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    In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues 

    Source: Journal of Nuclear Engineering and Radiation Science:;2017:;volume( 003 ):;issue: 002:;page 20909
    Author(s): Prasad, Sumit V.; Nayak, A. K.
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and ...
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    Experimental Study on Melt Coolability Capability of Calandria Vault Water During Severe Accident in Indian PHWRs for Prolonged Duration 

    Source: Journal of Nuclear Engineering and Radiation Science:;2018:;volume( 004 ):;issue: 003:;page 31014
    Author(s): Prasad, Sumit V.; Nayak, A. K.
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: The present experimental investigation in a scaled facility of an Indian pressurized heavy water reactors (PHWRs) is focused on the heat transfer behavior from the calandria vessel (CV) to the calandria vault during a ...
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    Experimental Evaluation of Critical Heat Flux in Downward-Facing Boiling on SS304 L Flat Plate Relevant to In-Calandria Retention in PHWRs 

    Source: Journal of Nuclear Engineering and Radiation Science:;2020:;volume( 006 ):;issue: 003
    Author(s): Prasad, Sumit V.; Nayak, A. K.
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: Retention of corium inside the calandria vessel (CV) by externally cooling it by calandria vault water is essential to mitigate severe accidents in pressurized heavy water reactor (PHWR). The thermal failure of CV can be ...
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    Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident 

    Source: Journal of Nuclear Engineering and Radiation Science:;2019:;volume( 005 ):;issue: 004:;page 41601
    Author(s): Pandey, Pradeep; Kulkarni, Parimal P.; Nayak, Arun; Prasad, Sumit V.
    Publisher: American Society of Mechanical Engineers (ASME)
    Abstract: In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause ...
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    Retention of Molten Corium in Calandria Vessel of PHWR With Moderator Drain Pipe 

    Source: Journal of Nuclear Engineering and Radiation Science:;2020:;volume( 007 ):;issue: 001:;page 011702-1
    Author(s): Pandey, Pradeep; Kulkarni, Parimal P.; Nayak, Arun; Prasad, Sumit V.
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: Retention of molten corium inside calandria vessel is crucial for arresting accident progression in pressurized heavy water reactors (PHWRs) during severe accidents. Our earlier tests have demonstrated corium retention and ...
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    In-Vessel Retention of PHWRs: Experiments at Prototypic Temperatures 

    Source: Journal of Nuclear Engineering and Radiation Science:;2020:;volume( 006 ):;issue: 001:;page 011601-1
    Author(s): Prasad, Sumit V.; Kulkarni, P. P.; Yadav, D. C.; Verma, P. K.; Nayak, A. K.
    Publisher: The American Society of Mechanical Engineers (ASME)
    Abstract: In pressurized heavy water reactors (PHWRs), multiple failures of engineered safety features may cause a failure of core cooling eventually leading to core collapse. The failed fuel and fuel channels relocate to the bottom ...
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    Experimental Demonstration of Decay Heat Removal by Submerged Feeders in a Full-Scale Test Facility of a Natural Circulation Boiling Water Reactor 

    Source: Journal of Nuclear Engineering and Radiation Science:;2019:;volume( 005 ):;issue: 004:;page 41203
    Author(s): Nayak, A. K.; Kumar, Mukesh; Prasad, Sumit V.; Jain, V.; Chandraker, D. K.
    Publisher: American Society of Mechanical Engineers (ASME)
    Abstract: Removal of decay heat with nonavailability of active systems is a safety issue especially during station blackout (SBO) in a light water reactor. Passive systems are being incorporated in the new designs of nuclear reactors ...
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    Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop 

    Source: Journal of Nuclear Engineering and Radiation Science:;2018:;volume( 004 ):;issue: 002:;page 21005
    Author(s): Kumar, Mukesh; Nayak, A. K.; Prasad, Sumit V.; Verma, P. K.; Singh, R. K.; Jain, Vikas; Chandraker, D. K.
    Publisher: The American Society of Mechanical Engineers (ASME)
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