PSI - Issue 71

Available online at www.sciencedirect.com

ScienceDirect

Procedia Structural Integrity 71 (2025) 218–225

a Post Irradiation Examination Division, Nuclear Fuels Group, Bhabha Atomic Research Centre, Trombay, Mumbai-400085, India , P. P. Nanekar Evaluating Effect of Accelerated Thermal Ageing on RPV Steel 5 th International Structural Integrity Conference & Exhibition (SICE 2024) Abstract Reactor pressure vessel (RPV) is the primary pressure boundary component of a light water nuclear power plant. It operates under harsh environments such as high temperature, high pressure, neutron irradiation, corrosion and fatigue for a long time. These lead to change in mechanical properties of RPV. In the present work a study was carried out to evaluate the effect of thermal ageing on tensile and microstructural properties of Cr-Mo-V-Ni low alloy RPV steel. Thermal aging refers to a phenomenon that the microstructure of material changes when operating under high temperature environment for a long time, and this leads to changes in the mechanical properties (Linjun Zhou et. al., 2022). Long term service of nuclear Reactor Pressure Vessel (RPV) materials in high temperature environment leads to thermal aging embrittlement. Archive (as fabricated) base material (Cr-Mo-V-Ni steel) of a reactor pressure vessel was subjected to accelerated thermal ageing at 500 o C for 1200 and 1800 h. Small punch test (SPT) was used to evaluate tensile properties of archive and thermal aged specimens. The results indicate insignificant change in tensile properties of steel for the ageing time studied. However, there was observation of cleavage feature on fracture surface of 1800 h aged specimen compared to ductile failure for archive and 1200 h aged specimens. No change in hardness, grain size, carbide number density was observed in aged specimen compared to archive specimens. SPT of as-fabricated base was also carried out at sub-ambient temperatures with the aim to determine DBTT of RPV. However, DBTT could not be evaluated because the lower shelf energy was not reached even at lowest temperature of -130 o C. © 2025 The Authors. Published by ELSEVIER B.V. This is an open access article under the CC BY-NC-ND license (https://creativecommons.org/licenses/by-nc-nd/4.0) Peer-review under responsibility of SICE 2024 organizers Ganesh Nigudage * , Priti Kotak Shah, J. S. Dubey

Keywords: Reactor pressure vessel steel; Accelerated thermal ageing; Small Punch Test; tensile properties.

* Corresponding author. Tel.: +91-22-25596111; fax: +91-22-25505151. E-mail address: gannig@barc.gov.in

1. Introduction

Nomenclature ε t E SP E m

In tensile testing, total elongation (%) SP fracture energy is calculated as the area under the force-displacement curve up to v f . SP total energy is calculated as the area under the force-displacement curve up to v m.

F m

In SP testing, maximum force (N)

F 0.1mm, offset

In SP testing, force at the intersection between the test record and a line parallel to the slope of the initial linear region with an offset of 0.1 mm (N)

h o

In SP testing, initial specimen thickness (mm)

2452-3216 © 2025 The Authors. Published by ELSEVIER B.V. This is an open access article under the CC BY-NC-ND license (https://creativecommons.org/licenses/by-nc-nd/4.0) Peer-review under responsibility of SICE 2024 organizers 10.1016/j.prostr.2025.08.030