Decay heat removal for prolonged period of station blackout (SBO) is a safety concern of the nuclear reactors. Aftermath of Fukushima, safety evaluation (performance under severe conditions: stress test) of the reactors was carried out worldwide. It includes establishment of grace period of the reactors. Similar exercises for advanced heavy water reactor (AHWR) were also performed and the design of AHWR was established for its robustness against such events. Decay heat removal during extended SBO is such a condition to be qualified. In this regard, experiments in the integral test loop (ITL), a full scale test facility of AHWR, were conducted for continuous 7 days of extended SBO. Experiment was started with 6.8 MPa as the initial reactor pressure and decay heat removal was demonstrated for 7 days of SBO by passive means. It is observed that the pressure falls down to 1 MPa in 3 h. The design of AHWR was evaluated from safety critical aspects during such an event experimentally. During this event, the clad surface temperature was found to be well within safe limits of operations. As a result of this experiment, it can be concluded that the design of AHWR is capable to remove decay heat for 7 days of SBO with sufficient safety margins.
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October 2019
Research-Article
Experimental Demonstration of Safety During Extended Station Blackout in an Integral Test Loop of a Natural Circulation Boiling Water Reactor
A. K. Nayak,
A. K. Nayak
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay, Mumbai 400 085, India
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay, Mumbai 400 085, India
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Mukesh Kumar,
Mukesh Kumar
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
e-mail: mukeshd@barc.gov.in
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, Indiae-mail: mukeshd@barc.gov.in
1Corresponding author.
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A. K. Vishnoi,
A. K. Vishnoi
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, India
Search for other works by this author on:
Vikas Jain,
Vikas Jain
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, India
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D. K. Chandraker
D. K. Chandraker
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, India
Search for other works by this author on:
A. K. Nayak
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay, Mumbai 400 085, India
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay, Mumbai 400 085, India
Mukesh Kumar
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
e-mail: mukeshd@barc.gov.in
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, Indiae-mail: mukeshd@barc.gov.in
A. K. Vishnoi
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, India
Vikas Jain
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, India
D. K. Chandraker
Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Reactor Design and Development Group,
Bhabha Atomic Research Centre,
Trombay
, Mumbai 400 085, India
1Corresponding author.
Manuscript received September 30, 2018; final manuscript received March 4, 2019; published online July 19, 2019. Assoc. Editor: Liangzhi Cao. This work was prepared while under employment by the Government of India as part of the official duties of the author(s) indicated above, as such copyright is owned by that Government, which reserves its own copyright under national law.
ASME J of Nuclear Rad Sci. Oct 2019, 5(4): 041205 (9 pages)
Published Online: July 19, 2019
Article history
Received:
September 30, 2018
Revised:
March 4, 2019
Citation
Nayak, A. K., Kumar, M., Vishnoi, A. K., Jain, V., and Chandraker, D. K. (July 19, 2019). "Experimental Demonstration of Safety During Extended Station Blackout in an Integral Test Loop of a Natural Circulation Boiling Water Reactor." ASME. ASME J of Nuclear Rad Sci. October 2019; 5(4): 041205. https://doi.org/10.1115/1.4043198
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