PSI - Issue 71

Ganesh Nigudage et al. / Procedia Structural Integrity 71 (2025) 218–225

219

SPT T SP

Small Punch Test Ductile to brittle transition temperature determined from Small punch test

v

Punch displacement (mm)

v f

In SP testing, punch displacement corresponding to a 20 % force drop with respect to maximum force (mm)

v m

In SP testing, punch displacement at maximum force (mm)

Yield strength in tensile testing Tensile strength in tensile testing

The core of the light water nuclear reactor is housed by Reactor pressure vessel (RPV). It is primary pressure boundary and operates at temperature of 270-320 °C for entire design life of 40 years or more and is irreplaceable part of reactor. So, safety and integrity of RPV is very important. Low alloy steels are used as structural material for reactor pressure vessel because of their high temperature strength, toughness, and good weldability. There are two broad types of low alloy steels that are used to manufacture RPV-a) Mn-Ni-Mo steels (Western grades), b) Cr-Mo-V-Ni steels (Eastern grades). The chemical composition and heat treatment (quenching followed by tempering) are chosen to ensure the required values of mechanical strength, ductility, toughness and satisfactory weldability are achieved. Operation of RPV in high temperature results in thermal ageing embrittlement. Thermal ageing is time and temperature dependent phenomena and involves diffusion of impurities and alloying elements and changes in the microstructure. Degradation in mechanical properties of RPV involves two mechanisms- 1) segregation of impurities like P, Sn, As to grain boundaries/interfaces is termed as non-hardening mechanism of embrittlement as no new barrier is formed due to phosphorus segregation, 2) changes in microstructure by precipitation or coarsening of precipitates which are prone to decohesion at precipitate matrix interface during deformation. Any one mode of degradation cannot be generalised for thermal embrittlement because it depends on various factors like chemical composition, heat treatment, microstructure, thermal ageing temperature (Tenneti Sharma et. al., 2019). The surveillance specimens obtained from the thermal zone of reactor are tested at regular intervals to know about the current state of RPV due to thermal ageing degradation of reactor. In order to predict the properties of RPV after long exposure time, either extrapolation of properties of thermal surveillance specimens can be done or accelerated thermal ageing, to get the idea of change in mechanical properties in advance, can be performed at temperature higher than reactor operating temperature. Present study involves the 2 nd approach i.e. use of accelerated thermal ageing. In this study long term thermal exposure (in reactor at around 320 o C) of Cr-Mo-V-Ni steel is simulated by performing accelerated thermal ageing at 500° C for 1200 and 1800 hours. Tensile properties of the as-fabricated and thermal aged base specimens have been evaluated using small punch test (SPT) to study the change in mechanical properties due to thermal ageing. Details of small punch test set-up and steps used to generate the correlation constants for estimating mechanical properties from SPT have been explained in (Ganesh et al, 2024). This is the first-time thermal ageing studies have been carried out on existing Indian nuclear reactor RPV material.

2. Experimental Details 2.1. Material and thermal ageing

The material used in this study is as-fabricated base material of Cr-Mo-V-Ni based low alloy steel in quenched and tempered condition. Half portion of Charpy specimens having length of 27.5mm and width and thickness both 10mm obtained after impact testing of standard Charpy V-notch specimens have been used for accelerated thermal ageing treatment. Typical chemical composition of the steel used in this study is given in Table 1.

Table 1. Typical chemical composition of the steel in wt%.

C

Mn

Si

P

Cr

Ni

Mo

V

S

0.13-0.18 0.3-0.6 0.10 In the present study, ageing time calculation is based on diffusion and segregation of phosphorus at prior austenite grain boundaries assuming that single degradation mechanism is occurring at high temperature. Thermal ageing carried out at 500 o C for 1200 and 1800 hours. The specimens were vacuum sealed in quartz capsules to avoid oxidation during thermal ageing and were placed in muffle furnace with the help of stainless-steel container. A K-type thermocouple was used to monitor the temperature. 0.17-0.37 0.010 1.8-2.3 0.4-1.0 0.5-0.7 0.10