Environmental Engineering Reference
In-Depth Information
fl ux between specimens and containers, lack of neutron monitors in most of
containers and no suitable temperature monitors (Brumovsky and Zdarek,
2005). The location of surveillance specimens does not assure similar condi-
tions as the beltline region of reactor pressure vessels. A modifi ed surveil-
lance programme for VVER-1000/V-320С type reactors was designed and
implemented at the Temelin NPP in the Czech Republic. The technical fea-
tures of the surveillance test assemblies provide opportunities for implemen-
tation of an integrated surveillance programme, using samples from several
VVER-1000 units: Temelin 1 and 2 (Czech Republic); Belene (Bulgaria);
Rivne 3 and 4, Khmelnitsky 2 and Zaporozhie 6 (Ukraine); and Kalinin 3
(Russia). Irradiation of these archive materials together with the a refer-
ence steel JRQ (of ASTM A 533-B type) and reference steel VVER-1000
allowed a comparison of the irradiation embrittlement of these materials,
and an opportunity to obtain more reliable and objective results, as no reli-
able predictive formulae exist up to now because of a higher nickel content
in the welds. Irradiation of specimens from the cladding region will help in
the evaluation of resistance of pressure vessels against PTS regimes.
Several mitigation measures have been identifi ed for the VVER-1000 RPVs.
Based on the fracture mechanics analysis, heating up the hydro-accumulator
water to 55°C was recommended to prevent injection of ECCS water with
temperatures below 20°C for all the plants. The use of low neutron leak-
age core loading patterns in VVER-1000 reactors would reduce RPV wall
fl uences by approximately 30%. For reducing the neutron fl ux on the reac-
tor vessel, low leakage core design was introduced at some plants (i.e. fuel
assemblies with high burn-up to be placed at the core periphery). In addition,
the quality of manufacturing and alloy composition ensure the possibility of
LTO for VVER-1000 reactors (Vasiliev & Kopiev, 2007).
The steam generators for VVER-1000 have been designed on the same
principles as the VVER-440 plants, however the SGs at VVER-1000 plants
are replaceable. At some units, throughout the design service life of the SG,
there were problems resulting in necessary SG replacement. At the same
time, the SGs at some plants could be operated beyond design service life.
As operating experience has shown, it is the water chemistry of the second-
ary circuit that is the main factor infl uencing operability of the SG tubing,
as in the case of VVER-440 plants. Tube integrity is inspected by the eddy
current method; the results of the testing can be used to determine the plug-
ging criterion for defected tubes. Proper defi nition of the plugging criterion
is an important challenge.
The ageing problems of the SGs at VVER-1000 plants are as follows
(Trunov
et al
., 2006a):
cracking at headers of the cold collectors of the heat-exchange tubes
degradation of the welded zone at hot collector headers
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