Ya. I. Shtrombakh

Assessment of relative contributions from different mechanisms to radiation embrittlement of reactor pressure vessel steels

Abstract Experimental data on radiation embrittlement in pressure vessel steels of both Russian and American grades, obtained by the authors and also taken from the literature, have been analyzed to assess the relative contributions from the following mechanisms: radiation-induced hardening, inter- and intragranular segregation of impurities at precipitate/matrix interfaces. It is demonstrated that radiation-induced intragranular segregation of impurities frequently provides a significant contribution to radiation embrittlement of pressure vessel steels.

Intergranular and intragranular phosphorus segregation in Russian pressure vessel steels due to neutron irradiation

Russian reactor pressure vessel steels have been studied in three conditions: initial, irradiated and annealed. It has been established that irradiation induces both intergranular as well as intragranular phosphorus segregation. Fractographic studies demonstrated that brittle intergranular and ductile intergranular fracture surfaces of Charpy specimens appear as a result of intergranular and intragranular segregation, respectively. Transmission electron microscope (TEM) studies have revealed radiation-induced precipitates on interface boundaries to which intragranular phosphorus segregation occurs. Auger electron spectroscopy (AES) has been applied to detect phosphorus enrichment of fracture surfaces in the regions of brittle and ductile intergranular fractures.

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.

Radiation embrittlement of low-alloy steels

Abstract Results of phosphorus, copper and nickel effect on radiation induced yield stress increase and ductile-to-brittle transition temperature (DBTT) shift are presented. The synergetic interaction between phosphorus and nickel is observed. The results of Russian VVER-440 and VVER-1000 surveillance programs and results of research programs on reactor pressure vessel (RPV) steel irradiation in surveillance channels of power reactors are discussed. The basic regularities of VVER-440 and VVER-1000 RPV steel are discussed. Trend curves for VVER-440 and VVER-1000 RPV steels are developed. The annealing effectiveness for VVER-440 and VVER-1000 RPV steel grades was compared. DBTT recovery of VVER-1000 RPV steels was found to be much lower than for VVER-440 RPV steels. Nickel was supposed to increase the post-irradiation residual DBTT shift of VVER-1000 type steels. Models for prediction of the post-irradiation residual DBTT shift of VVER-440 and VVER-1000 type steels were suggested.

Radiation damage of graphite and carbon-graphite materials

Abstract The results of the study of graphite and carbon-graphite materials obtained in the Russian Research Center “Kurchatov Institute” during a period of fifteen years are presented. The main structural effects and changes in pyrographites, nuclear (polycrystalline) graphites and carbon-graphite materials under irradiation are investigated. Irradiation temperatures and neutron fluences have been taken from wide ranges: ≈ 500–1200°C, and ≈ (0.01–2) × 10 26 n/m 2 respectively, ( E > 0.18 MeV). The following characteristics of materials are discussed: linear and volume changes in size, lattice parameters, radiation defect parameters, pore structure evolution, Youngs modulus change, etc. Particular attention is given to mechanisms, governing material radiation stability as a function of processing and irradiation conditions. The widely known effect of radiation dimensional change in irradiated graphite has been observed in the experiment. It has been demonstrated that the radiation defect parameters and magnitudes of radiation dimensional change are defined by initial sizes of graphite crystallites (for irradiation temperatures ≥ 500°C). It has been shown also that this effect determines the radiation stability in polycrystalline nuclear graphites. Distinctions in behavior of carbon-graphite materials under irradiation are shown.

The principal structural changes proceeding in Russian pressure vessel steels as a result of neutron irradiation, recovery annealing and re-irradiation

Abstract A wide range of pressure vessel steels – in the initial state (i.e. unirradiated), after irradiation, recovery annealing and re-irradiation – have been studied using microstructural and fractographic methods. The analysis of the data has allowed quality explanations of the key features of radiation embrittlement (RE) in reactor pressure vessel steels (RPVS) and resulting from initial- and re-irradiation to be proposed, and also to justify the existing concepts concerning the mechanisms of RE.

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Abstract Within the framework of the 6 month WANO program, small samples were cut 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 characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state ( T ko ) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance 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 with calculated data were the principle objectives of the study. Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C. The estimated value of T ko , for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement 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 predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T ko = 50 °C. The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.

Radiation embrittlement kinetics of the first generation of VVER-440 RVPs after post-irradiation annealing

Abstract A new approach for estimation of re-irradiation kinetics of VVER-440 reactor pressure vessel steels is presented. Effect of gamma flux on re-irradiation kinetics is discussed. Comparision of predicted and observed values of re-irradiation ductile-to-brittle transition termperature shifts for VVER-440 reactor pressure vessel core welds is provided.

Assessment of irradiation response of WWER-440 welds using samples taken from Novovoronezh unit 3 and 4 reactor pressure vessels

Abstract The results of the study on Novovoronezh unit 3 and 4 (NV NPP-3 and 4) reactor pressure vessel (RPV) radiation embrittlement measured using subsize impact specimens (5×5×27.5 mm 3 ) fabricated from samples taken from the corresponding RPV walls are presented. The post-irradiation annealing effectiveness and the embrittlement kinetics of Novovoronezh unit 3 and 4 RPV welds under re-irradiation are discussed. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy (TT 10×10 ) and subsize impact (TT 5×5 ) specimens of trepans cut out from Novovoronezh unit 2 RPV are compared. A new relation between TT 10×10 and TT 5×5 has been developed.

Assessment of Irradiation Conditions in WWER-440 (213) RPV Surveillance Location

The key component of WWER is the Reactor Pressure Vessel (RPV). The evaluation and prognosis of RPV material embrittlement and the allowable period of its safe operation are performed on the basis of impact test results of irradiated surveillance specimens (SS). The main problem is that the SS irradiation conditions (temperature of SS, neutron flux and neutron spectrum) have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV condition. In particular, the key issue is the possible difference between the irradiation temperature of the SS and the actual RPV temperature. It is recognized that the direct measurement of temperature by thermocouples during reactor operation is the only way for receiving reliable information. In addition, the neutron field’s parameters for surveillance specimens have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level. The COBRA project (http://ie.jrc.cec.eu.int/ames/), which started in August 2000 and had a duration of three years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contain state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during the reactor operation. The selected reactor was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of the capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5±4°C. The results obtained show that there is not need in temperature correction when data of surveillance specimens studies are used for assessment of WWER-440/213 reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals ∼2.7·1012 cm−2 s−1 with E>0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.Copyright