Zobrazeno 1 - 10
of 30
pro vyhledávání: '"K K VAZE"'
Publikováno v:
Nuclear Engineering and Technology, Vol 45, Iss 5, Pp 605-612 (2013)
Removal of decay heat from an operating reactor during a prolonged station blackout condition is a big concern for reactor designers, especially after the recent Fukushima accident. In the case of a prolonged station blackout condition, heat removal
Externí odkaz:
https://doaj.org/article/8441f3909e294ba686e4b8fbffd3737b
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2012 (2012)
In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculat
Externí odkaz:
https://doaj.org/article/4139c9f59cd94be2a52e797ec4940f6d
Autor:
K. K. Vaze
Publikováno v:
The Mind of an Engineer: Volume 2 ISBN: 9789811513299
I obtained my B.Tech. degree in Mechanical Engineering from I.I.T. Mumbai in 1973 and joined the 17th batch of Bhabha Atomic Research Centre (BARC) Training School immediately thereafter. After completing the training, I joined Indira Gandhi Centre f
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::b58e70002578042e530b6d33c4ace41d
https://doi.org/10.1007/978-981-15-1330-5_26
https://doi.org/10.1007/978-981-15-1330-5_26
Publikováno v:
Science and Technology of Nuclear Installations, Vol 2011 (2011)
In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners
Externí odkaz:
https://doaj.org/article/457ccdbf9ce94364b4d10539697e842d
Publikováno v:
Fatigue & Fracture of Engineering Materials & Structures. 40:145-156
In this paper, we present and demonstrate a methodology to improve probabilistic fatigue crack growth (FCG) predictions by using the concept of Bayesian updating using Markov chain Monte Carlo simulations. The methodology is demonstrated on a cracked
Publikováno v:
IndraStra Global.
This paper presents the estimation of the reliability levels associated with a cracked pipe found acceptable as per the failure assessment diagram (FAD) based acceptance criteria of ASME Section XI, Appendix H. This acceptance criterion is built on t
Publikováno v:
Kerntechnik. 78:411-421
High Level Liquid Radioactive Waste (HLLRW) produced during reprocessing of spent fuel from nuclear reactors is encased in the canisters after vitrification. The vitrified waste has high heat generation rate due to decay heat and needs interim storag
Publikováno v:
Kerntechnik. 76:98-103
In water cooled power reactors, significant quantities of hydrogen could be produced following a severe accident (loss-ofcoolant-accident along with non availability of Emergency Core Cooling System) from the reaction between steam and zirconium at h
Publikováno v:
IEEE Transactions on Nuclear Science. 58:479-489
Three element Steam Drum (SD) Level Controller has been conventionally used for most of the boilers, Nuclear power plant steam generator & Boiling Water Reactor (BWRs). Based on the process dynamic studies it was found that this scheme does not work
Autor:
K. K. Vaze
Publikováno v:
Transactions of the Indian Institute of Metals. 63:187-193
A component test programme was initiated to address the applicability of leak-before-break to Indian PHWRs. After completing the tests with constant amplitude loading under air environment, the interactive nature of fatigue was addressed. The paramet