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Research Papers

Experimental Demonstration of Safety of AHWR during Stagnation Channel Break Condition in an Integral Test Loop

[+] Author and Article Information
Mukesh Kumar

Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre Trombay,
Mumbai 400 085, India
e-mail: mukeshd@barc.gov.in

A. K. Nayak, Sumit V. Prasad, P. K. Verma, R. K. Singh, Vikas Jain, D. K. Chandraker

Reactor Engineering Division,
Reactor Design and Development Group,
Bhabha Atomic Research Centre Trombay,
Mumbai 400 085, India

1Corresponding author.

Manuscript received May 16, 2017; final manuscript received December 6, 2017; published online March 5, 2018. Assoc. Editor: Jovica R. Riznic.

ASME J of Nuclear Rad Sci 4(2), 021005 (Mar 05, 2018) (6 pages) Paper No: NERS-17-1053; doi: 10.1115/1.4038899 History: Received May 16, 2017; Revised December 06, 2017

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

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References

Sinha, R. K. , and Kakodkar, A. , 2006, “ Design and Development of the AHWR—the Indian Thorium Fuelled Innovative Nuclear Reactor,” Nucl. Eng. Des., 236(7–8), pp. 683–700. [CrossRef]
Sharma, M. , Pilkhwal, D. S. , Vijayan, P. K. , and Sinha, R. K. , 2011, “ Stagnation Channel Break Analysis of AHWR and ITL Using RELAP5/Mod 3.2 Computer Code,” 21st National and 10th ISHMT-ASME Heat and Mass Transfer Conference, Chennai, India, Dec. 27–30, pp. 33–40.
Singh, R. K. , and Rama Rao, A. , 2011, “ Steam Leak Detection in Advance Reactors Via Acoustics Method,” Nucl. Eng. Des., 241(7), pp. 2448–2454. [CrossRef]
Rao, G. S. S. P. , Vijayan, P. K. , Jain, V. , Borgohain, A. , Sharma, M. , Nayak, A. K. , Beloker, D. G. , Pal, A. K. , Saha, D. , and Sinha, R. K. , 2002, “AHWR Integral Test Loop Scaling Philosophy and System Description,” Bhabha Atomic Research Centre, Mumbai, India, Report No. BARC/2002/E/017. https://inis.iaea.org/search/search.aspx?orig_q=RN:34073647

Figures

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Fig. 2

Schematic of emergency core cooling system

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Fig. 3

Schematic of MHTS in AHWR (total 452 channels only two representative channels are shown)/ITL (two channels) and probable locations of stagnation channel break

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Fig. 4

Integral test loop

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Fig. 5

Fuel channel simulators of ITL

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Fig. 6

Setup for stagnation channel break

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Fig. 7

Channel flow and FCS power at 7 MPa

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Fig. 8

Acoustic signature during the test at 7 MPa

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Fig. 9

Steam drum pressure and level at 7 MPa

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Fig. 10

Heater pin surface temperature at 7 MPa

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