Environmental Engineering Reference
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108% of reactor thermal power, the CUF is equal to 0.027 due to vibration
even if a pipe wall thinning of 50% is assumed.
Analysis of the corrosion of piping wall must question whether the
erosion-corrosion allowance applied in the design provides suffi cient mar-
gin for 50+10 years of operation. Only a few cases are expected where the
existing corrosion-erosion monitoring programme using COMSY software
will have to be extended.
In analysing for material property change of the steam generator tubes,
the main fi nding of the study is that the thermal ageing of 08H18N10T
material used for heat-exchange tubes is negligible at operating tempera-
tures ~290°C. Results of laboratory tests show that there is no change in
the fatigue crack propagation rate due to LTO at 288°C (NPO Hidropress,
2007). An operational time of 60 years is justifi ed in this respect.
Reactor pressure vessel and internals
For the justifi cation of operability of RPV and RPV internals for extended
operational lifetime, the following analyses have to be performed.
PTS analyses for RPV test the structural integrity against brittle frac-
ture (fast fracture) of the RPV; it is ensured if the factual ductile-brittle
transition temperature (DBTT) of its critical components is less than the
maximum allowable component-specifi c DBTT. The analysis is based on
the comparison of the static fracture toughness of the material and stress
intensity factor calculated from the given loading situation (Linear Elastic
Fracture Mechanics (LEFM) concept). The steps in the analysis are pre-
sented by Katona, Rátkai and Pammer (2011). The fi nal conclusion of the
analyses is that the RPVs at Paks NPP can be safely operated for at least
60 years. For the sake of completeness of the studies, some additional anal-
yses are still ongoing regarding PTS sequences initiated by internal fi res,
fl ooding and earthquakes under shutdown conditions. The neutron fl uences
also have to be modifi ed taking into account the new fuel design introduced
after power up-rate.
Analysis of fracture toughness of structures within the reactor pressure
vessel was undertaken. According to the preliminary results the irradiation-
assisted stress corrosion cracking and void swelling may be of interest. The
stud joints fi xing the polygon mantle to the core basket can be critical in
both ageing mechanisms. Measures may be identifi ed after visual inspec-
tion of the core basket and review of inspection procedure. The possibil-
ity of implementation of a non-destructive volumetric test method for the
bolts is also a consideration. With respect to void swelling, the possibility of
implementation of ultrasonic measurements as well as gamma heating and
a replacement programme are being investigated.
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