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Safe Operation of VVER Reactor

#0797.2


Methodical Development for WWER-1000 Reactor Pressure Vessel Safe Operation Lifetime Evaluation Allowing for Anticorrosive Cladding

Tech Area / Field

  • FIR-MAT/Materials and Materials Conversion/Fission Reactors

Status
8 Project completed

Registration date
12.03.2003

Completion date
11.02.2008

Senior Project Manager
Alexandrov K A

Leading Institute
Engineering Center of Nuclear Equipment Strength, Reliability and Lifetime, Russia, Moscow

Supporting institutes

  • OKB Gidropress, Russia, Moscow reg., Podolsk\nNIKIMT (Institute of Assembly Technology), Russia, Moscow\nVNIIEF, Russia, N. Novgorod reg., Sarov\nTsNIIKM PROMETEY (Construction Materials), Russia, St Petersburg

Collaborators

  • National Nuclear Corporation, UK, Warrington\nTecnatom, s.a., Spain, Madrid\nEDF / Centre d'Ingénierie Déconstruction et Environment, France, Villeurbanne\nFramatome ANP GmbH, Germany, Erlangen

Project summary

The main purpose of this project is to develop methods for the determination of the residual VVER-440 and VVER-1000 reactor pressure vessels lifetime by fracture mechanics criteria.

In performing the project the following problems will be decided:


- generalization and analysis of methodical developments and experimental results obtained in the reactor materials science for last 10-15 years in Russia;
- harmonization of normative approaches used in Russian nuclear power engineering and those in Western practice (ASME, KTA and RCC-M Codes);
- perfection of breaking resistance calculation methods due to justified size diminution of the postulated design defect, use real brittle strength data for specific reactor pressure vessels (RPV), consideration of crack arrest and anticorrosive cladding as a constructive element;
- perfection of methods for RPV irradiation embrittlement evaluation taking into consideration the thermal annealing and actual flux;
- development of test program and technical requirements to improve the experimental basic for cracking resistance researches in irradiated reactor materials and establishment of correlation dependence between tests of surveillance specimens and large-dimension full-scale cross-section specimens cut from the spent lifetime Novo-Voronezh nuclear power plant VVER-440 reactor pressure vessel of the second unit.

The aims of the project proposed meet one of the basic ISTC objectives, namely: to support applied researches and technical developments in nuclear reactor power production and safety.

The project realization will allow specifying the currently operated VVER-440 and VVER-1000 reactor pressure vessels lifetime without reducing their safe operation due to more reliable and fundamental evaluation of RPV failure risk. In addition, the obtained results and developed approaches, methods and techniques can be applied by Western companies for evaluating safe operation and lifetime of their own PWR-type reactor pressure vessels including those after the thermal annealing.

The qualified scientists and engineers possessing extensive knowledge and practice experience in such areas as equipment development for special purpose nuclear power transportation facilities; design, production and operation of VVER reactor pressure vessels; development and production of nuclear ammunition will take part in this work.


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