Evaluation of Mechanical Properties of the Reactor Pressure Vessel Materials Changes by Small Punch Test Application

Abstract

The reactor pressure vessel (RPV) is the most important component of nuclear power plants. RPV steel near the reactor core is subject of irradiation degradation due to the fast neutron flux. Irradiation processes are rather complex but after all the damage of the steel crystal lattice lead to the changes of RPV mechanical properties as well as the shift of the transition temperature to higher values. Hence, monitoring of the RPV material irradiation changes must be proved during the all nuclear power plant (NPP) operation. The new surveillance specimen programs (SSP) at all Slovak NPPs reactors included, among the standard mechanical tests, also new types of evaluation mechanical properties due to method Small Punch Test (SPT).

Interim results of the reactor pressure vessel materials evaluation within the framework of the implemented Advanced Surveillance Specimen Programme

“…SPT practice requires a small volume of tested materials what substantially reduces the total activity of the individual irradiated specimen. There's no need for special hot cells when testing SPT sample, a simple lead brick shielding of a testing machine is adequate [6,7]. Because there is a lack of RPV experimental materials, the great advantage is using the SPT methods for evaluation of their basic mechanical properties [8].…”

“…Using this specimen holder, the specimens are prevented from cupping upward during punching, and therefore, plastic deformation is concentrated in the region below the punch rod. Using relatively simple system with recorders of the load and deflection values, we can obtain the following data of basic mechanical properties [6,7]:…”

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

“…Sampling locations are defined with care so that there is no requirement for subsequent repair or monitoring. Some of the small specimens have been irradiated in the Halden research reactor at a fluence of 4.02 × 10 23 n m −2 [79], which showed an expected increase in the tensile and yield stress properties. Comparisons between the fracture appearance transition temperature values obtained from the SP tests and standard charpy tests provided good agreement between material in the initial state and the irradiated condition.…”

“…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.…”

“…Material properties of the heat affected zone of welds, and the ductile-to-brittle transition temperature (DBTT) can be determined by the use of the SP testing technique [78,79].…”