nikkkom said:
It was a serious event indeed. I have a few questions.
Why reactor head wasn't examined every refueling - the head is removed and reinstalled anyway, it should be relatively easy to inspect it?
Why (it seems) only the inner surface of the head is clad in stainless steel? Can't the entire head be made from a better steel?
If it really is not practical, why at least the internal surfaces of all penetrations aren't clad in stainless steel, making this scenario impossible?
There are ASME code requirements per 10CFR50.55 do to inspection of a percentage of vessel components on a regular basis. These tests look for general trends, since flaws tend to develop slowly and across the entire vessel. Additionally anytime the RCS is taken apart (reactor coolant system), prior to restart, pressure tests have to be done to look for leaks. This leak probably was not a through-wall leak yet during the previous cycle.With regards to this penetration and weld issue, there was a lot of evidence that there were problems, for over 12 years, and it was mostly ignored. The biggest piece was when they had to change out their containment air filters every few days rather than every few months (as this was when the leak reached its worst point, and the discoloring and clogging of the filters was metal from the vessel accumulating).
With regards to other people's questions about LOCA (loss of coolant accident), this SB LOCA would have been well enveloped by the intermediate break or interfacing systems break LOCA, which are far worse accidents. In the event this leak actually punched through, the leak likely would not have been so bad as to cause a draindown of reactor coolant. The normal CV charging pumps (chemical volume control pumps) would likely have been able to maintain reactor coolant inventory. The operators would have recognized a high unidentified leak rate in the containment though, based on the containment sump accumulating water combined with increased charging flow without an increased letdown flow. Station technical specifications (not sure what the PWR tech spec is, in BWRs with standard technical specifications its in section 3.6) would direct them to have the plant cooled down within a 36 hour timeframe, and the leak would be found by inspection.
If for some unknown reason, the leak was large enough to actually cause an increase in drywell pressure or a reduction in reactor coolant system inventory, a LOCA trip would have shut the reactor down and the ECCS system would have injected to ensure core inventory was maintained. If the leak was large enough to cause this situation, the reactor would depressurize through the leak slowly, and the reactor would be brought to a hot shutdown condition (mode 4), then decay heat removal would bring it to cold shutdown (mode 5), with no core/fuel damage. The big issue in this case, is the reactor would have been blown down faster than 100 degrees F per hour, which is a safety limit on reactor vessels, and extensive analysis would have to be performed prior to restart to ensure the vessel did not embrittle beyond safety margins. There's always a small possibility that the vessel may never be used again.
In all cases, the plant would have to be shut down and the management and operational causes for the issue corrected prior to restart, as we have already seen at Davis Besse. Leaks in the ASME Class 1 reactor coolant system pressure boundary are not allowed.
What really got Davis Besse in trouble about this, was that they were told by the NRC to shut down to do reactor head inspections, based on indications other plants have seen of cracking. They said their head inspection program had no indications and that they sampled (most/all) of the head penetrations recently and none had indications. They asked for permission to wait a months or two until their next shutdown, and the NRC agreed based on their statements. This was a false statement, and constituted deliberate and willful misconduct, since the NRC made a safety decision based on apparent false information. Several people were indicted, and at least one is banned from US nuclear activities (and is currently having a lot of trouble getting an engineering job ANYWHERE).
Ultimately in all cases, the impact to the public would have been negligible under design basis conditions.
I suggest those who are interested go online and read some more. Everyone in the US nuclear industry gets trained on this every 2-3 years (along with TMI, Chernobyl, and soon Fukushima). There is a LOT of info available publicly.
