How to cite:Adamech, M., Petzová, J., Březina, M., & Kapusňák, M. 2018. Using of SPT method for estimation of mechanical properties changes of RPV steels after irradiation in the Halden reactor. Ubiquity Proceedings, 1(S1): 2Abstract: The paper deals with experimentally estimation and comparison of the mechanical properties changes of RPV steels before and after irradiation of samples in Halden reactor in Norway coming from Unit #3 and #4 of NPP Mochovce (still under construction). Altogether 180 SPT and 30 mini-tensile samples in two sets were prepared for irradiation, obtained from weld metal material (Sv10ChMFT) and base material (15Ch2MFA). In general, a good agreement between results obtained by SPT technique and using mini-tensile specimens was found. Both, base and weld metals of RPVs were found to be bainitic. After that, the first set of samples was irradiated in Halden reactor at temperature Tirr = 270 -280°C with intention to use two fluence values: ∼ 1.0 × 10 24 n/m 2 and ∼ 2.0 × 10 24 n/m 2 (> 1 MeV), respectively. Specimens after 1st irradiation were successfully tested and preliminary results show small increase of the strength characteristics (R e , R m ) if compare to "zero condition" testing results. FATTs, evaluated by the temperature dependence of the SPT energy, exhibit transition behaviour and shift towards higher temperatures.
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.
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 reactor pressure vessel (RPV) is the most critical component of every nuclear power plant (NPP) and continuous evaluation of its mechanical properties is a necessity for long and safe operation. Standard tests require a collection of large-dimension samples coming from the precious and archive materials, usually produced as control segments. Since SPT testing samples are quite small, high activity of irradiated materials is no longer an issue. The SPT technique therefore represents a very useful and effective method applied for characterization of mechanical properties such as ultimate tensile strength (Rm), yield strength (Re), and fracture appearance transition temperature (FATT). Monitoring of structural components in nuclear power plants receives much attention, particularly, in the context of long term operation (LTO) of current plants where the amount of material available for destructive testing is considerably limited.
By the non-destructive testing of a dissimilar weld joint (DWJ) of cold collector DN 1100 (CC) on a steam generator, indications were found on inner-side cold collector’s surface at the root position of the examined weld. All the identified indications were very similar in shape and form, therefore, it was decided to cut out a part of the damaged site from this type of DWJ DN 1100 and get the obtained ring (real piece of material) for complex metallographic analysis. This paper briefly describes the results and recommendations found for the future reference during the next long‑term operation induced ageing and degradation of critical steam generator parts in NPP Bohunice Unit 4. There are summarized the results obtained from evaluation of original DWJ material.
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