Author: Publisher:
ISBN:
Category : Nuclear power plants
Languages : en
Pages : 137
Book Description
Stress Analysis of Pressurized Water Reactor Steam Generator Tube Denting Phenomena
Author: Publisher:
ISBN:
Category : Nuclear power plants
Languages : en
Pages : 137
Book Description
Stress analysis of pressurized water reactor steam generator tube denting phenomena : interim report
Author: Jerrell M. ThomasSteam Generator Group Project
Author: E. B. SchwenkPublisher:
ISBN:
Category : Eddy currents (Electric)
Languages : en
Pages : 120
Book Description
Degradation of Steam Generator Tubing and Components by Operation of Pressurized-water Reactors
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Experience in operating pressurized water reactors (PWR) has shown a number of materials degradation processes to have occurred in their steam generators. These include stress corrosion cracking (SCC), intergranular attack, generalized dissolution, and pitting attack on steam generator tubes; mechanical damage to steam generator tubes; extensive corrosion of tubing support plates (denting); and cracking of feedwater lines and steam generator vessels. The current status of the understanding of the causes of each of these phenomena is reviewed with emphasis on their possible significance to reactor safety and directions the nuclear industry and the NRC should be taking to reduce the rate of degradation of steam generator components.
Energy Research Abstracts
Author: Denting of Inconel Steam Generator Tubes in Pressurized Water Reactors. Final Report
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Denting is a form of damage that results from the rapid, linear growth of magnetite on carbon steel tube support plates in tube-to-support plate crevices. The pressure of the corrosion products distorts the tubes as well as the support plates. Although denting was first thought to be confined to those recirculating steam generators that had been converted from phosphate to AVT, it has now also been observed in plants still on phosphate, as well as in some that started on AVT. In some units, slightly abnormal eddy current signals have been observed at the top of tube sheets. No denting has been observed so far in the B and W once-through steam generators. In all, some 14 plants are affected. Inconel 600 tube defects and leaks were first observed at Surry and Turkey Point, which were severely dented. The cracks originated from the primary side, and were found to be caused by primary side stress corrosion cracking of highly strained tubing. Some minor leaks have subsequently occurred in a few other steam generators. Suggestions are included of work to elucidate the quantitative aspects of stress corrosion cracking.
Tube Failures and Corrosion Problems in Steam Generators of Pressurized Water Reactors
Author: Manubhai C. PatelPublisher:
ISBN:
Category : Steam-boilers
Languages : en
Pages : 290
Book Description
Metallography in Failure Analysis
Author: J. McCallPublisher: Springer Science & Business Media
ISBN: 146132856X
Category : Technology & Engineering
Languages : en
Pages : 301
Book Description
Detailed analyses of failures of material components have proved to be valuable in many ways; by preventing further failures, by assessing the validity of designs and the selection of materials, by uncovering shortcomings in the processing of the materials in volved through characterizations of defects, and by revealing problems introduced during the manufacture or fabrication of the component. Increased recognition of the value of performing failure analyses has caused the field to develop into a very active area of tech nical endeavor. Failure analysis has been employed in numerous different technical dis ciplines and has proven beneficial. The increased activity has caused many new and im proved methods for performing these analyses to be developed. Among these are many methods which can be characterized as generally belonging to the field of metallography. In recognition of the important role that metallography plays in the performance of failure analyses, the absence of a text that specifically discusses this subject, and the be lief that communication of information on the subject would be of technical interest, The American Society for Metals and The International Metallographic Society co sponsored a symposium. The intent was to bring together world-recognized authorities working in various aspects of the failure analysis and metallographic fields to share meth ods they use, results they have obtained, and the purposes to which they utilized these results. The symposium, entitled "Metallography in Failure Analysis", was held in Hous ton, Texas, USA, July 17-18, 1977.
PWR Steamline Break Analysis Assuming Concurrent Steam Generator Tube Rupture
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
Results are presented for a steamline break analysis for a typical, two-loop, 2560 MW(t) pressurized water reactor. The calculations were performed using the IRT reactor system transient analysis code. Included are the analyses of steamline break transients assuming concurrent steam generator tube rupture (up to 30 steam generator tubes). Graphical and tabular results are presented.
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
Author: Publisher:
ISBN:
Category :
Languages : en
Pages :
Book Description
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.