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 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 Evaluation of Steam Generator Tube Rupture Events written by L. B. Marsh and published by . This book was released on 1980 with total page 136 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 Assessment of RELAP5 MOD3 1 with the LSTF SB SG 06 Experiment Simulating a Steam Generator Tube Rupture Transient written by and published by . This book was released on 1996 with total page 108 pages. Available in PDF, EPUB and Kindle. Book excerpt: The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

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 Transient Analysis of the NuScale Power Helical coil Steam Generator Tube Rupture Using RELAP5 3D written by Kyle P. Johnson and published by . This book was released on 2021 with total page 0 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 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 Analysis of RELAP5 MOD3 3 Prediction of 2 inch Loss of coolant Accident at Kr ko Nuclear Power Plant written by Iztok Parzer and published by . This book was released on 2010 with total page 41 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Steam Line Break Analysis Using RELAP5 MOD3 3 for Steam Generator Blowdown Load Assessment written by Han-Sik Jung and published by . This book was released on 2017 with total page 29 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Feedwater Line Break Analysis Using RELAP5 MOD3 3 for Steam Generator Blowdown Load Assessment written by Han-Sik Jung and published by . This book was released on 2016 with total page 31 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Steam Line Break Analysis Using RELAP5 MOD3 3 for Steam Generator Blowdown Load Assessment written by Han-Sik Jung and published by . This book was released on 2017 with total page 29 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 Investigation of a Steam Generator Tube Rupture Sequence Using VICTORIA written by and published by . This book was released on 1995 with total page 20 pages. Available in PDF, EPUB and Kindle. Book excerpt: VICTORIA-92 is a mechanistic computer code for analyzing fission product behavior within the reactor coolant system (RCS) during a severe reactor accident. It provides detailed predictions of the release of radionuclides and nonradioactive materials from the core and transport of these materials within the RCS. The modeling accounts for the chemical and aerosol processes that affect radionuclide behavior. Coupling of detailed chemistry and aerosol packages is a unique feature of VICTORIA; it allows exploration of phenomena involving deposition, revaporization, and re-entrainment that cannot be resolved with other codes. The purpose of this work is to determine the attenuation of fission products in the RCS and on the secondary side of the steam generator in an accident initiated by a steam generator tube rupture (SGTR). As a class, bypass sequences have been identified in NUREG-1150 as being risk dominant for the Surry and Sequoyah pressurized water reactor (PWR) plants.

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: