Download or read book Sensitivity Studies of Failure of Steam Generator Tubes During Main Steam Line Break and Other Secondary Side Depressurization Events written by and published by . This book was released on 2007 with total page 65 pages. Available in PDF, EPUB and Kindle. Book excerpt:
Download or read book Resolution of Generic Safety Issue 188 written by U.s. Nuclear Regulation Comion and published by CreateSpace. This book was released on 2014-07-25 with total page 26 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report addresses Generic Safety Issue (GSI) 188, “Steam Generator Tube Leaks or Ruptures Concurrent with Containment Bypass from Main Steam Line or Feedwater Line Breaches,” which concerns the potential for additional tube leakage or ruptures from the growth of existing cracks in steam generator tubes resulting from the dynamic loads following a main steam line break (MSLB) or feedwater line break (FWLB). To address the issue, this report provides the technical findings from thermal-hydraulic transient analyses and sensitivity studies, a simplified finite-element model of steam generator support structures and tubes, and structural analyses and sensitivity studies of the potential for crack growth. The results show that the additional dynamic loads from an MSLB are greater than those from an FWLB. The report concludes that dynamic loads from an MSLB are low and do not affect the structural integrity of tubes and do not lead to additional leakage or ruptures beyond what would be determined using differential pressure loads alone. Therefore, GSI-188 is closed, and the staff recommends no changes to existing regulations or guidance with respect to the dynamic loads induced by a breach of the main steam or feedwater line.
Download or read book FLASH Predictions of the MB 2 Steam Line Break Tests written by and published by . This book was released on 1992 with total page 6 pages. Available in PDF, EPUB and Kindle. Book excerpt: If a main steam line from a pressurized water reactor (PWR) steam generator were to rupture, the effect would be a depressurization of the secondary side and a consequential overcooling transient on the primary side. Analyses must accurately predict the effects of the rapid cooldown of the reactor vessel coolant on positive nuclear-kinetic reactivity feedback to the core plus thermal shock to the reactor vessel and other primary system components. Many early studies of the steam line break (SLB) transient made extremely conservative assumptions to maximize the primary to secondary heat transfer which in turn maximized the reactor vessel cooldown rate. Among the more significant of these assumptions was that flow from the break was pure steam and that the tube bundle remained covered until the secondary mass inventory was significantly reduced. The Model F commercial PWR steam generator testing performed in the Model Boiler No. 2 (MB-2) facility located at the Westinghouse Engineering Test Facility in Tampa, Florida provided data to better qualify the actual variation in these key parameters. A conclusion of this analysis is that the MB-2 steam line break data base is accurate and of sufficient detail to provide a valuable basis for making comparisons relative to code predictions. Results obtained using the FLASH transient safety analysis code were found to be in excellent agreement with the data.
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 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 Once through Steam generator Sensitivity Calculations written by and published by . This book was released on 1988 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of TRAC-PF1/MOD2 thermal-hydraulic calculations has been performed to determine the effect of uncertainties in modeling once through steam-generator (OTSG) secondary-side phenomena on the calculated behavior of Babcock and Wilcox power plants. The calculations were performed by varying parameters in correlations for the secondary-side phenomena. The parameters and transients were chosen to show the maximum expected sensitivity of the calculated results to the parameter variations. The parameters were then varied over a range representing the estimated uncertainty in the correlation. In this manner, the sensitivity if the calculated plant behavior to the modeling uncertainties was determined with a reasonable number of calculations. The sensitivity of calculated plant behavior to variations in interfacial heat-transfer in the OTSG secondaries was determined in a series of steam-generator overfill transient calculations. Calculations were performed for a main steam line break (MSLB) transient to quantify the sensitivity to variations in interfacial drag in the secondaries; the interfacial drag was varied in these calculations to indicate the effects of entrainment and de-entrainment processes, for which no specific models exist in the code. In addition to the transient calculations, a series of steady-state calculations was performed to determine the sensitivity of the OTSG primary-to-secondary heat transfer to the assumed fraction of tubes wetted by the auxiliary feedwater (AFW) injection. The plant model used for the sensitivity calculations was qualified by performing a benchmark calculation for a natural circulation test in the TMI-1 plant.
Download or read book Loss of Feed Flow Steam Generator Tube Rupture and Steam Line Break Thermohydraulic Experiments written by and published by . This book was released on 1986 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.
Download or read book Steam Generator Tube Failures written by and published by . This book was released on 2001 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.
Download or read book Evaluation of Steam Generator U tube Integrity During PWR Station Blackout with Secondary System Depressurization written by and published by . This book was released on 1999 with total page 72 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 Loss of Feed Flow Steam Generator Tube Rupture and Steam Line Break Thermohydraulic Experiments written by O. J. Mendler and published by . This book was released on 1986 with total page 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 Carryover and Heat Transfer During a Steam Line Break Accident written by Alexander George Parlos and published by . This book was released on 1985 with total page 312 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 Failure Investigation of Boiler Tubes A Comprehensive Approach written by Paresh Haribhakti and published by ASM International. This book was released on 2018 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: Failures or forced shutdowns in power plants are often due to boilers, and particularly failure of boiler tubes. This comprehensive resource deals with the subject of failure investigation of boiler tubes from basic fundamentals to practical applications. Coverage includes properties and selection of materials for boiler tubes from a metallurgical view point, damage mechanisms responsible for failure of boiler tubes, and characterization techniques employed for investigating failures of boiler tubes in thermal power plants and utility boilers of industrial/commercial/institutional (ICI) boilers. A large number of case studies based on the actual failures from the field are described, along with photographs and microstructures to allow for easy comprehension of the theory behind the failures. This book is geared to practicing engineers and for studies in the major area of power plant engineering. For non-metallurgists, a chapter has been devoted to the basics of material science, metallurgy of steels, heat treatment, and structure-property correlation. A chapter on materials for boiler tubes covers composition and application of different grades of steels and high temperature alloys currently in use as boiler tubes and future materials to be used in supercritical, ultra-supercritical and advanced ultra-supercritical thermal power plants. A comprehensive discussion on different mechanisms of boiler tube failure is the heart of the book. Additional chapters detailing the role of advanced material characterization techniques in failure investigation and the role of water chemistry in tube failures are key contributions to the book. The authors have long-standing experience in the field of metallurgy and materials technology, failure investigation, remaining life assessment (RLA) and fitness for service (FFS) for industrial plant and equipment, including power plants. They have conducted a large number of failure investigations of boiler tubes and have recommended effective remedial measures in problem solving for power and utility boilers.
Download or read book Loss of feedwater Steam Generator Tube Rupture and Steam Line Break Experiments written by O. J. Mendler and published by . This book was released on 1987 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt:
