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 1987 with total page 1176 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Publications of Los Alamos Research written by Los Alamos National Laboratory and published by . This book was released on 1986 with total page 184 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Proceedings of the International ANS ENS Topical Meeting on Thermal Reactor Safety San Diego California U S A February 2 6 1986 written by American Nuclear Society and published by . This book was released on 1986 with total page 414 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Technology written by and published by . This book was released on 1990 with total page 472 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 Analysis of Multiple tube Ruptures in Both Steam Generators for the Three Mile Island 1 Pressurized Water Reactor written by and published by . This book was released on 1985 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - °F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue.
Download or read book INIS Atomindex written by and published by . This book was released on 1986 with total page 1190 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Nuclear Regulatory Commission Issuances written by U.S. Nuclear Regulatory Commission and published by . This book was released on 1991 with total page 728 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Reports of the Technical Assessment Task Force Chemistry Thermal hydraulics Core damage WASH 1400 reactor safety study Alternative event sequences written by United States. President's Commission on the Accident at Three Mile Island. Technical Assessment Task Force and published by . This book was released on 1979 with total page 240 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Mitigation of Steam Generator Tube Rupture in a Pressurized Water Reactor with Passive Safety Systems written by and published by . This book was released on 1994 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.
Download or read book The Need for Change the Legacy of Tmi Report of the President s Commission on the Accident at Three Mile Island written by United States President's Commission on and published by Franklin Classics Trade Press. This book was released on 2018-11-10 with total page 210 pages. Available in PDF, EPUB and Kindle. Book excerpt: This work has been selected by scholars as being culturally important and is part of the knowledge base of civilization as we know it. This work is in the public domain in the United States of America, and possibly other nations. Within the United States, you may freely copy and distribute this work, as no entity (individual or corporate) has a copyright on the body of the work. Scholars believe, and we concur, that this work is important enough to be preserved, reproduced, and made generally available to the public. To ensure a quality reading experience, this work has been proofread and republished using a format that seamlessly blends the original graphical elements with text in an easy-to-read typeface. We appreciate your support of the preservation process, and thank you for being an important part of keeping this knowledge alive and relevant.
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 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 Three Mile Island written by U.S. Nuclear Regulatory Commission. Special Inquiry Group and published by . This book was released on 1980 with total page 214 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Evaluation of a Main Steam Line Break with Induced Multiple Tube Ruptures written by and published by . This book was released on 1995 with total page 23 pages. Available in PDF, EPUB and Kindle. Book excerpt: This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.
Download or read book Maintenance of Process Instrumentation in Nuclear Power Plants written by H.M. Hashemian and published by Springer Science & Business Media. This book was released on 2006-11-09 with total page 309 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book provides a training course for I and C maintenance engineers in power, process, chemical, and other industries. It summarizes all the scattered literature in this field. The book compiles 30 years of knowledge gained by the author and his staff in testing the I and C systems of nuclear power plants around the world. It focuses on process temperature and pressure sensors and the verification of these sensors’ calibration and response time.
