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Author : J. Dufresne
Publisher :
ISBN 13 :
Total Pages : 152 pages
Book Rating : 4.3/5 (91 download)
Download or read book Probabilistic Study on the Rupture of Light Water Reactor Pressure Vessel written by J. Dufresne and published by . This book was released on 1977 with total page 152 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : W. E. Vesely
Publisher :
ISBN 13 :
Total Pages : 64 pages
Book Rating : 4.3/5 (91 download)
Download or read book The Octavia Computer Code written by W. E. Vesely and published by . This book was released on 1978 with total page 64 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :
Publisher :
ISBN 13 :
Total Pages : 108 pages
Book Rating : 4.:/5 (319 download)
Download or read book A Survey of Water Reactor Primary System Conditions Pertinent to the Study of Pipe Rupture written by and published by . This book was released on 1964 with total page 108 pages. Available in PDF, EPUB and Kindle. Book excerpt: Designers, builders and operators of water cooled reactor systems have been queried regarding the range of parameters which should be investigated in a study of piping rupture.
Author :
Publisher :
ISBN 13 :
Total Pages : 26 pages
Book Rating : 4.:/5 (946 download)
Download or read book Initial Probabilistic Evaluation of Reactor Pressure Vessel Fracture with Grizzly and Raven written by and published by . This book was released on 2015 with total page 26 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. The first application of Grizzly has been to study fracture in embrittled reactor pressure vessels (RPVs). Grizzly can be used to model the thermal/mechanical response of an RPV under transient conditions that would be observed in a pressurized thermal shock (PTS) scenario. The global response of the vessel provides boundary conditions for local models of the material in the vicinity of a flaw. Fracture domain integrals are computed to obtain stress intensity factors, which can in turn be used to assess whether a fracture would initiate at a pre-existing flaw. These capabilities have been demonstrated previously. A typical RPV is likely to contain a large population of pre-existing flaws introduced during the manufacturing process. This flaw population is characterized stastistically through probability density functions of the flaw distributions. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation during a transient event. This report documents initial work to perform probabilistic analysis of RPV fracture during a PTS event using a combination of the RAVEN risk analysis code and Grizzly. This work is limited in scope, considering only a single flaw with deterministic geomerty, but with uncertainty introduced in the parameters that influence fracture toughness. These results are benchmarked against equivalent models run in the FAVOR code. When fully developed, the RAVEN/Grizzly methodolgy for modeling probabilistic fracture in RPVs will provide a general capability that can be used to consider a wider variety of vessel and flaw conditions that are difficult to consider with current tools. In addition, this will provide access to advanced probabilistic techniques provided by RAVEN, including adaptive sampling and parallelism, which can dramatically decrease run times.
Author : Mark P. Rubin
Publisher :
ISBN 13 :
Total Pages : 44 pages
Book Rating : 4.3/5 (91 download)
Download or read book The Probability of Intersystem LOCA written by Mark P. Rubin and published by . This book was released on 1980 with total page 44 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book THE INCLUSION OF INNER SURFACE BREAKING FLAWS IN PROBABILISTIC FRACTURE MECHANICS ANALYSES OF REACTOR VESSELS SUBJECTED TO PLANNED NORMAL COOL-DOWN TRANSIENTS1 written by and published by . This book was released on 2008 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that lightwater nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses/publications, consistent with the assumptions utilized for this particular reactor in the PTS reevaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify/generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.
Author : D. O. Harris
Publisher :
ISBN 13 :
Total Pages : pages
Book Rating : 4.:/5 (5 download)
Download or read book Probabilistic Analysis of Rupture in Nuclear Reactor Coolant Piping Paper II-7 of Topical Meeting in Probabilistic Analysis of Nuclear Reactor Safety written by D. O. Harris and published by . This book was released on 1978 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : K. V. Cook
Publisher :
ISBN 13 :
Total Pages : 16 pages
Book Rating : 4.:/5 (244 download)
Download or read book Detection and Characterization of Flaws in Segments of Light Water Reactor Pressure Vessels written by K. V. Cook and published by . This book was released on 1987 with total page 16 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : M. Brumovský
Publisher :
ISBN 13 :
Total Pages : 64 pages
Book Rating : 4.:/5 (256 download)
Download or read book Safety of light-water reactor pressure vessels against brittle fracture written by M. Brumovský and published by . This book was released on 1979 with total page 64 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : J. Dufresne
Publisher :
ISBN 13 :
Total Pages : 0 pages
Book Rating : 4.:/5 (141 download)
Download or read book PWR reactor pressure vessel failure probabilities 4th International conference on pressure vessel technology written by J. Dufresne and published by . This book was released on with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author :
Publisher :
ISBN 13 :
Total Pages : 0 pages
Book Rating : 4.:/5 (682 download)
Download or read book Light Water Reactor Lower Head Failure Analysis written by and published by . This book was released on 1993 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.
Download or read book A SCOPING STUDY written by and published by . This book was released on 2011 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of a small set of comprehensive event trees and fault trees and recommendation for future work.
Author : J. R. Hawthorne
Publisher :
ISBN 13 :
Total Pages : 152 pages
Book Rating : 4.3/5 (91 download)
Download or read book Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program: Postirradiation notch ductility and tensile strength determinations for PSF simulated surveillance and through-wall specimen capsules written by J. R. Hawthorne and published by . This book was released on 1984 with total page 152 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : Paul E. Roege
Publisher :
ISBN 13 :
Total Pages : 292 pages
Book Rating : 4.:/5 (227 download)
Download or read book Potential Effects of Leak-Before-Break on Light Water Reactor Design written by Paul E. Roege and published by . This book was released on 1985 with total page 292 pages. Available in PDF, EPUB and Kindle. Book excerpt: Detailed analyses and substantial structures are required in Light Water Reactor power plants, to protect against the damaging effects of pipe breaks. As an alternative, one might show that the growth of any crack in a particular pipe would lead to fluid leaks which would be detected long before such a crack would result in a pipe break. This principle is referred to as Leak-Before-Break. This thesis presents the results of a study which was conducted to assess the potential impact of Leak-Before-Break on the design of modern Light Water Reactors. It was determined that a majority of pipe rupture restraints could be eliminated in a typical plant, with a potential cost savings of tens of millions of dollars per plant. Assumptions about reactor operating conditions and leak detection capability are critical. Some recommendations made in this thesis are: safety margins to be used in leak-before-break determination should be based on actual risk; and requirements for leak detection sensitivity in nuclear plants should be based on specific needs.
Author : Hongdi Wang
Publisher :
ISBN 13 : 9789180407038
Total Pages : 0 pages
Book Rating : 4.4/5 (7 download)
Download or read book Thermo-mechanical Assessment of Reactor Pressure Vessels of Light Water Reactors During Severe Accidents written by Hongdi Wang and published by . This book was released on 2023 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Reactor Pressure Vessel Failure Probability Following Through-wall Cracks Due to Pressurized Thermal Shock Events written by and published by . This book was released on 1986 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Author : AS. Amar
Publisher :
ISBN 13 :
Total Pages : 13 pages
Book Rating : 4.:/5 (125 download)
Download or read book Residual Life Assessment of Light Water Reactor Pressure Vessels written by AS. Amar and published by . This book was released on 1989 with total page 13 pages. Available in PDF, EPUB and Kindle. Book excerpt: The service-dependent degradation (aging) of light water reactor (LWR) pressure vessels due to irradiation embrittlement is discussed in this paper. The major variables which influence the irradiation embrittlement of LWR vessels are the copper and nickel content of the vessel materials and the fluence. The vessel beltline region is subjected to the largest fluences. A surveillance program, which consists of tension and Charpy-V-notch (CVN) testing of irradiated specimens of base, heat-affected-zone, and weld materials, is required to monitor changes in their embrittlement. Three main unresolved technical issues are: (1) the limited range and accuracy of the current correlations for calculating shifts in the reference temperature for nil-ductility transition (RTNDT) and changes in the Charpy upper shelf energy; (2) the need to demonstrate the conservatism of using CVN-based RTNDTshifts for certain sensitive reactor pressure vessel materials; and (3) the type of surveillance program required for any renewed operating license period. The damage caused by irradiation embrittlement can impact plant operating procedures, including heat up/cool down and hydrostatic test procedures, as well as the acceptability of various plant transients.
