Utilization of SPT Techniques for Evaluation of Changes in the Properties of the VVER-440 Reactor Pressure Vessel Steels After Their Irradiation in the Research and Power Reactors

Abstract

The paper describes the testing procedures and the basic results of the evaluation of the Small Punch Test (SPT) specimens after their irradiation in the Halden reactor in Norway. The SPT technique was used for estimation of basic mechanical properties as ultimate tensile strength and yield stress of the tested materials as well as the Fracture Appearance Transition Temperature (FATT). The main aim of the work was to compare the SPT results obtained from the surveillance specimen programs implemented in the Slovak power reactors with the SPT results from the specimens irradiated in the research reactor in Halden. For the project there were chosen 3 types of steels used for construction of the reactor VVER 440/213 type in Bohunice NPPs in the Slovak Republic. The experimental materials were two bainitic steels — base metal and weld metal of the reactor pressure vessel wall and austenitic cladding of the reactor wall. Two sets of SPT specimens together with mini-tensile specimens prepared from the experimental materials were irradiated in the Halden reactor. The samples were irradiated at 275°C to two fluence values which are equivalent to approximately of 4 and 6 campaigns in the power reactor. Obtained results are compared to up-to-date SPT results from the surveillance specimen program applied at Bohunice NPP in the Slovak Republic.

Small specimen techniques for estimation of tensile, fatigue, fracture and crack propagation material model parameters

“…92 Since FATT from small punch specimens and Charpy specimens was linearly correlated, small punch test results could be used to estimate bulk FATT. 36 It was observed that irradiation shifted FATT estimated from SPT results to higher temperatures for ferritic materials 23 which generally agreed with standard specimen testing results.…”

“…Fracture appearance transition temperature (FATT) was investigated for steels exclusively 23,28,36,92,95 using small punch specimens. One of the specimens had a semi-circular notch (Figure 23(c)) 92 and another had a notch through thickness (Figure 22(a)).…”

The role of small specimen creep testing within a life assessment framework for high temperature power plant

“…Currently the SP testing technique has gained an important role for monitoring nuclear power plant operation and critical areas, such as the heat affect zone of welds. In particular, the material state of all Slovak reactor pressure vessels is monitored through SP testing (several thousand material samples) as part of the advanced surveillance specimen programmes, which aims to extend the reactor pressure vessels life in the WWER-440 pressurised light water reactors by at least 20 years [79,80], with typical operating conditions at 297°C and 123 bar [81]. The focus of this surveillance programme is to provide a through life assessment of the tensile strength and ductile–brittle transition temperature at various locations within the reactor, with material extracted using the Rolls-Royce SSamTM-2 device with comparison against reference material blocks.…”

“…Based on the current maturity level of the testing techniques outlined in the section ‘The role and application of small specimen creep testing’, it is proposed that the most practical approach to consider at the moment is the further development and deployment of the impression creep method, particularly since it has been exploited on ageing plant by Brett [13–15,109] for purposes of component risk ranking and condition assessment. The extensive programme of sampling undertaken on Slovakian reactors [79,80,110] is also noteworthy and is another example of the proactive and systematic use of these small specimen testing techniques. It is worth noting that other methods such at the two-bar creep test specimen may provide additional benefits due to the potential to model the tertiary phase of creep, however further work on miniaturisation is likely to be required.…”

“…(a) Variations of minimum creep strain rates at 600°C and at a stress of 155 MPa, with room temperature hardness for Grade P91 steels (with different service histories) from Brett [12], and (b) variations of uniaxial rupture life, t f , at 600°C and at a stress of 155 MPa, with room temperature hardness for Grade P91 steels (with different service histories), from Brett [12]. Slovak reactor pressure vessels is monitored through SP testing (several thousand material samples) as part of the advanced surveillance specimen programmes, which aims to extend the reactor pressure vessels life in the WWER-440 pressurised light water reactors by at least 20 years [79,80], with typical operating conditions at 297°C and 123 bar [81]. The focus of this surveillance programme is to provide a through life assessment of the tensile strength and ductile-brittle transition temperature at various locations within the reactor, with material extracted using the Rolls-Royce SSamTM-2 device with comparison against reference material blocks.…”