INVESTIGATION OF SAMPLES TAKEN FROM KOZLODUY UNIT-2 REACTOR PRESSURE-VESSEL

Within the framework of the 6 month WANG program, small samples were c ut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal chara cteristics required for estimating radiation-limited lifetime-the duct ile-to-brittle transition temperature (DBTT) in the initial state (T-k o) and the phosphorus and copper contents which affect the radiation s tability of steel-were not determined during manufacturing. The Kozlod uy unit 2 pressure vessel had no surveillance program. Radiation stabi lity was evaluated using dependencies based on analysis results for su rveillance samples taken from other VVER-440 reactors. For this reason , the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental wit h calculated data were the principle objectives of the study. Instrume nted impact tests were carried out on sub-size specimens of base and w eld metal. Correlation dependencies were used with standard tests to d etermine DBTTs for the base and weld metal(in accordance with Russian standards): base metal before annealing 40 degrees C, after annealing 16 degrees C; weld metal before annealing 212 degrees C, after anneali ng 70 degrees C. The estimated value of T-ko, for the initial, unirrad iated weld metal, was 50 degrees C. The experimental results were comp ared with a prediction of the extent of radiation-induced embrittlemen t of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel. The p redicted increase in the DBTT of the weld metal as a result of irradia tion (about 165 degrees C) is practically equal to the experimental re sult (162 degrees C). However, the value of T-f obtained from tests be fore annealing (212 degrees C) is about 40 degrees C higher than the e stimated value, i.e. the calculation does not produce a conservative e stimate. This was explained by a low estimate of T-ko (10 degrees C), which had been calculated using data from chemical analysis of the wel d metal, performed by the manufacturer. The investigations on the samp les, however, yielded an estimated value of T-ko = 50 degrees C. The e ffectiveness of annealing in restoring the mechanical properties of ir radiated VVER-440 reactor pressure vessels was confirmed. Recovery ann ealing lowered the DBTT of the weld metal by 85% or more of its radiat ion-induced shift.