Different approaches to estimation of reactor pressure vessel material embrittlement

The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV) material embrittlement. The beltline materials (base and weld metal) were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF) has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt). Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal

Evaluation of WWER-1000 vessel materials fracture toughness

The lifetime of WWER-1000-type reactor vessels is finally conditioned by the fracture toughness (crack growth resistance) of RPV materials. Up to now in line with the regulations the fracture toughness is characterized by the critical temperature of brittleness determined by the results of the Charpy specimen impact testing. Such approach is typical for all countries operating the water pressure reactors. However, regulatory approach is known from the western specialists not always to characterize adequately the crack growth resistance of the vessel materials and in some cases to underestimate their characteristics in the reference state that leads to unreasonably high conservatism. Excessive conservatism may lead to the invalid restrictions in the operating modes and the service life of the reactor vessel. Therefore there appeared the necessity to apply another approaches based on the state-of-the-art experimental methods of the fracture mechanics and allowing evaluating the fracture toughness parameters sufficiently. The paper presents the results of the comparison of the regulatory approach and the Master curve approach from the point of view of the adequate determination of the vessel material crack growth resistance parameters. Analysis of the experimental data of the surveillance specimens illustrated the potential possibility of applying the new statistical method for the WWER-1000- type reactor vessel lifetime extension

VVER-1000 reactor vessel metal state monitoring in Ukraine

On the basis of the “hot” cells of the Institute for Nuclear Research of National Academy of Sciences of Ukraine reconstitution technique of VVER reactor vessel metal surveillance specimens, and the main provisions of the reactor vessel radiation loading monitoring procedures and surveillance specimen dosimetry are described in the paper. The first results of the reconstituted irradiated surveillance specimen tests are presented, which together with the data on the radiation loading of the reactor vessel provide information for substantiation of the reactor safe operation term from the point of view of the brittle fracture resistance

Stress Effect on Embrittlement of WWER-440 Reactor Pressure Vessel Steels.

Abstract not availableJRC.F-Institute for Energy (Petten

TAREG 2.01/00 Project - Validation of Neutron Embrittlement for VVER 1000 and 440/213 RPVs, with Emphasis on Integrity Assessment

The irradiation embrittlement and integrity of the VVER reactors has been an important issue in many EC supported TACIS and PHARE projects since 1990. In the EC annual programs two TACIS projects (TAREG 2.01/00 and 2.01/03) were launched on the issue in order to improve the neutron irradiation embrittlement databases, elaborate new trend curves for the embrittlement and to assess the integrity of the RPVs (Reactor Pressure Vessel) by analysing PTS transients (Pressurized Thermal Shock) for some selected Russian and Ukrainian VVER 1000 and 440/213 NPPs. In this paper the TAREG 2.01/00 project is briefly described with some details from the twin project 2.01/03, which served as a materials testing project, providing inputs for the 1st project. As a result of the project new trend curves for neutron irradiation embrittlement was elaborated based on upgraded and more reliable surveillance results databases. The PTS study show that the integrity of the selected VVER RPVs can be ensured to the end of RPV design life.JRC.F.5-Nuclear Reactor Safety Assessmen