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Nuclear Safety Cooperation

R2.09/94 INTEGRITY ASS. VVER 1000 RPV's

Status
Closed
Russia
Benefitting Zone
Eastern Europe / North Asia
€ 481,451.87
EU Contribution
Contracted in 1995
TACIS
Programme
Technical Assistance to the Commonwealth of Independent States

Details

Type of activity

Design safety

Nature

Services

Method of Procurement

(FR2007) Restricted Call for Tender - External Actions

Duration

17/10/1995 - 17/05/1997

Contractor

TEKNOLOGIAN TUTKIMUSKESKUS VTT

Project / Budget year

WW9406 Nuclear Safety 1994 / 1994

Objectives

The objective of the present project was to perform a limited integrity assessment of the VVER 1000 Reactor Pressure Vessel (RPV) taking into account the irradiation embrittlement effect.

The project covered four major areas:

Collection of the material data on Russian VVER 1000 RPVs.
Evaluation of baseline material properties.
Evaluation of the surveillance material data.
Limited RPV integrity assessment for one unit.
In addition, some experimental results from test reactor programmes have been analysed to complete the limited surveillance data.

Results

The surveillance test results of the 3 Russian VVER 1000 units, i.e. Balakovo 1, Kalinin 1 and Novovoronezh 5, show that the embrittlement rates of the high-Ni welds (No. 4) of the RPVs could be higher than those specified by the Russian Guide, but supplementary material data for higher neutron doses is needed for assessing the end-of-life material properties.
Besides the low specimen neutron doses, uncertainties associated with the irradiation conditions (neutron dose determination, irradiation temperature) as well as the small number of test results reduce the applicability of the present surveillance test results.

As a general conclusion from the material data and the results of the PTS assessment, showing no or a small margin in respect to the predicted fracture toughness, mitigation measures may be needed for most of the Russian VVER 1000 RPVs against the embrittlement of the beltline weld (No. 4).
According to the preliminary evaluation performed, heating the ECCS water from 20?C to 55?C could be a sufficient measure.
More relevant test data, at least on the high-nickel welds, as well as plant specific integrity assessments are necessary for addressing the need of further mitigation measures against irradiation embrittlement.