Junk Plant Indian Point: IAEA Says Maintenance Philosophy is "Run To Failure"?

Updated 4/2

Many plants have reconfigured their core into a upflow design. Why didn't Indian do this???

"To date, baffle bolt cracking was observed only in the “down-flow” design."

***So everyone knew baffle-reformer bolts degradation was a direct threat to safety in the 1980s. They knew the mechanism was related to radiation exposure in 1980s. While Indian Point waited until a 2016 outage to preform their first inspection? Is this being safety conservative. They knew it was a difficult and expensive job, a high potential for extending a outage, so their priority system kept putting off replacing the bolts. 

It goes to the question how these plants will enter into end-of-life? Will a plant end its life as a grand old lady who served society's greater purpose or will they shutdown under a national scandal? Will the permanent shutdown scandals proliferate further depressing the USA's public Nuclear Power approval rate? With a low public approval rate, can you imagine the hay the politicians and their competitors can make with this as a election issue. 

Remember Indian Point is a special case. The giant state governor is out to shutdown the plant. They have been negatively in the news with problems for years. This makes the plant highly susceptible to magnifying and amplifying events at the plant. Will controversy ruin the rest of Entergy's nuclear fleet in the eyes of the NRC and Wall Street.

 

What will a cascade accident look like. When large percentages of our population through the media panic uncontrollable. When nobody can control outcome like TMI. Here comes massive re-regulations? 

 

This is going to have very expensive implications at other plants.     

IAEA: Assessment and Management of Ageing of Major Nuclear Power Plant Components Important to Safety: PWR Vessel Internals 

2007 Update

7.5. CRACKING OF THE BAFFLE BOLTS

7.5.1. General 

In 1980’s inspections indicated baffle bolts cracking. The bolts are made of 316 cold worked stainless steels. They failed by intergranular cracking. Normally, 316 steel is not prone to IGSCC in this water environment and all the bolts cracked were located in the second and third rows from the bottom, that is exactly the place corresponding to the highest neutron 48 irradiation. This demonstrates that the neutron irradiation is a significant feature for this cracking, even if the exact mechanism is unknown now. It is difficult to conclude whether the bolts cracked by IASCC, or due to irradiation embrittlement, or by other IGSCC phenomena. To date, baffle bolt cracking was observed only in the “down-flow” design.

This cracking is a concern and made necessary the development of ultrasonic methods for the non destructive examination of the bolts. Table 13 shows Baffle-Former Bolts inspection results in some IAEA Member States.

…Creep is a function of stress level, temperature and time at temperature. Fast neutron exposure enhances austenitic stainless steels to creep. Some creep/relaxation of baffle bolts has been observed during testing and replacement of baffle bolts in the USA, France, Japan, and Belgium.

…IAEA-TECDOC-1119 documented ageing assessment and management practices for PWR reactor vessel internals (RVIs) that were current at the time of its finalization in 1998. It concluded that while irradiation assisted stress corrosion cracking (IASCC) of PWR internals ad not been observed for structural components globally so far, it may be of concern with time. After the issuance of the TECDOC, an inspection of baffle bolts in the United States discovered cracking in two of the four plants inspected. As the baffle and former assembly provides an interface between the core and the core barrel region and is important to safety because it provides a high concentration of the reactor coolant flow in the core region, IASCC on the baffle-former bolts is of safety concern. In addition, swelling (void formation), which was not addressed in IAEA-TECDOC-1119, could become an issue in PWR internals because of low displacement rates and increased temperature due to gamma heating. Concern of fretting wear of control rod guide tubes has also been raised in Japan. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1119 should be updated by incorporating those new events and their countermeasures.