Boiler Stack Economizer Tube Failure

Keyword(s):  
2013 ◽  
Vol 13 (5) ◽  
pp. 513-520
Author(s):  
Ryan J. Haase ◽  
Larry D. Hanke
Keyword(s):  

1975 ◽  
Author(s):  
M.K. Carlson
Keyword(s):  

2020 ◽  
Vol 7 (3) ◽  
pp. 19-00548-19-00548
Author(s):  
Akihiro UCHIBORI ◽  
Hideki YANAGISAWA ◽  
Takashi TAKATA ◽  
Jiazhi LI ◽  
Sunghyon JANG

Author(s):  
David Cho ◽  
Danny H. B. Mok ◽  
Steven X. Xu ◽  
Douglas A. Scarth

Technical requirements for analytical evaluation of in-service Zr-2.5Nb pressure tubes in CANDU(1) reactors are provided in the Canadian Standards Associate (CSA) N285.8. The evaluation must address all in-service degradation mechanisms including the presence of in-service flaws. Flaws found during in-service inspection of CANDU Zr-2.5Nb pressure tubes, including fuel bundle scratches, debris fretting flaws, fuel bundle bearing pad fretting flaws, dummy bundle bearing pad fretting flaws, erosion-shot flaws and crevice corrosion flaws, are volumetric and blunt in nature. These in-service flaws can become crack initiation sites during pressure tube operation and potentially lead to pressure tube failure. Any detected flaws that do not satisfy the criteria of acceptance as per CSA N285.4 must be analytically evaluated to justify continued operation of the pressure tube. Moreover, the risk of pressure tube failure due to presence of in-service flaws in the entire reactor core must be assessed. A review of assessment of the risk of pressure tube failure due to presence of in-service flaws in CANDU reactor core is provided in this paper. The review covers the technical requirements in the CSA N285.8 for evaluating degradation mechanisms related to flaws in the reactor core. Current Canadian industry practice of probabilistic assessment of reactor core for pressure tube failure due to presence of in-service flaws is described, including evaluation of flaws for crack initiation, subsequent crack growth to through-wall penetration, and pressure tube rupture due to unstable crack growth prior to safe shutdown of the reactor. Operating experience with the application of probabilistic assessment of reactor core for the risk of pressure tube failure due to presence of in-service flaws is also provided.


Author(s):  
Akihiro Uchibori ◽  
Shin Kikuchi ◽  
Akikazu Kurihara ◽  
Hirotsugu Hamada ◽  
Hiroyuki Ohshima

Multiphysics analysis system was newly developed to evaluate possibility of failure propagation occurrence under heat transfer tube failure accident in a steam generator of sodium-cooled fast reactors. The analysis system consists of the computer codes, SERAPHIM, TACT, RELAP5, which are based on the mechanistic numerical models. The SERAPHIM code calculates the multicomponent multiphase flow involving sodium-water chemical reaction. In this study, numerical models for the chemical reaction about production of a sodium monoxide and its transport process were constructed to enable evaluation of a wastage environment. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The TACT code was integrated by the numerical models of the fluid-structure thermal coupling, the temperature and stress evaluation, the wastage evaluation and the failure judgment. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the rapidly heated tube in the present work.


2015 ◽  
Vol 30 (2) ◽  
pp. 187
Author(s):  
Rakesh Kumar ◽  
AnilKumar Pandey ◽  
Sellam Karunanithi ◽  
ChetanD Patel ◽  
SanjayKumar Sharma ◽  
...  

Sign in / Sign up

Export Citation Format

Share Document