A. Kryukov

The role of chemical composition in WWER RPV irradiation embrittlement

The main chemical elements influencing WWER RPV steel radiation embrittlement are copper, phosphorus, nickel and manganese. Copper and phosphorus are the main influencing elements for WWER-440 RPVs. The influence of copper is implemented in the forming of nanocluster centres rounded by other elements which are the effective barriers to dislocation movements. Phosphorus segregates inside the grains, interacting with matrix defects and attracted to the Cu-type precipitates. Phosphorus also migrates to grain boundaries through diffusion processes. Due to the higher nickel content and very low copper and phosphorus present in WWER-1000 RPV steel, the main influencing element for this RPV type is nickel. The manganese content also has to be taken into account due to synergism of Ni and Mn. The semi-mechanistic model is based on key embrittlement mechanisms of chemical elements influence: matrix damage, irradiation-induced precipitation and element segregation proposed.

International Database on Reactor Pressure Vessel Materials for Long-Term Operation of NPPs

This paper presents the capabilities of the International Database on Reactor Pressure Vessel (RPV) Materials (the Database) for precise and comprehensive RPV lifetime assessment, aimed at supporting the long-term operation of nuclear power plants. The Database was created in the framework of the International Atomic Energy Agency activities. Fourteen countries, including the United States, France, and Russia, supplied large amounts of surveillance results and data from national and international research programs. The recent achievements and open issues in the area of RPV radiation embrittlement assessment are considered. They concern mainly the effects resulting from long irradiation times and high neutron fluences (neutron flux influence, late blooming phases), nickel and manganese synergism, and further validation of appropriate safety concepts (the Master Curve approach). New information from ongoing surveillance and research programs has to be incorporated into the Database for the most effective RPV radiation embrittlement prediction. These additional data will greatly support the development of embrittlement correlations and embrittlement trend curves valid for long irradiation times.

Use of a Semi-Mechanistic Analytical Model to Analyze Radiation Embrittlement of Model Alloys: Cu and P Effects

We performed analysis of the basic mechanism of radiation embrittlement of steels and weld seams with account of the direct matrix fracture, precipitation and segregation of chemical elements. A model is proposed, which describes matrix fracture due to its neutron bombardment and for 11 model alloys provides an accurate description of the processes of primary embrittlement and re-embrittlement after scheduled heat treatment (annealing). The distinctive features of the proposed model are: a possibility of explaining the embrittlement processes before and after annealing for alloys with low (or zero) Ni content, and its applicability to the analysis of operational behavior of WWER materials.

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

Neutron Embrittlement of VVER 1000 and 440/213 RPVs: Learning From EC Projects on RPV Integrity

Since 1991 the European Commission has financed a significant number of Technical Assistance Projects to the Commonwealth of Independent States (TACIS) and EURATOM R&D actions addressing the main safety issues on RPV material embrittlement and integrity assessment. Since the VVER 440 reactors of the reference series 213 are made from recognised neutron embrittlement resistant materials and include comprehensive surveillance programmes, a standard plant life management procedure can be applied to address long-term concerns, mostly aiming at reducing uncertainties in the assessment techniques. Therefore, the open issues (flux effect, upgrading of surveillance results, implementation of toughness measurements and relevant acceptance criteria, behaviour of the cladding) are quite universal. The efficiency of late annealing (at fast [E>0.5 MeV] neutron doses over 1020 n/cm2 ) and the re-embrittlement after annealing remain key issues for any final decision for their operational lifetime. The more recently developed VVER 1000 reactors have some well-known features arising from the original design and manufacturing process (high nickel content in the core weld, location of the surveillance specimens), which have to be carefully considered if appropriate mitigation measures are to be implemented during operation. A precise identification of the issues related to the surveillance programme has been achieved thanks to research on dosimetry evaluation, representativeness (temperature and flux effect) of the specimens and optimisation of the evaluation of their results. Nickle, just as copper and phosphorus, is now recognised as having a detrimental effect on neutron embrittlement in synergy with other elements (e.g. manganese). The analyses of available data for CrNiMn steels do not show a significant effect for fast [E>0.5 MeV] neutron doses below 7.1019 n/cm2 , but their consistency and relevance might be questionable. A way has already been pioneered which shows how valuable results can be obtained using the existing surveillance programmes specimens. A systematic application on the Russian & Ukrainian plants is now planned in order to get updated figures on design end of life (EOL) integrity assessment. This includes updated dosimetry assessment, multiple specimen testing (reconstitution, impact and static toughness tests) and advanced integrity analyses. An optimised database of representative surveillance results (up to the design end of life) is expected, which should provide a sound basis for further understanding and setting up of relevant prediction tools, considering at the same time any other specific R&D results. The global integrity assessment will also provide for preparing and implementing adequate mitigation measures in due time, if necessary. The paper will report about the knowledge on RPV embrittlement effects, providing evidence of recent contributions to solve shortcomings of the VVER 440/213 and 1000 units. The current state-of-the-art and the remaining open issues have been assessed recently by a group of international experts. The planned R&D activities and the detailed scope of the latest TACIS projects are described.Copyright