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).

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This paper deals with the evaluation of material properties of safety-related components of the primary circuit of nuclear power plants (NPP) such as reactor pressure vessel (RPV), primary piping (PP) or steam generator (SG). The main degradation mechanism of NPP’s components is radiation damage, but these processes are situated only in the core region of the reactor pressure vessel (RPV). The main mechanical loading of all individual parts of NPP’s primary circuit are the influence of high pressure at elevated temperature till 300°C. These processes lead to the fatigue damage of structural materials, which is characterized as the thermal fatigue or thermal ageing. The reactors in Slovakia have been behind several decades in operation. Therefore, to ensure the safe and reliable operation of NPP, it is necessary to monitor and evaluate these changes. While the monitoring of the radiation degradation is standardly performed using surveillance programs during all plant life operation, the monitoring of thermal ageing at primary circuit components was realized only after few years of all NPP units operation. The concept of the thermal ageing assessment was divided into two possible approaches to evaluation. The first is long-term exposition of surveillance specimens. And the second approach is a direct surface sampling of heavy components after several years of operation. The Small Punch Test (SPT) methods are mainly used for evaluation of materials actual state. By the SPT technique it is possible to evaluate the basic tensile properties as the ultimate tensile strength and the yield stress of the tested materials from a very small amount of obtained material. The details of the original VUJE design program of thermal ageing monitoring of NPP primary circuit materials in Slovakia is described in this paper.

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.

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The paper deals with the evaluation of mechanical properties of safety-related components of the primary circuit of nuclear power plants (NPPs). During a long-term operation of NPPs, changes of mechanical properties occur. To ensure the safe operation of NPP it is necessary to monitor and evaluate these changes. One possibility how to solve this problem is a direct sampling and the assessment of the actual mechanical properties using the small punch test (SPT) technique. By the SPT technique it is possible to evaluate the basic tensile properties such as the ultimate tensile strength and the yield stress of the tested materials as well as the transition temperature. The authors describe model examples of surface sampling using the special Rolls Royce equipment SSam™-2. The SPT specimens were prepared from the removed samples by application of the wire electrical discharge machining and fine grinding. The prepared specimens were tested and the obtained data were evaluated. Original pieces of primary tubes with diameter 500 mm made of stainless steel type 08Ch18N10T (AISI 321) were used for these model examples. Another experimental material used was a block of the reactor pressure vessel wall cut from NPP Greifswald Unit 7 which contains a weld joint. Those experiments were a part of the preparatory activities for the planned sampling and the assessment of the actual mechanical properties of primary components such as the main primary coolant piping, the steam generator shell and the reactor pressure vessel.

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During a long-term operation of nuclear power plants (NPP), the changes of structural material properties occur. To ensure the safe and reliable operation, it is necessary to monitor and evaluate these changes mainly on components from primary circuit of NPPs. One of the dominant ageing mechanisms of NPP components besides the radiation embrittlement and the fatigue loads is the thermal ageing. The thermal ageing is the temperature, material and time dependent degradation mechanisms due to long-term exposure at the operating temperature of 570 K.This paper describes the project for thermal ageing monitoring at primary piping in NPP Bohunice Unit 3. There are summarized the results obtained from evaluation of original primary piping material.