Download or read book Analytic Studies Pertaining to Steam Generator Tube Rupture Accidents written by B. A. Kashiwa and published by . This book was released on 1985 with total page 80 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Risk Assessment of Severe Accident induced Steam Generator Tube Rupture written by and published by . This book was released on 1998 with total page 218 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC's Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs.
Download or read book Analysis of Steam Generator Tube Rupture as a Severe Accident Using MELCOR 1 8 4 written by Hongrun Yang and published by . This book was released on 1999 with total page 110 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Consequence Analysis for a Steam Generator Tube Rupture Accident written by Chʻun-fa Chuang and published by . This book was released on 1983 with total page 217 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Evaluation of Steam Generator Tube Rupture Events written by L. B. Marsh and published by . This book was released on 1980 with total page 132 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book A Simulation of Steam Generator Tube Rupture Accident by Safety Analysis Code RELAP5 MOD1 written by Akio Gofuku and published by . This book was released on 1989 with total page 57 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Effectiveness of Surveillance Sampling Strategies for Detecting Steam Generator Tube Degradation written by Andrew Jordan Clark and published by . This book was released on 2015 with total page 110 pages. Available in PDF, EPUB and Kindle. Book excerpt: Nuclear power plants seeking to extend their operating license must first address the degradation of systems, structures, and components (SSCs) to ensure they can maintain a satisfactory level of reliability into the extended lifetime. Passive SSCs play an important role in determining the feasibility of life extension. Part of the feasibility analysis requires plants to demonstrate the viability and reliability of passive SSCs into the extended lifetime. The research carried out toward this thesis considers primary water stress corrosion cracking (PWSCC) of steam generator (SG) tubes as an example degradation mechanism. An empirical model for PWSCC crack growth is adopted to simulate crack growth over a 40-year operating lifetime. Surveillance and maintenance strategies similar to those performed by the industry are integrated with the PWSCC crack growth model to determine the effectiveness of surveillance strategies for detecting SG tube degradation. The results of this analysis were applied to a specific accident scenario in which steam generator tubes rupture following a depressurization of the secondary side due to the sudden rupture of a steam-line caused by flow-accelerated corrosion. Likelihood of a spontaneous steam generator tube rupture is also assessed. The analysis and application of the specific accident scenario indicates a maximum core damage frequency in the 16th year. Sensitivity analyses into the probability of detection (POD) and crack growth rates were also performed. As expected, the likelihood of the accident scenario occurring increased significantly as the maximum POD was decreased. When crack growth rates were slowed down, the overall likelihood of the accident scenario decreased and the expected occurrence of the accident scenario was delayed.
Download or read book Analysis of Steam generator Tube rupture Events Combined with Auxiliary feedwater Control system Failure for Three Mile Island Unit 1 and Zion Unit 1 Pressurized Water Reactors written by and published by . This book was released on 1986 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx. 63 K (approx. 113°F) for TMI-1 and approx. 44 K (approx. 80°F) for Zion-1.
Download or read book Energy Research Abstracts written by and published by . This book was released on 1995 with total page 782 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book RELAP5 MOD3 3 Analysis of Steam Generator Tube Rupture SGTR Accident for NPP Kr ko written by V. Benčik and published by . This book was released on 2014 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book RELAP5 MOD3 3 Analysis of Steam Generator Tube Rupture SGTR Accident for NPP Kr ko written by V. Benčik and published by . This book was released on 2014 with total page 55 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book The Key Issues in Transient Analysis Calculations which Affect the Radiological Consequences Following Design basis Steam Generator Tube Rupture Accidents written by C. J. Ablott and published by . This book was released on 1996 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Steam Generator Tube Rupture Effects on a LOCA written by and published by . This book was released on 1979 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process.
Download or read book Steam Generator Tube Rupture Analysis for Maanshan Nuclear Power Plant written by Song-Feng Wang and published by . This book was released on 1989 with total page 74 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Radioactivity Transport Following Steam Generator Tube Rupture written by Joram Hopenfeld and published by . This book was released on 1985 with total page 37 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book ERDA Energy Research Abstracts written by and published by . This book was released on 1985 with total page 676 pages. Available in PDF, EPUB and Kindle. Book excerpt:
