANC Specification 60139

Adjoint sensitivity analysis of the RELAPS/MOD3.2 two-fluid thermal-hydraulic code system

International Nuclear Information System (INIS)

Ionescu-Bujor, M.

2000-10-01

This work presents the implementation of the Adjoint Sensitivity Analysis Procedure (ASAP) for the non-equilibrium, non-homogeneous two-fluid model, including boron concentration and non-condensable gases, of the RELAP5/MOD3.2 code. The end-product of this implementation is the Adjoint Sensitivity Model (ASM-REL/TF), which is derived for both the differential and discretized equations underlying the two-fluid model with non-condensable(s). The consistency requirements between these two representations are also highlighted. The validation of the ASM-REL/TF has been carried out by using sample problems involving: (i) liquid-phase only, (ii) gas-phase only, and (iii) two-phase mixture (of water and steam). Thus the 'Two-Loops with Pumps' sample problem supplied with RELAP5/MOD3.2 has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when only the liquid-phase is present. Furthermore, the 'Edwards Pipe' sample problem, also supplied with RELAP5/MOD3.2, has been used to verify the accuracy and stability of the numerical solution of the ASM-REL/TF when both (i.e., liquid and gas) phases are present. In addition, the accuracy and stability of the numerical solution of the ASM-REL/TF have been verified when only the gas-phase is present by using modified 'Two-Loops with Pumps' and the 'Edwards Pipe' sample problems in which the liquid and two-phase fluids, respectively, were replaced by pure steam. The results obtained for these sample problems depict typical sensitivities of junction velocities and volume-averaged pressures to perturbations in initial conditions, and indicate that the numerical solution of the ASM-REL/TF is as robust, stable, and accurate as the original RELAP5/MOD3.2 calculations. In addition, the solution of the ASM-REL/TF has been used to calculate sample sensitivities of volume-averaged pressures to variations in the pump head. (orig.) [de

Parametric calculations of fatigue-crack growth in piping

International Nuclear Information System (INIS)

Simonen, F.A.; Goodrich, C.W.

1983-06-01

This study presents calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors. The results indicate that the present flaw acceptance standards of ASME Section XI provide a relatively consistent set of allowable flaw sizes because the predicted life of flawed piping is relatively insensitive to pipe wall thickness, flaw aspect ratio, and piping material (ferritic versus austenitic). On the other hand, the results show that flaws that are acceptable under ASME Section XI can grow at unacceptable rates if the cyclic stresses are at the maximum level permitted by the design rules of ASME Section III. However, a review of the conservatisms inherent to the ASME code rules is presented to explain the low occurrence of piping fatigue failures in service. It is concluded that decreases in the allowable flaw sizes are not justified

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

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

Utilização de Ecolife® e Acibenzolar-s-metil (ASM no controle da antracnose da banana em pós-colheita Use of Ecolife® and Acibenzolar-S-metil (ASM on the control of antracnosis in banana post-harvest

Directory of Open Access Journals (Sweden)

Luciano Marinho Furtado

2010-09-01

Full Text Available O objetivo deste trabalho foi avaliar a ação dos produtos ASM e Ecolife no controle da antracnose pós-colheita em frutos de banana. Frutos sadios de banana, variedades maçã, prata, pacovan e cacau , em fase intermediária de maturação, foram imersos previamente em soluções de ASM e Ecolife e inoculados com Colletotrichum musae. Utilizou-se o delineamento inteiramente casualizado, em esquema fatorial com quatro repetições. Os resultados apresentados demonstraram a eficácia dos produtos nas concentrações de 5 ml/ L (Ecolife e 0,50g/ L (ASM no controle da antracnose nas variedades analisa das. A variedade cacau apresentou menor lesão quando tratada com o Ecolife (5,79 mm. Com relação ao efeito do ASM, a bana na prata demonstrou um melhor resultado, com tamanho médio de lesão de 5,62 mm. Com o decorrer do processo de maturação dos frutos houve um decréscimo na severidade da doença nas quatro variedades estudadas, exceto no tratamento testemunha, que continuou apresentando aumento no tamanho das lesões nos frutos e atingir a polpa ao final da maturação.The aim of this work was to evaluate the effects of ASM and Ecolife on the control of post-harvest antracnosis in banana. Banana fruits of Maçã, Prata, Pacovan and Figo varieties, at intermediate stage of maturation, were immersed in solutions of ASM and Ecolife and inoculated with Colletotrichum musae A completely randomized design was used, at a factorial scheme 4x2 (four varieties x two products with four replicates. The presented results demonstrated the susceptibility of the fruits to the disease, mainly Maça variety with lesion of 17,99 mm. It was demosntrated the effectiveness of products at concentrations of 5ml. L-1 (Ecolife and 0.50g. L-1 (ASM on antracnosis control. The Figo variety presented the smallest injuries when treated with Ecolife (5.79 mm. Regarding ASM effects, Prata variety demonstrated the best performance, with 5.62 mm of injuries diameter

Current Status of the Elevated Temperature Structure Design Codes for VHTR

International Nuclear Information System (INIS)

Kim, Jong-Bum; Kim, Seok-Hoon; Park, Keun-Bae; Lee, Won-Jae

2006-01-01

An elevated temperature structure design and analysis is one of the key issues in the VHTR (Very High Temperature Reactor) project to achieve an economic production of hydrogen which will be an essential energy source for the near future. Since the operating temperature of a VHTR is above 850 .deg. C, the existing code and standards are insufficient for a high temperature structure design. Thus the issues concerning a material selection and behaviors are being studied for the main structural components of a VHTR in leading countries such as US, France, UK, and Japan. In this study, the current status of the ASME code, French RCC-MR, UK R5, and Japanese code were investigated and the necessary R and D items were discussed

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

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)

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.

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.

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

Application of the risk-informed methodology for APR1400 P-T limits curve

Energy Technology Data Exchange (ETDEWEB)

Kim, K.; Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-07-01

A reactor pressure vessel (RPV) in a nuclear power plant has operational limits of pressure and temperature to prevent a potential drastic propagation of cracks due to brittle fracture. We call it a pressure-temperature limits curve (P-T limits curve). Appendix G of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, provides deterministic procedures to develop the P-T limits curve for the reactor pressure vessel. Recently, an alternative risk-informed methodology has been added in the ASME Code. Risk-informed means that we can consider insights from a probabilistic risk assessment by using this methodology. This alternative methodology provides a simple procedure to develop risk-informed P-T limits for heat up, cool down, and hydrostatic test events. The risk-informed P-T limits curve is known to provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. In this paper, we developed both the deterministic and a risk-informed P-T limits curve for an APR1400 reactor using Appendix G of the ASME Code, Section XI and compare the results in terms of additional operational margin. (author)

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...

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.

[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.

FY17 Status Report on the Initial Development of a Constitutive Model for Grade 91 Steel

Energy Technology Data Exchange (ETDEWEB)

Messner, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Phan, V. -T. [Argonne National Lab. (ANL), Argonne, IL (United States); Sham, T. -L. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-07-01

Grade 91 is a candidate structural material for high temperature advanced reactor applications. Existing ASME Section III, Subsection HB, Subpart B simplified design rules based on elastic analysis are setup as conservative screening tools with the intent to supplement these screening rules with full inelastic analysis when required. The Code provides general guidelines for suitable inelastic models but does not provide constitutive model implementations. This report describes the development of an inelastic constitutive model for Gr. 91 steel aimed at fulfilling the ASME Code requirements and being included into a new Section III Code appendix, HBB-Z. A large database of over 300 experiments on Gr. 91 was collected and converted to a standard XML form. Five families of Gr. 91 material models were identified in the literature. Of these five, two are potentially suitable for use in the ASME code. These two models were implemented and evaluated against the experimental database. Both models have deficiencies so the report develops a framework for developing and calibrating an improved model. This required creating a new modeling method for representing changes in material rate sensitivity across the full ASME allowable temperature range for Gr. 91 structural components: room temperature to 650° C. On top of this framework for rate sensitivity the report describes calibrating a model for work hardening and softening in the material using genetic algorithm optimization. Future work will focus on improving this trial model by including tension/compression asymmetry observed in experiments and necessary to capture material ratcheting under zero mean stress and by improving the optimization and analysis framework.

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

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

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

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.

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.

Problems to be solved about inelastic behaviour of materials and inelastic analysis of structures at elevated temperature

International Nuclear Information System (INIS)

Ledermann, P.; Escatha, Y. d'.

1981-01-01

At elevated temperature, ASME CODE CASE N 47 demands, in its design and analysis part to demonstrate that none of eight damages, related to the monotonic and cyclic inelastic behaviour of the material and structure, will happen during the whole life of the reactor. However this demonstration, for strain limits and creep fatigue failure, using a purely elastic analysis as in the ASME CODE Section III, is usually impossible. Inelastic analysis is then necessary. We review some of the research work (theorical and experimental) which is being done to qualify methods for an inelastic analysis of structures at elevated temperature [fr

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...

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.)

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

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)

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)

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

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.

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)

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...

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...

High cyclic fatigue of PWR primary piping generated by the pressure pulsations in coolant

International Nuclear Information System (INIS)

Zd'arek, J.; Pecinka, L.; Zeman, V.

1999-01-01

The protection of nuclear piping Class 1, 2 and 3 against fatigue failure is according to standard western practise and is based on - determining the cumulative usage factor (CUF) using equation (11) of ASME Code, Section III, Article NB 3653 for Class 1 piping; - Markl experiments and equation (10) of ASME Code, Section III, Article NC/ND 3653 for Class 2/3 piping. These evaluations cover only low cyclic loading and the possible influence of high cyclic loading as for example vibratory stresses generated by the main circulating pumps are not taken into account. This problem is fully covered in the Czech and Russian codes. The goal of this paper is 1. to clarify the basic principles; 2. to discuss in detail the methodology for the calculation of high frequency vibratory stresses; and 3. to demonstrate with a numerical example, the degree of influence of the CUF. (orig.)

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.

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)

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

Flujo espiratorio máximo en niños asmáticos: Casos y controles

Directory of Open Access Journals (Sweden)

Arturo Recabarren Lozada

1995-04-01

Full Text Available Con el fin de determinar las variaciones del Flujo Espiratorio Máximo (PEF, se estudiaron a 38 niños asmáticos (CASOS en período intercrítico de la enfermedad y a 38 niños sanos (CONTROLES, de ambos sexos comprendidos entre los 5 y 15 años de edad. Los niños asmáticos fueron clasificados por parámetros clínicos en determinado grado de severidad de asma bronquial, determinando el PEF de cada niño objeto de estudio conel mini-Wright Peak Flow Meter, en 2 registros diarios a los 06 y 18 horas, durante 7 días consecutivos, obteniendo la variabilidad del mismo. Se encuentra diferencia en la variabilidad global de niños asmáticos de todos los grados de severidad de la enfermedad comprada con la de los niños normales, con diferencia estadística altamente significativa (p<0.000001. Las variaciones diurnas del PEF ayuda en el diagnóstico del asma bronquial y también son útiles para realizar la catalogación de severidad de la enfermedad. El PEF correlaciona bien con los síntomas presentados por los pacientes y por lo tanto guarda correspondencia con la Hiperreactividad bronquial (HRB del niño asmático. Postulamos que un niño con historia clínica sugestiva, una variabilidad global mayor del 8% indica que el diagnóstico de asma es altamente probable (Rev Med Hered 1995; 6: 76-82

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

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.)

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.)

Integrity evaluation of the pressure vessels of Angra-2 and Angra-3 reactors by stress analysis

International Nuclear Information System (INIS)

Gomes, E.

1978-01-01

The integrity of the reactor pressure vessel of the unit II/III of the Nuclear Power Station at 'Angras do Reis' is evaluated by stress analysis, through the dynamics relaxation method. For the solution of the problem an axisymmetric model is fixed. Initially, the data of the Oak Ridge Vessel V-7 is compared with those obtained by two computer programs used in this study. The methods used in the computer programs are FEM and DEM. A11 the results are compared with the ASME Code Section III 1974 edition. The range deviation is determined to 99% confidence limit, in order to minimize the error probabilities. Finally, the equivalent intensity stress obtained is calculated and compared with the acceptable values of the ASME Code Section III, 1974 edition [pt

Design improvement for partial penetration welds of Pressurizer heater sleeves to head junctures

International Nuclear Information System (INIS)

Kim, Jin-Seon; Lee, Kyoung-Jin; Park, Tae-Jung; Kim, Moo-Yong

2007-01-01

ASME Code, Section III allows partial penetration welds for openings for instrumentation on which there are substantially no piping reactions and requires to have interference fit or limited diametral clearance between nozzles and vessel penetrations for the partial penetration welds. Pressurizer heater sleeves are nonaxisymmetrically attached on the hill-side of bottom head by partial penetration welds. The excessive stresses in the partial penetration weld regions of the heater sleeves are induced by pressure and thermal transient loads and also by the deformation due to manual welding process. The purpose of this study is 1) to improve design for the partial penetration welds between heater sleeves to head junctures, 2) to demonstrate the structural integrity according to the requirements of ASME Code, Section III and 3) to improve welding procedure considering the proposed design

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)

Analysis of inservice inspection relief requests

International Nuclear Information System (INIS)

Aldrich, D.A.; Cook, J.F.

1989-08-01

Nuclear Regulatory Commission (NRC) regulations require inspection (ISI) of boiling or pressurized water-cooled nuclear power plants be performed in accordance with a referenced edition and addenda of Section XI, ''Rules for Inservice Inspection of Nuclear Power Plant components,'' of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The regulations permit licensees to request relief from the NRC from specific ASME Code requirements that are determined to be impractical for the specific licensee. The NRC evaluates these requests and may grant such relief, but the NRC may also impose alternative or augmented inspections to assure structural reliability. The purpose,of this task was to evaluate the basis for ISI nondestructive examination (NDE) relief requests and to evaluate the effect of proposed ASME Code changes that would reduce the need for such requests or provide for more complete information in relief requests. This report contains the results of an analysis of an ISI relief request data base that has been expanded to include 1195 ISI relief requests versus the 296 relief requests covered in the first report in April 1987, EGG-SD-7430. Also relief requests were added to the data base which came from both first and second 10-year inspection intervals for several facilities. This provided the means to analyze the effect of recently approved ASME Code cases and updated Code requirements, some of which have been published as a result of earlier work on this task

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

Effects of ASM-024, a modulator of acetylcholine receptor function, on airway responsiveness and allergen-induced responses in patients with mild asthma.

Science.gov (United States)

Boulet, Louis-Philippe; Gauvreau, Gail M; Cockcroft, Donald W; Davis, Beth; Vachon, Luc; Cormier, Yvon; O'Byrne, Paul M

2015-01-01

To evaluate the safety, tolerability and clinical activity of ASM-024, a new cholinergic compound with dual nicotinic and muscarinic activity, in mild allergic asthma. The present study involved 24 stable, mild allergic asthmatic subjects. In a cross-over design, ASM-024 (50 mg or 200 mg) or placebo were administered once daily by nebulization over three periods of nine consecutive days separated by a three-week washout. The effect of each treatment on the forced expiratory volume in 1 s (FEV1), provocative concentration of methacholine causing a 20% decline in FEV1 (PC20), early and late asthmatic responses, and allergen-induced inflammation were measured. Seventeen subjects completed the study. During treatment with ASM-024 at 50 mg or 200 mg, the PC20 value increased respectively from a mean (± SD) 2.56±3.86 mg/mL to 4.11 mg/mL (P=0.007), and from 3.12±4.37 mg/mL to 5.23 mg/mL (P=0.005) (no change with placebo). On day 7 (day preceding allergen challenge), postdosing FEV1 increased by 2.0% with 50 mg (P=0.005) and 1.9% with 200 mg (P=0.008) (placebo -1.1%). ASM-24 had no inhibitory effect on early and late asthmatic responses, nor on sputum eosinophil or neutrophil levels. ASM-024 induced no serious adverse events, but caused cough in 22% and 48% of the subjects with 50 mg and 200 mg, respectively, compared with 10% who were on placebo. ASM-024 did not inhibit allergen-induced asthmatic response and related airway inflammation, but reduced methacholine airway responsiveness and slightly improved lung function. The mechanism by which ASM-024 improves these outcomes requires further study.

IL-17A acts via p38 MAPK to increase stability of TNF-alpha-induced IL-8 mRNA in human ASM.

Science.gov (United States)

Henness, Sheridan; van Thoor, Eveline; Ge, Qi; Armour, Carol L; Hughes, J Margaret; Ammit, Alaina J

2006-06-01

Human airway smooth muscle (ASM) plays an immunomodulatory role in asthma. Recently, IL-17A has become of increasing interest in asthma, being found at elevated levels in asthmatic airways and emerging as playing an important role in airway neutrophilia. IL-17A predominantly exerts its neutrophil orchestrating role indirectly via the induction of cytokines by resident airway structural cells. Here, we perform an in vitro study to show that although IL-17A did not induce secretion of the CXC chemokine IL-8 from ASM cells, IL-17A significantly potentiates TNF-alpha-induced IL-8 protein secretion and gene expression in a concentration- and time-dependent manner (P ASM cells, acting via a p38 MAPK-dependent posttranscriptional pathway to augment TNF-alpha-induced secretion of the potent neutrophil chemoattractant IL-8 from ASM cells.

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

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

International Nuclear Information System (INIS)

Khaleghi, H.; Fallah, E.

2003-01-01

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

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

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

International Nuclear Information System (INIS)

Elfving, K.

1998-11-01

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

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)

Bolted flanged connections subjected to longitudinal bending moments

International Nuclear Information System (INIS)

Blach, A.E.

1992-01-01

Flanges in piping systems and also pressure vessel flanges on tall columns are often subjected to longitudinal bending moments of considerable magnitude, be it from thermal expansion stresses in piping systems or from wind or seismic loadings on tall vertical pressure vessels. Except for the ASME Code, Section III, Subsections NB, NC, and ND, other pressure vessel and piping codes do not contain design ASME Nuclear Power Plant Code (Section III), an empirical formula is given, expressing a longitudinal bending moment in bolted flanged connections in terms of an equivalent internal pressure to be added to the design pressure of the flange. In this paper, an attempt is made to analyse the stresses on flanges and bolting due to external bending moments and to compare flange thicknesses thus obtained with thicknesses required using the equivalent design pressure specified in Subsections NB, NC, and ND. A design method is proposed, based on analysis and experimental work, which may be suitable for flange bending moment analysis when the rules of the Nuclear Power Plant Code are not mandatory. (orig.)

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.

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

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

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

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)

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)

Some considerations for establishing seismic design criteria for nuclear plant piping

International Nuclear Information System (INIS)

Chen, W.P.; Chokshi, N.C.

1997-01-01

The Energy Technology Engineering Center (ETEC) is providing assistance to the U.S. NRC in developing regulatory positions on the seismic analysis of piping. As part of this effort, ETEC previously performed reviews of the ASME Code, Section III piping seismic design criteria as revised by the 1994 Addenda. These revised criteria were based on evaluations by the ASME Special Task Group on Integrated Piping Criteria (STGIPC) and the Technical Core Group (TCG) of the Advanced Reactor Corporation (ARC) of the earlier joint Electric Power Research Institute (EPRI)/NRC Piping ampersand Fitting Dynamic Reliability (PFDR) program. Previous ETEC evaluations reported at the 23rd WRSM of seismic margins associated with the revised criteria are reviewed. These evaluations had concluded, in part, that although margins for the timed PFDR tests appeared acceptable (>2), margins in detuned tests could be unacceptable (<1). This conclusion was based primarily on margin reduction factors (MRFs) developed by the ASME STGIPC and ARC/TCG from realistic analyses of PFDR test 36. This paper reports more recent results including: (1) an approach developed for establishing appropriate seismic margins based on PRA considerations, (2) independent assessments of frequency effects on margins, (3) the development of margins based on failure mode considerations, and (4) the implications of Code Section III rules for Section XI

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}.

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.

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.

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

Application of the French codes to the pressurized thermal shocks assessment

Energy Technology Data Exchange (ETDEWEB)

Chen, Mingya; Wang, Rong Shan; Yu, Weiwei; Lu, Feng; Zhang, Guo Dong; Xue, Fei; Chen, Zhilin [Suzhou Nuclear Power Research Institute, Life Management Center, Suzhou (China); Qian, Guian [Paul Scherrer Institute, Nuclear Energy and Safety Department, Villigen (Switzerland); Shi, Jinhua [Amec Foster Wheeler, Clean Energy Department, Gloucester (United Kingdom)

2016-12-15

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the 'screening criterion' for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no 'screening criterion'. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

Application of the French Codes to the Pressurized Thermal Shocks Assessment

Directory of Open Access Journals (Sweden)

Mingya Chen

2016-12-01

Full Text Available The integrity of a reactor pressure vessel (RPV related to pressurized thermal shocks (PTSs has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the “screening criterion” for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no “screening criterion”. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

Application of the French codes to the pressurized thermal shocks assessment

International Nuclear Information System (INIS)

Chen, Mingya; Wang, Rong Shan; Yu, Weiwei; Lu, Feng; Zhang, Guo Dong; Xue, Fei; Chen, Zhilin; Qian, Guian; Shi, Jinhua

2016-01-01

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the 'screening criterion' for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no 'screening criterion'. In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed

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)

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...

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

Energy Technology Data Exchange (ETDEWEB)

Elfving, K

1998-11-01

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

Mechanical Design and Analysis of LCLS II 2 K Cold Box

Science.gov (United States)

Yang, S.; Dixon, K.; Laverdure, N.; Rath, D.; Bevins, M.; Bai, H.; Kaminski, S.; Ravindranath, V.

2017-12-01

The mechanical design and analysis of the LCLS II 2 K cold box are presented. Its feature and functionality are discussed. ASME B31.3 was used to design its internal piping, and compliance of the piping code was ensured through flexibility analysis. The 2 K cold box was analyzed using ANSYS 17.2; the requirements of the applicable codes—ASME Section VIII Division 2 and ASCE 7-10—were satisfied. Seismic load was explicitly considered in both analyses.

The Effect of Cold Work on Properties of Alloy 617

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-01

Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was submitted in 1992 to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed.1 Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. The mechanical and physical properties of Alloy 617 were extensively characterized for the VHTR programs in the 1980’s and incorporated into the 1992 draft Code Case. Recently, the properties of modern heats of the alloy that incorporate an additional processing step, electro-slag re-melting, have been characterized both to confirm that the properties of contemporary material are consistent with those in the historical record and to increase the available database. A number of potential issues that were identified as requiring further consideration prior to the withdrawal of the 1992 Code Case are also being re-examined in the current R&D program. Code Cases are again being developed to allow use of Alloy 617 for nuclear design within the rules of the ASME Boiler and Pressure Vessel Code. In general the Code defines two temperature ranges for nuclear design with austenitic and nickel based alloys. Below 427°C (800°F) time dependent behavior is not considered, while above this temperature creep and creep-fatigue are considered to be the dominant life-limiting deformation modes. There is a corresponding differentiation in the treatment of the potential for effects associated with cold work. Below 427°C the principal issue is the relationship between the level of cold work and the propensity for stress corrosion cracking and above that temperature the primary concern is the impact of cold work on creep-rupture behavior.

49 CFR 178.338-16 - Inspection and testing.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Inspection and testing. 178.338-16 Section 178.338... PACKAGINGS Specifications for Containers for Motor Vehicle Transportation § 178.338-16 Inspection and testing... Section VIII of the ASME Code. The welder and the welding procedure must be qualified in accordance with...

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

Jaimot Jimenez, J. J.

2013-07-01

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

Dikkers Valves for nuclear industry

International Nuclear Information System (INIS)

Anon.

1975-01-01

Most countries have adopted the ASME Boiler and Pressure Vessel Code Section III, as the basis of their national requirements for licensing nuclear components. This Code gives clear directives for defining design requirements coupled with a controlled manufacturing system. It has always been and still is the policy of Dikkers to manufacture high-quality products. Dikkers manufacture nuclear products in accordance with this Code, Section III; indeed many features exceed these minimum requirements. At the Nuclex Exhibition in Basel, Dikkers Valves BV will exhibit its main products for use in nuclear power plants. (Auth.)

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

Flexural behavior of concrete beam with mechanical splices of reinforcement subjected to cyclic loading

International Nuclear Information System (INIS)

Nab, H. S.; Kim, W. B.

2008-01-01

In nuclear power plant structures, the mechanical rebar splices are designated and constructed on the basis of ACI and ASME code. Regardless of good performance on mechanical rebar splices, these splicing methods that did not be registered on ASME code have not restricted to apply to construction site. In this study, the main candidate splice is cold roll formed parallel threaded splice. This was registered newly in ASME Section III division 2 CC 4333 'Mechanical Splices' in 2004. To compare the traditional rebar splice with mechanical rebar splices, concrete beams were made to evaluate the ductility of spliced reinforcing bars. Based on Experimental results, it was identified that the mechanical rebar splices by parallel threaded coupler had better accumulated dissipation energy capacity to resist seismic behavior than the traditional lapping splices. It showed that concrete specimens with D36 reinforcing bar coupler are 1.8 times better performance and that concrete specimens with D22 reinforcing bar coupler are 2.8 times better performance. (authors)

Evaluation of local tensions through finite elements applied to a large diameter pipe subjected to vacuum condition of a petroleum refinery; Avaliacao das tensoes locais atraves de elementos finitos aplicada a uma tubulacao de grande diametro sujeita a condicao de vacuo de uma dada refinaria de petroleo

Energy Technology Data Exchange (ETDEWEB)

Neves, Julio C. Goes; Balbi, Diego J. G. [Promom Engenharia, Rio de Janeiro, RJ (Brazil)

2012-07-01

The objective of this paper is to present an evaluation of the results obtained in the study of local stress in wall of large diameter pipe. The case study consists of to analyze a pipeline system with 66 inch, which is responsible for transporting oil, oven to the Tower of vacuum distillation unit in a petroleum refining. The absence of internal pressure leads to a critical with respect to the collapse of the walls of the tube in long sections, without the presence of additional elements increase the rigidity of the geometry. The ASME Section VIII Division 1 advocates the use of additional plates, called stiffeners, which aim to curb the efforts from this condition. Thus, it is necessary structural assessment of critical portions of the system in implementing this solution. Therefore, complementary approaches have been proposed, passing by ASME B31.3, Section VIII Division 1, moreover, a computer simulation of stresses through the finite element method, which the results were analyzed according to criteria of tensions presents in ASME Code Section VIII Division 2. (author)

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

Development of an inelastic stress analysis code 'KINE-T' and its evaluations

International Nuclear Information System (INIS)

Kobatake, K.; Takahashi, S.; Suzuki, M.

1977-01-01

Referring to the ASME B and PVC Code Case 1592-7, the inelastic stress analysis is required for the designs of the class 1 components in elevated temperature if the results of the elastic stress analysis and/or simplified inelastic analysis do not satisfy the requirements. Authors programmed a two-dimensional axisymmetric inelastic analysis code 'KINE-T', and carried out its evaluations and an application. This FEM code is based on the incremental method and the following: elastic-plastic constitutive equation (yield condition of von Mises; flow rule of Prandtl-Reuss; Prager's hardening rule); creep constitutive equation (equation of state approach; flow rule of von Mises; strain hardening rule); the temperature dependency of the yield function is considered; solution procedure of the assembled stiffness matrix is the 'initial stress method'. After the completion of the programming, authors compared the output with not only theoretical results but also with those of the MARC code and the ANSYS code. In order to apply the code to the practical designing, authors settled a quasi-component two-dimensional axisymmetric model and a loading cycle (500 cycles). Then, an inelastic analysis and its integrity evaluation are carried out

Application of improved quality control technology to pressure vessels

International Nuclear Information System (INIS)

Kriedt, F.

1985-01-01

Within the last decade, ASME Boiler and Pressure Vessel Code Section VIII-1 instituted requirements for a formal written quality control system. The results, good and bad, of this requirement are discussed. The effects are far reaching from a national economic standpoint. Quality control technology has improved. These improvements are discussed and compared to existing requirements of the CODE. Recommended improvements are suggested

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.

Pump testing - Comparison of factory vs. field test of centrifugal pumps

International Nuclear Information System (INIS)

Fehlau, R.

1992-01-01

Testing of pumps in situ, i.e., as installed in a system, will typically yield somewhat different performance results from the original manufacturer's factory test. This paper discusses some of the reasons for these variations. It shows that the factory test curves can be used for evaluation of initial acceptance tests but not for reference in normal inservice testing (IST). This is the basis for reference values used in American Society of Mechanical Engineers (ASME) Section 11 specifications and the revised ASME Code 1990

Frame Synchronization Without Attached Sync Markers

Science.gov (United States)

Hamkins, Jon

2011-01-01

We describe a method to synchronize codeword frames without making use of attached synchronization markers (ASMs). Instead, the synchronizer identifies the code structure present in the received symbols, by operating the decoder for a handful of iterations at each possible symbol offset and forming an appropriate metric. This method is computationally more complex and doesn't perform as well as frame synchronizers that utilize an ASM; nevertheless, the new synchronizer acquires frame synchronization in about two seconds when using a 600 kbps software decoder, and would take about 15 milliseconds on prototype hardware. It also eliminates the need for the ASMs, which is an attractive feature for short uplink codes whose coding gain would be diminished by the overheard of ASM bits. The lack of ASMs also would simplify clock distribution for the AR4JA low-density parity-check (LDPC) codes and adds a small amount to the coding gain as well (up to 0.2 dB).

The status of Korean nuclear codes and standards

International Nuclear Information System (INIS)

Namha Kim; Jong-Hae Kim

2005-01-01

Korea Electric Power Industry Code (KEPIC), a set of integrated standards applicable to the design, construction and operation of electric power facilities including nuclear power plants, has been developed on the basis of referring to the prevailing U.S. codes and standards which had been applied to the electric power facilities in Korea. Being the developing and managing organization of KEPIC, Korea Electric Association (KEA) published its first edition in 1995, the second in 200,0 and is expected to publish the 2005 edition. KEPIC was applied to the construction of Ulchin Nuclear Units 5 and 6 in 1997, and will be applicable to the construction of forthcoming nuclear power plants in Korea. Along with the effectuation of the Agreement on Technical Barriers to Trade (TBT) in 1995, the international trend related to codes and standards is changing rapidly. The KEA is, therefore, making its utmost efforts so as for KEPIC to keep abreast with the changing environment in international arena. KEA notified ISO/IEC Information Centre of its acceptance of the Code of Good Practice in the Agreement on TBT. The 2005 KEPIC edition will be retrofitted according to the ISO/IEC Guide 21- Adoption of International Standards as regional or national standards. KEA's efforts will help KEPIC correspond with international standards such as ISO/IEC standards, and internationally recognized standards such as ASME codes and standards. (authors)

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.)

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

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