SYMPOSIA PAPER Published: 01 January 1992

Evaluation of Tarapur Atomic Power Station Pressure Vessel Surveillance Results


SA 3020 (nickel-modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210-MW boiling water reactors of the Tarapur Atomic Power Station Charpy V-notch impact surveillance specimens representing the pressure vessel belthne base, weld, and the heat-affected zone were irradiated at the wall and shroud location These specimens were evaluated after 6 5 effective full power years of reactor operation The neutron fluences at the locations were 5.31 × 1017 and 4.88 × 1018 n/cm2 (E > 1 MeV), respectively.

The increase in the reference temperature and the decrease in the upper shelf energy obtained from the surveillance program were compared with the corresponding values predicted using USNRC Regulatory Guide 1.99, Revisions 1 and 2 It has been brought out that the regulatory position regarding the heat-affected zone needs to be clarified.

The adjusted reference temperature was estimated from the surveillance results and from Revision 2 of Regulatory Guide 1.99 as 324 K The reference fracture toughness of the vessel wall temperature of 411 K. This ensures the safety of the pressure vessel until the end of the expected 40-year life of the plant.

Author Information

Anantharaman, S
Bhabha Atomic Research Centre, Bombay, India
Chatterjee, S
Bhabha Atomic Research Centre, Bombay, India
Sivaramakrishnan, KS
Bhabha Atomic Research Centre, Bombay, India
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Developed by Committee: E10
Pages: 56–63
DOI: 10.1520/STP17861S
ISBN-EB: 978-0-8031-5187-1
ISBN-13: 978-0-8031-1477-7