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 Steam Generator Tube Rupture Analyses for Maanshan Nuclear Power Plant written by Jan-Ru Tang and published by . This book was released on 1987 with total page 70 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Steam Generator Tube Rupture Analysis Deficiency written by and published by . This book was released on 1988 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book International Standard Problem No 20 written by OECD Nuclear Energy Agency. Committee on the Safety of Nuclear Installations and published by . This book was released on 1988 with total page 76 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 NRC Report on the January 25 1982 Steam Generator Tube Rupture at R E Ginna Nuclear Power Plant written by and published by . This book was released on 1982 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
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 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 Analysis of Steam Generator Tube Rupture Events at Oconee and Ginna written by Institute of Nuclear Power Operations (U.S.) and published by . This book was released on 1982 with total page 403 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Data Analysis written by Yuhao Huang and published by . This book was released on 1990 with total page 40 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report provides an analysis of the data collected during a one-month visit to a 2-unit, non-U.S. PWR. The data consist of results from interviews (largely with plant operators and former shift engineers) and from reviews of videotapes covering crew responses to steam generator tube rupture training exercises. The interviews were aimed at indicating perceptions of individual and group skills. The analysis shows that the interview results are fairly consistent, that the time required to perform key actions does not generally correlate very well with the team quality ratings obtained from the interviews, and that the team quality ratings obtained from interviews correlate reasonably with ratings of the team performances (during the exercises) developed using the 7-dimension scale described in PNL-7250.
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 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 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 Steam Generator Secondary PH During a Steam Generator Tube Rupture written by and published by . This book was released on 1991 with total page 132 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL to perform an analytical assessment of secondary coolant system (SCS) pH during an SGTR. Design basis thermal and hydraulic calculations were used together with industry standard chemistry guidelines to determine the SCS chemical concentrations during an SGTR. These were used as input to the Facility for Analysis of Chemical Thermodynamics computer system to calculate the equilibrium pH in the SCS at various discrete time during an SGTR. The results of this analysis indicate that the SCS pH decreases from the initial value of 8.8 to approximately 6.5 by the end of the transient, independent of PWR design.
Download or read book Main Steam Line Break Accident Analysis for Maanshan Nuclear Power Plant Using RELAP5 MOD1 Code written by Kuo-Shing Liang and published by . This book was released on 1986 with total page 95 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Analysis of Recovery Procedures for a Single Steam Generator Tube Rupture written by Charles L. Nalezny and published by . This book was released on 1984 with total page 11 pages. Available in PDF, EPUB and Kindle. Book excerpt:
