We continue the chapter by describing the major components within the reactor which are subject to corrosive damage.

1.2.2 Reactor pressure vessel (RPV)

Reactor vessel heads (RVH) can experience different types of corrosion. In 2002 boric acid crystal deposits and iron oxide were found to have flowed out from several openings in the lower service structure support skirt after removal of insulation from the Davis-Besse RVH, after an accumulated «16 effective full power years (EFPYs) of operation. A large corrosion cav­ity was found on the downhill side of the low-alloy steel RVH.1 Boric acid corrosion wastage occurred on the RPV head surface and lead to a total low-alloy steel loss of ~4.3 cm3. Boric acid corrosion was not the only mech­anism involved in the degradation: it was supposed that erosion-corrosion may have played a role in the initial cavity formation; galvanic corrosion between the low-alloy steel and the stainless steel occurred around the perimeter of the exposed cladding; and axial stress corrosion cracks were observed in five control rod drive mechanism (CRDM) nozzles adjacent to the J-groove weld.1 Both Alloy 600 and 182 weld metal failed by pri­mary water stress corrosion cracking (PWSCC). There was no conclusive evidence that hot cracking contributed to the J-groove weld cracking. The Davis-Besse event illustrates the severe consequence of in-service crack­ing of RVH-penetration components fabricated from Ni-base Alloy 600 and 182 weld metal in which PWR water leaking from the cracked nozzles severely corroded the RPV head low-alloy steel material down to the 308 stainless steel cladding material.2

In the vessel, internals are exposed to irradiation. Under neutron flux the microstructure of the material can evolve: segregation at the grain boundaries associated with dechromization and hardening induced by the recombination of point defects. The first cracked baffle-former bolts were observed in 1988 in Bugey Unit 2 (PWR, France), during ultrasonic testing (UT) controls. Several bolts were examined3-6 and the failure was attributed to a particular case of SCC: irradiation assisted stress corrosion cracking (IASCC). Periodic inspections and a replacement program were set up in the affected reactor types. The assessment of the damage affecting the bolts revealed that significant differences in cracking behaviors exist between the various reactors. For instance, taking into account the number of cracked bolts, Bugey Unit 2 (100 cracked bolts in 140 000 h) and Fessenheim Unit 2 (46 cracked bolts in 140 000 h) are the most affected reactors (the remaining reactors were mostly less than 30 cracked bolts). Additionally, their bolts were made from the same heat, suggesting the influence of initial composi­tion and microstructure.