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Special Section Papers

Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

[+] Author and Article Information
Rajee Guptan

Department of Atomic Energy,
Nuclear Power Corporation of India Limited,
C-3 NUB, Anushaktinagar,
Mumbai 400094, India
e-mail: grajee@npcil.co.in

A. K. Vijaya

Department of Atomic Energy,
Nuclear Power Corporation of India Limited,
C-3 NUB, Anushaktinagar,
Mumbai 400094, India
e-mail: akvijaya@npcil.co.in

Vibha Hari

Department of Atomic Energy,
Nuclear Power Corporation of India Limited,
C-3 NUB, Anushaktinagar,
Mumbai 400094, India
e-mail: vibhahari@npcil.co.in

Rajeev Nama

Department of Atomic Energy,
Nuclear Power Corporation of India Limited,
B-3 NUB, Anushaktinagar,
Mumbai 400094, India
e-mail: rnama@npcil.co.in

Manuscript received July 19, 2016; final manuscript received January 6, 2017; published online March 1, 2017. Assoc. Editor: Thambiayah Nitheanandan.

ASME J of Nuclear Rad Sci 3(2), 020910 (Mar 01, 2017) (7 pages) Paper No: NERS-16-1074; doi: 10.1115/1.4035783 History: Received July 19, 2016; Revised January 06, 2017

Probabilistic safety assessment (PSA) of nuclear power plants is performed to yield insights into the safety, design, and performance of the plants and their potential environmental effects. This includes the identification of dominant risk contributors, determination of the vulnerabilities of plant and containment systems, and comparison of options for risk reduction. Three levels of PSA are recognized. Level-1 addresses the identification of plant failures leading to core damage and their frequencies of occurrence. Level-2 addresses the assessment of containment response leading together with level-1 results to the determination of containment release frequencies. A level-2 PSA analyses the challenges to the containment, the possible containment responses and their estimated probabilities, and an assessment of the consequent releases to the environment. Level-3 is the assessment of off-site consequences leading, together with the results of level-2 analysis, for estimation of public risks. A comprehensive level-2 PSA study of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is performed to assess the challenges to the containment, the possible containment responses and their estimated probabilities, and consequent releases to the environment. The dominating sequences consist of small-break loss of coolant accident (SBLOCA) and station black out (SBO) followed by containment isolation failure. The results of this are used as an input for developing the severe accident management guidelines (SAMG) measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced large early release frequency (LERF).

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References

Figures

Grahic Jump Location
Fig. 1

PSA framework (source: IAEA safety series No-P-8)

Grahic Jump Location
Fig. 5

Flow chart for system modeling

Grahic Jump Location
Fig. 6

Fault tree for air locks

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