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Experimental investigation of sagging of a completely voided pressure tube of Indian PHWR under heatup condition
Nuclear engineering and design, 2010-10, Vol.240 (10), p.3504-3512
2010
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Study of ballooning of a completely voided pressure tube of Indian PHWR under heat up condition
Nuclear engineering and design, 2012-02, Vol.243, p.301-310
2012
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ASTEC adaptation for PHWR limited core damage accident analysis
Nuclear engineering and design, 2014-06, Vol.272, p.273-286
2014
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Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor
Nuclear engineering and design, 2014-10, Vol.278, p.360-366
2014
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Severe accident management strategy verification for VVER-1000 (V320) reactor
Nuclear engineering and design, 2011-09, Vol.241 (9), p.3977-3984
2011
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Experimental investigation of symmetric and asymmetric heating of pressure tube under accident conditions for Indian PHWR
Nuclear engineering and design, 2013-01, Vol.254, p.300-307
2013
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Analyses for VVER-1000/320 reactor for spectrum of break sizes along with SBO
Annals of nuclear energy, 2010-03, Vol.37 (3), p.359-370
2010
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Generic Containment: Detailed comparison of containment simulations performed on plant scale
Annals of nuclear energy, 2014-12, Vol.74, p.165-172
2014
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A diagnostic system for identifying accident conditions in a nuclear reactor
Nuclear engineering and design, 2011, Vol.241 (1), p.177-184
2011
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Uncertainty analysis of large break LOCA for pressurized heavy water reactor
Nuclear engineering and design, 2012-04, Vol.245, p.180-188
2012
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Estimation of source term and related consequences for some postulated severe accident scenarios of Indian PHWR
Nuclear engineering and design, 2010-10, Vol.240 (10), p.3529-3538
2010
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An innovative approach for Steam Generator Pressure Control of a nuclear power plant
Nuclear engineering and design, 2011-12, Vol.241 (12), p.5113-5126
2011
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Level-1, -2 and -3 PSA for AHWR
Nuclear engineering and design, 2011-08, Vol.241 (8), p.3256-3269
2011
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Analysis of PHWR LP Turbine Steady State Using RELAP/SCDAP Code
Energy procedia, 2011, Vol.7, p.367-373
2011
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Estimation of crack opening area for leak before break analysis of nuclear reactor system
Nuclear engineering and design, 2009-02, Vol.239 (2), p.327-337
2009
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Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor
IEEE transactions on nuclear science, 2009-12, Vol.56 (6), p.3712-3725
2009
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Effect of steam environment on severe core damage behaviour for VVER-1000 with the ASTEC V1 code
Nuclear engineering and design, 2009-03, Vol.239 (3), p.559-565
2009
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Uncertainty analysis of LBLOCA for Advanced Heavy Water Reactor
Annals of nuclear energy, 2008-02, Vol.35 (2), p.323-334
2008
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Modelling of thermal and flow stratification for reactor pressure vessel pressurised thermal shock
Nuclear engineering and design, 2002-03, Vol.212 (1), p.75-84
2002
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