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Special Section Papers

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

[+] Author and Article Information
Sumit V. Prasad

Reactor Engineering Division,
Bhabha Atomic Research Centre,
Trombay 400085, Mumbai, India
e-mail: svprasd@barc.gov.in

A. K. Nayak

Reactor Engineering Division,
Bhabha Atomic Research Centre,
Trombay 400085, Mumbai, India
e-mail: arunths@barc.gov.in

1Corresponding author.

Manuscript received July 8, 2016; final manuscript received December 13, 2016; published online March 1, 2017. Assoc. Editor: Thambiayah Nitheanandan.

ASME J of Nuclear Rad Sci 3(2), 020909 (Mar 01, 2017) (8 pages) Paper No: NERS-16-1066; doi: 10.1115/1.4035691 History: Received July 08, 2016; Revised December 13, 2016

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 future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

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References

Joshi, J. B. , Nayak, A. K. , Singhal, M. , and Mukhopadhaya, D. , 2013, “ Core Safety of Indian Nuclear Power Plants (NPPs) Under Extreme Conditions,” Sadhana, 38(5), pp. 945–970. [CrossRef]
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Muktibodh, U. C. , 2011, “ Design, Safety and Operability Performances of 220MWe, 540MWe and 700Mwe PHWRs in India,” Inter-Regional Workshop on Advanced Nuclear Reactor Technology for Near Term Deployment, July 4–8, IAEA, Vienna, Austria.
Mathew, P. M. , Nitheanandan, T. , and Bushby, S. J. , 2008, “ Severe Core Damage Accident Progression Within a CANDU 6 Calandria Vessel,” ERMSAR Seminar, 3rd European Review Meeting on Severe Accident Research, Nesseber, Bulgaria, Sept. 23–25.
Sumit, V. P. , Nayak, A. K. , Kulkarni, P. P. , Vijayan, P. K. , and Vaze, K. K. , 2015, “ Study on Heat Removal Capability of Calandria Vault Water From Molten Corium in Calandria Vessel During Severe Accident of a PHWR,” Nucl. Eng. Des., 284, pp. 130–142. [CrossRef]

Figures

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Fig. 1

(a) PHWR core assembly and (b) coolant channel

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Fig. 2

Severe accident progression inside the calandria vessel

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Fig. 3

Experiment setup: (a) 2D schematic with no decay heat, (b) 2D schematic with decay heat, (c) 3D schematic, and (d) actual setup

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Fig. 4

(a) Schematic of stepped calandria and (b) stepped calandria vessel setup

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Fig. 5

Without decay heat: (a) temperature distribution inside molten pool at different radial heights, (b) crust thickness variation, (c) temperature distribution of inner and outer surface at different circumferential location, and (d) picture from top of the solidified melt after the test

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Fig. 6

With decay heat: (a) temperature distribution inside molten pool at different radial heights, (b) crust thickness variation, (c) temperature distribution of inner and outer surface at different circumferential location, and (d) picture from top of the solidified melt after the test

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Fig. 7

(a) Temperature distribution inside molten pool at different radial heights, (b) crust thickness variation with time, (c) temperature distribution in stepped test section shell, and (d) strain variation in stepped test section

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