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research-article

Progression of Severe Accidents, Containment Performance and Estimates of releases of radionuclides-Level-2 PSA IPHWR

[+] Author and Article Information
Rajee Guptan

Nuclear Power Corporation of India Limited, C-3 NUB, Anushaktinagar, Mumbai
grajee@npcil.co.in

A. Vijaya

Nuclear Power Corporation of India Limited, C-3 NUB, Anushaktinagar, Mumbai
akvijaya@npcil.co.in

Vibha Hari

Nuclear Power Corporation of India Limited, C-3 NUB, Anushaktinagar, Mumbai
vibhahari@npcil.co.in

Rajeev Nama

Nuclear Power Corporation of India Limited, B-3 NUB, Anushaktinagar, Mumbai
rnama@npcil.co.in

1Corresponding author.

ASME doi:10.1115/1.4035783 History: Received July 19, 2016; Revised January 06, 2017

Abstract

Probabilistic Safety Assessment (PSA) of Nuclear Power Plants are 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 Pressurised 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 SBLOCA and SBO followed by containment isolation failure. The results of this are used as an input for developing the SAMG measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced LERF.

Copyright (c) 2017 by ASME
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