ML16201A445
| ML16201A445 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 06/16/2016 |
| From: | NRC Region 1 |
| To: | |
| References | |
| Download: ML16201A445 (7) | |
Text
1 IPRenewal NPEmails From:
Dentel, Glenn Sent:
Thursday, June 16, 2016 4:32 PM To:
DR WILLIAM CORCORAN
Subject:
RE: Re: UCS bolt assessment of operability Dr. Corcoran, I am responding to your email dated June 8, 2016 (ML16168A073). In that email, you raised several questions regarding baffle-former bolting conditions at Indian Point.
The NRC has published a webpage dedicated to baffle-former bolts at nuclear power plants, which provides information on this subject including a background on the issue, safety implications, and the NRCâ's next steps. For many answers to your questions, please visit our page at http://www.nrc.gov/reactors/operating/ops-experience/baffle-former-bolts.html.
Baffle bolt degradation and failures were first identified in foreign reactors as early as 1988. In response, the NRC issued generic communication to owners of U.S. plants when this condition was first discovered and ensured it was addressed in plants in the U.S. as part of our license renewal reviews. Baffle bolt degradation has also been identified in the United States at D.C. Cook, H.B. Robinson, Point Beach and Prairie Island, among others, but to a more limited extent than that detected at Indian Point Unit 2. The owners of Indian Point Unit 2 identified the baffle bolt problems by completing the required examinations as a result of our license renewal reviews for these plants.
When Indian Point Unit 2 observed visual indications of bolting problems during its recent refueling outage, the on-site NRC resident inspectors were made aware of the problem and began to follow the issue closely.
Inspectors trained in materials science issues from the NRC Region I Office and NRC headquarters staff completed on-site inspections to observe bolt replacement activities, critically reviewed the licenseeâ's technical evaluations regarding bolt replacement controls and supporting analyses, and discussed in detail the results of our reviews. Our inspectors also reviewed the licenseeâ's evaluation of this information as it pertains to Indian Point Unit 3. While our reviews and inspections continue in this regard, we do not have any immediate safety concerns.
Our inspection results, which will include the NRCâ's assessment of Indian Point Units 2 and 3, will be provided in an upcoming publicly available inspection report by August 15, 2016. I will reach out to you once the report for Indian Point becomes available and confirm that all of your questions have been answered.
Sincerely, Glenn Dentel, Chief Division of Reactor Projects Branch 2 From: DR WILLIAM CORCORAN [1]
Sent: Wednesday, June 08, 2016 7:36 PM To: Paul ; Lew, David ; Dentel, Glenn ; Dorman, Dan ; Haagensen, Brian ; Trapp, James Cc: sgalef@aol.com; Paul Gallay ; Amy Rosmarin ; Susan Van Dolsen ; Andrew Bast ; Roger Witherspoon ;
marilyn elie ; Aaron Cantu ; Sam Thielman ; Nancy Cutler ; Sheehan, Neil ; Pickett, Douglas ; Burritt, Arthur ;
2 Dacimo ; Jk Canepa ; Ellen Weininger ; Ellen Cantarow ; Alison Rose Levy ; Sandra Galef ; Ellen Cantarow ;
John Sullivan ; Kathryn Miles ; Elizabeth Browning Studio ; Elijah Shaheen ; vaughey@aol.com
Subject:
[External_Sender] Re: UCS bolt assessment of operability How would the reactor have responded to any and all design basis events, including loss of coolant accidents (LOCAs) and Steam Line Breaks (SLBs), if one had occurred just before the plant shutdown for the core former inspection?
How will the NRC develop lines of inquiry?
How will Entergy develop lines of inquiry?
What are the failure modes? Did all of the failed bolts fail the same way?
What are the results of the metallographic examinations and tests? Were they all the same?
What is the time distribution of the failures?
What are the conditions, behaviors, actions, and inactions that resulted in the failures not being manifested before 2016?
What was the technical safety basis for the inspection intervals?
Were there any alarms from the Metal Impact Monitoring System (MIMS)?
Were there any abnormal conditions that could be seen on any instrumentation and/or recordings?
What are the results of a careful inventory accounting of all of the bolts, bolt pieces, bolt fragments, fasteners, etc. to nail down exactly what foreign material is going to be still inside the reactor coolant system?
(Even temporarily trapped material can be freed by mechanical, hydraulic, or other perturbations/ shocks to the system, in which case the foreign material can damage important equipment including steam generator tubes and fuel cladding.)
Did the Metal Impact Monitoring System (MIMS) detect any of the pieces from broken bolts?
If so, how did workers evaluate the quite valuable advice it provided?
If not, should it have detected the impacts from the metal pieces?
What are the other U.S. reactors that have bolts and similar fasteners inside the reactor coolant pressure boundary?
What were the circumstances of the âœover 8700â bolts Areva said it has replaced? How many other bolts have been replaced? What were the circumstances?
What are the other questions and lines of inquiry?
The Areva infomercial is at https://youtu.be/fR-lVJUcbtY It says that Areva has replaced over 8700 bolts.
3 The NRC Information Notice 98-11 is at NRC: Information Notice No. 98-11: Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants NRC: Information Notice No. 98-11:
Cracking of Reactor V...
The Nuclear Regulatory Commission, prot ecting people and the environment.
View on www.nrc.gov Preview by Yahoo Take care, Bill Corcoran William R. Corcoran, Ph.D., P.E.
21 Broadleaf Circle Windsor, CT 06095-1634 860-285-8779 William.R.Corcoran@1959.USNA.com http://www.linkedin.com/in/williamcorcoranphdpe https://www.box.com/shared/kfxg1lt9dh On Wednesday, June 8, 2016 12:24 PM, Paul wrote:
From: Dave Lochbaum <DLochbaum@ucsusa.org>
Date: June 8, 2016 at 11:32:41 AM EDT To: "Blanch, Paul" <pmblanch@comcast.net>
Subject:
Indian Point Unit 3 Hello Paul:
I've had a chance to read Entergy's licensee event report on the degraded baffle bolts on Unit 2 as well as NRC Inspection Manual Chapter 0326 (attached and hereafter referred to as MC 0326).
Entergy described the LER as covering "Unanalyzed Condition due to Degraded Reactor Baffle-Former Bolts" (cover page).
4 Entergy stated that the degraded baffle bolts did not meet the acceptable criteria in WCAP-18048-P to allow Unit 2 to restart - "The consequence of this is that baffle-former bolt replacements were required to be completed prior to returning IP2 back to service" (page 2, middle paragraphs).
Entergy stated that "The root cause of the baffle-former bolt failures is primarily Irradiation Assisted Stress Corrosion Cracking..." (page 3) and that "Failure of a critical number of bolts in a localized area subsequently increased loading on adjacent bolts thus increasing the probability of failure of the adjacent bolts and generating the clustered pattern seen in the inspection results" (page 3).
Entergy committed to "Perform inspection of the baffle-former bolts in refueling outage 2R23" (page 4). MRP-227-A only requires re-inspections every 10 years (every fifth refueling outage), so Entergy is inspecting the Unit 2 bolts on an accelerated schedule due to the extent of the degradation.
MC 0326 Section 04.05 states that "Circumstances that require an operability determination: a. Degraded conditions. b. Nonconforming conditions. c. Discovery of an unanalyzed condition." As noted above, Entergy's LER reported that an unanalyzed condition had been discovered; namely, the degraded condition of the baffle bolts which was nonconforming to the acceptance criteria in WCAP-18048-P. Any one factor would require an Operability Determination -- having all three present just makes the right call easier.
MC 0326 Section 04.04 defines the scope of operability determinations. That scope should include "The SSCs (systems, structures, and components) affected by the degraded or nonconforming condition" and "The extent of condition for all similarly affected SSCs." Based on past practice, these criteria mean that, for example, a degraded condition found on a 2-inch valve must be assumed to affect 1-inch, 3-inch, and other sized valves if the cause of the degradation is unrelated to the 2-inch size. Similarly then, the degraded condition on Unit 2 SSCs should invoke an Operability Determination evaluating this potential degraded condition on Unit 3's SSCs.
MC 0326 Section 04.03 outlines the Presumption of Operability. It's akin to the presumption of innocence in trials. The first full paragraph on page 7 explains the limitations on the Presumption of Operability - "...it would not be appropriate to presume operability based on the future results of an analysis where there is not a reasonable expectation that the system can perform its specified safety function during the interim."
Entergy's report indicated that Unit 2 could not be safely restarted until the degraded bolts were replaced. Until it is established that Unit 3's bolts are not degraded (either by an Operability Determination or by shut down to inspect the bolts), there's not a reasonable expectation that the core former's safety function can be performed.
Entergy's LER and NRC's MC 0326 provide me the documentation I needed to contend that Unit 3 should either be shut down for baffle bolt inspections or have its continued operation sans inspections justified by a formal Operability Determination reviewed by the NRC.
- Thanks, Dave Lochbaum
5 UCS
Hearing Identifier:
IndianPointUnits2and3NonPublic_EX Email Number:
7509 Mail Envelope Properties (f54c7d797a7e4e8a8e6af142b2b7116c)
Subject:
RE: Re: UCS bolt assessment of operability Sent Date:
6/16/2016 4:31:33 PM Received Date:
6/16/2016 4:31:34 PM From:
Dentel, Glenn Created By:
Glenn.Dentel@nrc.gov Recipients:
"DR WILLIAM CORCORAN" <William.R.Corcoran@1959.USNA.com>
Tracking Status: None Post Office:
R1PWMSMRS02.nrc.gov Files Size Date & Time MESSAGE 10228 6/16/2016 4:31:34 PM image001.jpg 948 Options Priority:
Standard Return Notification:
No Reply Requested:
No Sensitivity:
Normal Expiration Date:
Recipients Received: