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OFFICIAL USE ONLY SENSITIVE INTERNAL INFORMATION 50.59 site hazards analysis report associated with the proposed Spectra Energy 42-inch diameter natural gas pipeline. Mr. Blanch asserted that Entergy's 50.59 analysis was inaccurate and incomplete resulting in violations of 10 CFR 50.59, "Changes, tests, and experiments, 10 CFR 50.9, "Completeness and Accuracy of Information," 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and possibly 10 CFR 50.5, "Deliberate Misconduct." The petition, which gathered significant local stakeholder and political interest, was rejected by an NRC Petition Review Board by letter dated September 9, 2015 (ADAMS ML15251A023).
OFFICIAL USE ONLY SENSITIVE INTERNAL INFORMATION 50.59 site hazards analysis report associated with the proposed Spectra Energy 42-inch diameter natural gas pipeline. Mr. Blanch asserted that Entergy's 50.59 analysis was inaccurate and incomplete resulting in violations of 10 CFR 50.59, "Changes, tests, and experiments, 10 CFR 50.9, "Completeness and Accuracy of Information," 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and possibly 10 CFR 50.5, "Deliberate Misconduct." The petition, which gathered significant local stakeholder and political interest, was rejected by an NRC Petition Review Board by letter dated September 9, 2015 (ADAMS ML15251A023).
By letter dated November 6, 2015, NRR issued a response to 39 questions that Paul Blanch had presented during his July 15, 2015, presentation before the PRB regarding both the proposed and existing gas pipelines at IPEC.]]
By letter dated November 6, 2015, NRR issued a response to 39 questions that Paul Blanch had presented during his July 15, 2015, presentation before the PRB regarding both the proposed and existing gas pipelines at IPEC.))
IP3 Reactor Vessel Head Seal Leakage/Repair Following the IPEC Unit 3 Spring 2015 3R18 refueling outage, leakage was identified past the inner o-ring reactor vessel head closure flange. In order to isolate the inner seal leakage, the leak off line was secured and the outer o-ring was placed in service allowing start-up to continue. On, July 14th, a high temperature leak off line alarmed and cleared intermittently. The high leak off line temperature alarm annunciated and did not clear on July 19th, indicating an outer o-ring seal failure.
IP3 Reactor Vessel Head Seal Leakage/Repair Following the IPEC Unit 3 Spring 2015 3R18 refueling outage, leakage was identified past the inner o-ring reactor vessel head closure flange. In order to isolate the inner seal leakage, the leak off line was secured and the outer o-ring was placed in service allowing start-up to continue. On, July 14th, a high temperature leak off line alarmed and cleared intermittently. The high leak off line temperature alarm annunciated and did not clear on July 19th, indicating an outer o-ring seal failure.
Regional staff in DRS/DRP, Resident Inspectors, Nuclear Reactor Regulation staff comprised of a Senior Materials Engineer and the Senior Project Manager developed questions regarding the degraded condition and corrective action taken by the Licensee. On July 22, the licensee implemented an Operational Decision Making Issue (ODMI ) indicating different set points and associated triggers regarding the monitoring of containment temperature and humidity; telltale temp; and reactor coolant drain tank level (RCDT). In order to reduce the pressure between the inner and outer seal, both seals were placed in service.
Regional staff in DRS/DRP, Resident Inspectors, Nuclear Reactor Regulation staff comprised of a Senior Materials Engineer and the Senior Project Manager developed questions regarding the degraded condition and corrective action taken by the Licensee. On July 22, the licensee implemented an Operational Decision Making Issue (ODMI ) indicating different set points and associated triggers regarding the monitoring of containment temperature and humidity; telltale temp; and reactor coolant drain tank level (RCDT). In order to reduce the pressure between the inner and outer seal, both seals were placed in service.
On September 15, 2015, Unit 3 was shut down for a planned maintenance outage to replace the reactor vessel 0-rings. Following maintenance activities, the reactor became critical on September 25, 2015, and returned to full power operation on September 26, 2015.
On September 15, 2015, Unit 3 was shut down for a planned maintenance outage to replace the reactor vessel 0-rings. Following maintenance activities, the reactor became critical on September 25, 2015, and returned to full power operation on September 26, 2015.
[[Gas Pipelines (Existing and Proposed)
((Gas Pipelines (Existing and Proposed)
Presently, there are 2 natural gas pipelines (26" and 30") that run through the Owner Controlled Area of IPEC, and have been the subject of great interest with various stakeholders over the past few years. Spectra Energy approached Entergy during the summer of 2013 about plans to expand their natural gas pipeline capacity across the Hudson River with a new 42-inch diameter pipeline. On February 28, 2014, Spectra filed an application with the Federal Energy Regulatory Commission for a certificate to build a new 42-inch natural gas pipeline along a southern route on Indian Point property.
Presently, there are 2 natural gas pipelines (26" and 30") that run through the Owner Controlled Area of IPEC, and have been the subject of great interest with various stakeholders over the past few years. Spectra Energy approached Entergy during the summer of 2013 about plans to expand their natural gas pipeline capacity across the Hudson River with a new 42-inch diameter pipeline. On February 28, 2014, Spectra filed an application with the Federal Energy Regulatory Commission for a certificate to build a new 42-inch natural gas pipeline along a southern route on Indian Point property.
In accordance with NRC requirements, Entergy performed a site hazards analysis to determine the impact of the new natural gas pipeline on the site. On August 21 , 2014, Entergy voluntarily submitted a 50.59 evaluation and blast analysis for NRC review. A Region I DRS security inspector and Headquarters expert on blast analysis performed an ROP baseline inspection (71111 .18 - Plant Modifications) of the 50.59 and blast analysis. The results of the inspection OFFICIAL USE ONLY        SENSITIVE PRE-DECISIONAi INFORMATION 20
In accordance with NRC requirements, Entergy performed a site hazards analysis to determine the impact of the new natural gas pipeline on the site. On August 21 , 2014, Entergy voluntarily submitted a 50.59 evaluation and blast analysis for NRC review. A Region I DRS security inspector and Headquarters expert on blast analysis performed an ROP baseline inspection (71111 .18 - Plant Modifications) of the 50.59 and blast analysis. The results of the inspection OFFICIAL USE ONLY        SENSITIVE PRE-DECISIONAi INFORMATION 20
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                                                                                                                                         .2~50
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ar.R "CS Leakage is 91 days from now, or 11/ 5/2015 0 000 0 "' "' "' "'
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ar.R "CS
ar.R "CS Leakage is 91 days from now, or 11/ 5/2015 0 000 0 "' "' "' "'
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Latest revision as of 02:29, 17 March 2020

NRC-2018-000251 - Resp 1 - Final, Agency Records Subject to the Request Are Enclosed. (Part 2 of 2)
ML18354A760
Person / Time
Issue date: 11/27/2018
From:
NRC/OCIO
To:
Shared Package
ML18354A758 List:
References
FOIA, NRC-2018-000251
Download: ML18354A760 (375)
v • d • e


Text

From: Haagensen, Brian Sent: Wednesday, May 31, 2017 2:12 PM To: Rich, Sarah Cc: Siwy, Andrew

Subject:

FW: Assessment of IP3 Boric Acid Leakage - Recap Attachments: MP- PROC-MP- MP 3792AA[r006.00].pdf Attached is the Millstone 3 procedure for RPV o-ring replacement.

From: McKown, Louis Sent: Wednesday, May 31, 2017 2:10 PM To: Haagensen, Brian <Bria n.Haagensen @nrc.gov>

Cc: Ambrosini, Josephine <Josephine.Ambrosini@nrc.gov>; Highley, Christopher <Christopher.Highley@nrc.gov>

Subject:

RE: Assessment of IP3 Boric Acid Leakage - Recap Good Day, So with all of the containment entries do they have a hatch or a set of fancy revolving doors in series (that way we can credit redundancy for safety).

By the way when you are searching for o-ring, use 'o--ring' with two dashes ... you know because it's Millstone.

If you have any further questions, comments, or concerns please do not hesitate to contact me at the information provided below.

Very Respectfully, Lou McKown, Resident Inspector, Millstone Power Station Division of Reactor Projects, Branch 2 Nuclear Regulatory Commission, Region I 860-447-3170/3179 (Desk)

Louis.McKown@NRC.gov From: Haagensen, Brian Sent: W ednesday, May 31, 2017 1:58 PM To: McKown, Louis <Louis.McKown@ nrc.gov>

Cc: Ambrosini, Josephine <Josephine .Ambrosini@nrc.gov>; Highley, Christopher <Christopher.Highley@nrc.gov>

Subje ct: FW: Assessment of IP3 Boric Acid Leakage - Recap Lou - I'll show you mine if you'll show me yours (i.e. send me a copy of the Millstone 3 RPV o-ring installation procedure).

From: Haagensen, Brian Sent: Wednesday, May 31, 2017 8:30 AM To: Setzer, Thomas <Thomas.Setzer@nrc.gov>

Cc: Tifft, Doug (Doug.Tifft@nrc.gov) <Doug.Tifft@nrc.gov>; Greives, Jonathan <Jonathan.Greives@nrc.gov>; Guzman,

Richard <Richard.Guzman@nrc.gov>

Subject:

Assessment of IP3 Boric Acid Leakage - Recap Yesterday, Entergy entered the IP3 containment and identified a 6-8 foot boric acid deposit on the reactor pressure vessel flange area. The BA apparently leaked from the outer RPV flange o-ring.

(b)(4)

Presently, the RCS UIL is no longer increasing. Review of past leak rate data indicates that the leak has stabilized at -0.12 gpm. This may be indicative of the outer o-ring sealing to the flange when full pressure was applied. The BA leakage may have been a transient phenomenon, where by the leakage occurred during the period of time when IP3 shifted from the inner o-ring seal to the outer o-ring seal. Current leakage may be going through the outer seal or it may be leaking from another location.

2

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The challenge is that while the o-ring seating surface is stainless steel and therefore is not subject to BA corrosion, the RPV studs are carbon steel and may be subject to BA corrosion.

Alternate leakage indications in the containment show that containment sump leakage initially increased at the same approximate time that the inner o-ring seal failed. However, sump chemistry results are not indicative of an RCS (sump pH not decreasing, [BA] not increasing) and the containment particulate radiation monitor is no longer increasing (increased from 1.0 E-9 on 5/25 to 1.6 E-9 uCi/ml on 5/26 and stabilized).

IP3 is still analyzing the conditions and trying to make sense out of conflicting data.

At this point, we should let them take the time to get this right without trying to get ahead of the issue. No public statements are planned in the immediate future regarding this leak. However, we should be ready to answer questions at the AAM regarding this condition.

Brian C. Haagensen Senior Resident Inspector Indian Point Energy Center 914 739-9360 (Office)

(b)(6) (Cell) 3

From: Haagensen, Brian <bhaag90@entergy.com >

Sent: Monday, December 11, 2017 10:03 AM To: Haagensen, Brian

Subject:

[External_Sender] FW: RCS keakage trends on Unit 2 Brian C. Haagensen Senior Resident Inspector Indian Point Energy Center 914-739-9360 (Office)

!(b)(6)  !(cell)

In plant x5347 From: Safouri, Christopher Sent: Tuesday, November 28, 2017 1:35 PM To: Haagensen, Brian

Subject:

RE: RCS keakage trends on Unit 2 Understood, I'll check it out.

Thanks, Chris From: Haagensen, Brian Sent: Tuesday, November 28, 2017 9:02 AM To: Safouri, Christopher Cc: andrew.siwy@nrc.gov

Subject:

RCS keakage trends on Unit 2 Chris, Take a look at the Unit 2 total leakage trend below. Since 11/16, gross RCS leakage has slowly increased from 0.08 gpm to 0.13 gpm. This appears to be caused by an increase in non-RCS pressure boundary leakage (i.e. eves leakage) from 0.01 gpm to 0.05 gpm . These are still very small numbers so no action is needed but t hey apparently are seeing leakage through the charging pump packing again.

Brian

(b)(4) 2

Brian C. Haagensen Senior Resident Inspector Indian Point Energy Center 914-739-9360 (Office) j(b)(6) I(cell)

In plant x5347 3

From: Haagensen Brian To: Schroeder Daniel; Setzer Thomas: Henrioo Mark* Rossi Matthew

~~~m~~~~~; R~~l~~kaCghei:~:~~:-0n Unit 2 -l(b )( 5)

Cc:

Subject:

Date: Monday. December 11 , 201711:16:20AM '----------'

Attachments : Extemal Sender FW RCS keakage ueods on Voit 2 msg unideorified leakage srats treads rnsg External Sender msg Dan ,

l(b)(5)

Two weeks ago (on 11 /28), we identified a small but steady increase in the gross (total) RCS leak rate (see 1st attached email

- from 0.08 to 0.13 gpm). IP2 attributed this trend to an increase in eves leakage which did not make a lot of sense to me (CVCs leakage tends to trend in steps, not slow increases, when shifting charging pumps). However, the absolute value was tor the leak rate was still very small, On 12/8, we noted that IP2 would be in AL 1 (see 2nd attached email) on Monday (today) but the small absolute value (0.05 apm UIL) seemed to be inconsequential CR-IP2-2017-05071 (see 3rd attached email) showed that there may have been indications of increased RCS leakage into containment following the MBFP outage last September based on Xe-133 / Ar-41 ratio trending, but nobody jdentjfjed it at the time

- - . . , . . - - , - - - - - - - - - "The most battling piece is that there was never any 1n 1cat1on o a ea on e e - a e rain line from the external o-ring. In fact, the leakoff line temperature actually dropped slowly from 94 F to 73 F over the since September. Th is may have been caused by the line being plugged with boric acid from the previous inner a-ring leak.

The overall gross RCS leakage trend from September to December is as follows (see graph below) :

(Note - the gross leakage spike on 9/ 18 was caused by the MBFP outage)

(b)(4)

IPEC has written a CR to determine how thei r ODMI for the inner o-ring leak failed to catch the outer o-ring leak.

Brian C. Haagensen Senior Resident Inspector Indian Point Energy Center 914 739-9360 (Office)

(b )(6) (Cell)

f rom:

To:

Sil~Kt:

HJ~t.~J'l'!e", llria>n (ht1tNl_,Send111 rn': 11.( S lukl1,1 111nll1 an un~ l Oite: 1,W!o.l'f,Dettfflbetlt, 201710:0l. 26AM An,11,hmtr1t$:  !!!!!L~

Sr/,11n C. HJilgtnstn Senior Re51dent lnspectOf Indian Pomt Energy Center 914 /39 9360 (OU1<c) ilh}/6) j(celll tn plJnt M!°>.3117 From: Safourl, Ch,lstopher Sent: Tuesday, Novembtr 28, 2017 1:35 PM To: Haagensen, Brian Subj ect: RE: RCS t:e.-ikage trends on Unit 2 Understood, I'll check it out From: Haagensen. Brian Sent: Tue$di:ly, Novembe( 28, 2017 9:02 AM To: Safoun, Christopher Cc: andWY s~vtnrc sov Subj ect: RCS keallage uends on Unit 2 Chr1s, Take a look .it the Unit 2 total leakage trend below. Since 11/16, gross RCS leakage has slowtv increased from 0.08 gpm to O 13 epm This appears to be caused by an ina ease in non-RCS pressure boundary leakage (1.e CVCS le11k11get from O01 gpm to 0.05 gpm. Thes,e 11re still very sm11II numbtrs so no ;,ctlon Is needed lxit thev app.erently are seemg le;,kage ch rough the ch11rging pt.Jmp p11cklng ag;,in.

Brian This is the same graph that appeared in the prior email, which was been redacted on the basis of FO IA exemption 4.

11 Srl;,n C. Hn gensen

From: Alyse L. Peterson <Alyse.Peterson@nyserda.ny.gov>

Sent: Thursday, September 03, 2015 3:39 PM To: Tifft. Doug Cc: bridget.frymire@dps.ny.gov; McNamara, Nancy

Subject:

[External_Sender] Re: IP3 seal leakage update Thank you very much Doug. Please continue to keep us updated.

On Sep 3, 2015, at 3:07 PM, Tifft, Doug <Doug.Tifft@nrc.gov> wrote:

SENSITIVE INFORMATION - This email contains proprietary information related to a plant shutdown - NOT FOR PUBLIC RELEASE As we discussed earlier this week, Indian Point Unit 3 is currently working *through an issue where both their inner and outer reactor vessel head seals a,re leaking. And as discussed, NRC inspectors have been closely monitoring the licensee's actions in response to this issue. At the end of the call you requested to be informed should the situation change, so I would like to update you on some recent developments.

  • At the time of.the call, the inner reactor vessel head seal was in service. We discussed how the leakage from the inner seal was directed to the reactor coolant drain tank (RCDT). We noted that at the current leak rate, the operators were pumping down the tank about twice per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> shift. Subsequent to our call, the licensee chose to place the outer seal in service. The thought being that the outer seal was less degraded than the inner seal and would result in less leakage and lessen the burden on operators of pumping the RCDT down. When the licensee placed the outer seal into service, they began to see containment parameters that indicated that the leakage was not flowing-through the drain lines to the RCDT, but rather directly to containment across the outer o-ring. The licensee again plac~d the inner seal back in service.

With this new information the licensee is re-evaluating their decision to remain operating. They are still working through their engineering decision making process, but the licensee has indicated that they may shut the plant down in about 1 week or so to r~pair the seals. (The date we have heard is Friday, September 11, but my understanding is this a soft targe, and is weather dependent.) The NRC continues to monitor the licensee's actions and decisions as they work through this issue.

Please let me know if you have any questions.

-Doug

~<<? ~

~:p;)(~]

Regional State Liaison Officer 0-337-6911 1

From: Te\son, Ross Sent: Monday, June 26, 201712:28 PM To: Greives, Jonathan Cc: Anderson, Shaun; San,ders, Serita; King, Mark; Rosebrook, Andrew

Subject:

RE: Indian Point RPV Head 0-Ring Leak Jon, Thanks for the additional information. Actually, I was engaged in an email discussion with another staff member in IOEB and 1asked him a number of questions (not realizing t.hat they would make their way to you). Apologies if we distracted you from other priorities. I just tried to teach you by phone but without success ..

I had seen email from the region indicating (a) a pattern of repetitive (RPV Head-to-Vessel) 0-ring failures at JP-3 and unsuccessful licensee correctiv~ actions (to prevent repetition?), and (b) regional intent to inspect associated licensee cause determinations and corrective actions to prevent repetition. This caused me to question (c) whether the licensee (or NRC) had considered the failures to constitute one or more SCAQs and

( d) how proposed changes in NEI 16-07 (CAP-02) might impact the process and expectations. I think I understand now that no one is considering the issue to constitute an SCAQ.

As you may know, I'm on a team reviewing NEl 16-07 which proposes a number of relaxations to licensee Pl&R programs including (a) adjustments to SCAQ thresholds and (b) Crit. XVI SCAQ requirement interpretations specific to cause determination and corrective actions to prevent repetition. Your NEI 16-07 representative is Andrew Rosebrook.

It was not my intent to task re ion al resources nor to insert myself into regional inspection. However, I can be reached at M: (b)(6) if you have any insights you wish to share or questions for me. I would be happy to share perspectives. (*We recently moved offices and my desk phone is not working correctly yet.)

1

~U.S.NRC

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Ross Tefson (301} 115 22S. Temporarily oos) M !(b)(6)

Reactor Inspection Branch (IR'm} .___ _ _____.

NRR Div. of Inspection & Regional Support Office - O 13E20 PS: If you're curious, you can find Draft NEI 16-07 [Rev. A] at ML17152A233.

See 2.1 Definitions (Pg. 4) for definition of Corrective Action to Preclude Repetition (CAPR).

See 3.2 Significant Condition Adverse to Quality (SCAQ) (Pg. 6) for direction on cause determination & CAPR See Appendix A - Condition Report Significance Examples (Pg. A-1) for definition & examples of SCAQ.

From: King, Mark Sent: Monday, June 26, 2017 10:45 AM

  • To: Telson, Ross <Ross.Telson@nrc.gov>

Subject:

FW: Indian Point RPV Head 0-Ring Leak Your welcome Ross...ta/k about your timely support from R-1.

From: Greives, Jonathan Sent: Monday, June 26, 201710:31 AM To: Te Ison, Ross <Ross.Telson@nrc.gov>

Cc: King, Mark <Mark.King@nrq~ov>;-setzer, Thomas <Thomas.Setzer@nrc.gov>; Haagensen, Brian

<Brian.Haagensen@nrc.gov>; Rich, Sarah <Sarah.Rich@nrc.gov>; Siwy, Andrew <Andrew.Siwv@nrc.gov>; Henrion, Mark

<Mark.Henrion@nrc.gov>

Subject:

Indian Point RPV Head 0-Ring Leak Ross, We were told this morning that you had a questions regarding the CAP classification of the double a-ring leak for IP3's RPV head. Entergy did not classify the most recent leak as a significant conditions adverse to quality. In all cases, _they classified the leaks to a level B, condition adverse to quality, requiring an apparent cause evaluation. This is consistent with how they have classified previous occurrences. The residents are in the middle of a PIR sample on repeated leaks from RPV head o-rings and we expect it to be documented in the 2017 integrated inspection report.

Let me know if you have any additional questions.

Regards, Jon Jonathan Greives Branch Chief (Acting)

Branch 2- Division of Reactor Projects U.S. Nuclear Regulatory Commission- Region 1 610-337-5120 (W)

!(b)(6)  !(C) jonathan.greives@nrc.gov 2

From: Guzman, Richard Sent: Monday, June 26, 2017 2:07 PM To; Greives, Jonathan Cc: Setzer, Thomas

Subject:

Fwd: 2.206 Petition from T. Gurdziel dated June 11,.201 7 re; Indian Point Unit 3 Jon - see message below. Thanks for the assist.

Rich Guzman Sr. Project Manager NRR/DORL US NRC 301-415-1030 Begin Forwarded Message:

From: "Guzman, Richard" <Richard.Guzman@nrc.gov>

Subject:

2.206 Petition from T. Gurdziel dated.June 11, 2017 re: Indian Point Unit 3 Date; 22 June 2017 16:25

  • To: "'tgurdziel@twcny.rr.com"' <tgurdziel@twcny.rr.com>

Mr. Gurdziel, Good Afternoon. As we discussed, I have been assigned as a Petition Manager for the 10 CFR 2.206 petition you submitted to the U.S. Nuclear Regulatory Commission (NRC) on June 11 , 2017, regarding your concerns with the reactor vessel head a-rings at Indian Point Nudear Generating Unit No. 3.

Section 2.206 of Title 10 of the Code of Federal Regulations describes the petition process - the primary mechanism for the public to request enforcement action by the NRC in a public process. This process pennits anyone to petition NRC to take enforcement-type action related to NRC licensees or licensed activities. Depending on the results of its evaluation, NRC could modify, suspend or revoke an NRG-issued license or take any other app1ropriate enforcement action to resolve a problem. The NRC staffs guidance for the disposition of 2.206 petition requests is in Management Directive 8.11, which is publicly available.

The 2.206 process provides a mechanism for any member of the public to request enforcement action against NRC licensees. The 2.206 process is separate from the allegations process which aff'.ords individuals who raise safety concerns a degree of protection of their identity. In the 2.206 process, all of the information in your letter will be made public, including your identity.

You specifically requested in your e-mail for the NRC to keep Indian Point, Unit 3 (IP3) in cold shutdown until the condition of the reactor vessel head upper and lower surfaces are "proved to be identical to the as-purchased condition". The NRC considers your request as a short-term, immediate action given that IP3 is in the process of restarting from its maintenance outage. On June 22, 2017, your r,equest for immediate action was reviewed by members of the Petition Review Board (PRB), which includes st aff from the NRC's Office of Nuclear Reactor Regulation (NRR) and Region 1.* After its review and discussion, the PRB determined that there were no immediate safety concerns which would adversely impact the public's health and safety; therefore, the PRB denied yo1 ur request for immediate action in the restart of IP3. Specifically, the PRB noted

- that plant technical specifications require the licensee to monitor for unidentified leakage into containment and specifies actions if leakage were increase in excess of these limits, up to and including a plant shut down. Unidentified leakage at IP3 is currently within the limits specified by this requirement.

1

In accordance with NRG Management Directive 8.11, you have the opportunity to address the PRB, either in person at the NRC Headquarters in Rockville, MD, or by telephone conference. The purpose of this interaction is so that the petitioner can discuss the petition and verbaHy supplement the petition with any new information. During the meeting, the PRB is in listening mode and will not make any decisions regarding your petition. I understand from our conversation today that you would like to decline the opportunity to address the PRB at this time.

If you have other questions on the 2.206 process, or regarding the role as petition rnanager, please contact me at 301-415-1030.

Thank you, Rich Guzman Sr. PM, Division Operator Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Office: 0-9C7 I Phone: (301) 415-1030 Richard.Guzman@nrc.gov 2

From: Alley, David Sent: Wednesday, August 12, 2015 9:44 AM To: Tsao, John; Burritt, Arthur; Floyd, Niklas; Gray, Mel Cc: Holmberg, Mel; Shaikh, Atif; Drake, James; Elliott, Robert

Subject:

RE: Indian Poi nt Unit 3: Questions Regarding the Unit 3 inner and outer Rx Head 0 -

rings Leakage

All, I am work at home today and leave beginning tomorrow.

Despite the info notice, these leakoff lines appear to still be a source of confusion. When we wrote the info notice, I believe it was our intent that plants use the info notice as a basis for the need to conduct pressure tests. It is certainly our opinion that this piping should be code class piping and that it should be pressure tested. Indian Point is not currently alone in this issue. I have recently received calls on the subject from Jim Drake (ANO I think, inner O ring is leaking, plant doesn't want to stop the leak by closing leakoff line valve because line has not been pressure tested), Atif Shaikh and Mel Holmberg (Perry I think no leaks but discrepancy discovered during inspection)

Part of all this appears to be related to the age of the plant. If the plant predates Reg Guide 1.26 (again I think that is the right one), there is a bigger chance that the leakoff line could be other than code class pipe. If the leakoff line is appropriately classed as other than code class pipe, we may have a harder time with enforcing the pressure test. If the leakoff line is classed by the plant (or should be classed) as code class piping, i.e., the plant is committed to the reg guide, it would seem that we have a valid basis for challenging the lack of a pressure test.

As John mentioned in his email, we get a lot of relief requests for reducing the pressure for testing these lines.

The universal approach is to test the line with the refueling cavity full. This puts the line under about 10 to 15 psi. Not much but something. When we first started authorizing these alternatives we agonized over them for a long time. Some of the lines (generally BWRs if I remember) have a thermally operated valve. When leakage occurs, temperature goes up the valve opens and the control room is notified. We had less concern with these because the pressure stayed low. We did not consider the concept that the valve would be closed and the plant would continue to run for a long time. The rest of the plants had a pressure sensor in the leakoff line.

When leakage occurred, pressure built up and as some point the valve opened and the control room was notified. The pressure at which the valve opened was significantly higher than the test pressure, ergo the greater concern about the test pressure.

Aside from all the above, the question at hand, at least to some extent, involves the question of where the pressure boundary is located. We may be in a situation where the ASME code pressure boundary and the reactor coolant pressure boundary (tech specs) are in different places. We are very good sometimes in using similar words to have different meanings and then spending hours discussing what we did or did not mean.

Bottom line on this subject is that it is on my list of things to discuss with Rob Elliot of the tech spec branch. It is not, however, on the "must be handled immediately list". (If it needs to be elevated in priority, please let me know.)

I don't know that this has helped any, but at least you can all take comfort in that you are not alone in dealing with this and that we at headquarters are paying attention to the problem. I am open to suggestions as to how to proceed and how fast that needs to happen.

I promise that any responses to this that I get today will get more attention than those that come in while I am on leave. I probably will check email some while on leave b~Jt vom prjmary contact till I get back will be John Tsao. Should you really need to get hold of me, please ca1q(b)(6) 1

._____ __,I

Dave From: Tsao, John Sent: Wednesday, August 12, 2015 9:10 AM To: Burritt, Arthur; Floyd, Niklas; Gray, Mel Cc: Alley, David

Subject:

Indian Point Unit 3: Questions Regarding the Unit 3 inner and outer Rx Head 0-rings Leakage I have a concern regarding the licensee's response to Question # 1 which asked whether the vessel 0-ring drain lines have been pressure tested at Indian Point Unit 3 (please see email below). The licensee's interpretation of the ASME Code,Section XI, IWC-5221 is that a pressure test is not needed because there is no pressure in the drain lines during normal operation.

IWC-5221 states that "The system leakage test shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability (e.g., to demonstrate system safety function or satisfy technical specification surveillance requirements)."

My interpretation (as discussed in Information Notice 2014-02) is that the system pressure of the drain lines should be the pressure of the RCS even though there is no pressure inside the drain lines during normal operation. The drain lines are designed to support the RCS pressure in case there is a leak from the reactor vessel via a degraded inner 0-ring. The purpose of the system leakage test as required by IWA-5000 and IWC-5000 is to ensure that the pipe does not leak when it is needed. The drain lines are needed when leakage occurs. When leakage occurs, the drain lines will be pressurized.

Therefore, my position is that the drain lines need to be pressure tested using RCS pressure during normal operation.

However, we have recognized that pressure testing of the drain lines using the RCS system pressure is a hardship. Therefore, a licensee may request relief from using the RCS system pressure. As an alternative, licensees may use the hydrostatic head of the cavity water as the test pressure for the system leakage test and perform VT-2 exam prior to startup during the refueling outage. We have reviewed and approved such relief requests from licensees.

I think that Indian Point should submit a relief request and should include the pressure test of vessel 0-ring drain lines in its 10-year inservice inspection program.

Because Indian Point has never performed a pressure test on the 0-ring drain lines I do not know whether the Region Office should cite the licensee for non-compliance to the ASME Code,Section XI (i.e., 10 CFR 50.55a).

Please note that the above are my personal thoughts and they have not been reviewed by my supervisor, Dave Alley. I have included my supervisor Dave Alley in this email. Any final NRR decision should come from Dave Alley. Dave is on annual leave from August 12, 2015 to August 19, 2015.

John From: Burritt, Arthur Sent: Tuesday, August 11, 2015 5:56 PM To: Tsao, John <John.Tsao@nrc.gov>; Floyd, Niklas <Niklas.Floyd@nrc.gov>; Gray, M el <Mel.Gray@nrc.gov>

Subject:

FW: Questions Regarding the Unit 3 inner and outer Rx Head 0-rings Leakage 2

For your review and thoughts on next steps From: Stewart, Scott Sent: Tuesday, August 11, 2015 2:50 PM To: Floyd, Niklas <Niklas.Floyd@nrc.gov>

Cc: Burritt, Arthur <Arthur.Burritt@nrc.gov>; Pinson, Brandon <Brandon.Pinson@nrc.gov>; Rich, Sarah

<Sarah.Rich@nrc.gov>; Newman, Garrett <Garrett.Newman@nrc.gov>; Pickett, Douglas <Douglas.Pickett@nrc.gov>

Subject:

Questions Regarding the Unit 3 inner and outer Rx Head 0 -rings Leakage The licensee has prepared this response to questions we had raised regarding the flange leak on Unit 3. FYI, Scott

1. Has the head vent leak-off line been pressure tested according to ASME Code,Section XI, IWA-5000? If not, why?

We believe this is referring to the Vessel 0-ring leak-off lines under NRC Information Notice 2014-02 instead of the head vent. In brief, IPEC does not pressure test these lines . Response to the referenced IN is documented under WT-WTIPC-2014-00055 CA-06. See excerpts of the response below:

Even though IPEC is susceptible to the issue discussed above, the intent of the requirements of 10 CFR 50.55a (g) (4) are being met. The lines have been classified as Quality Group B {Class 2) for both IP2 and IP3. An 0-ring is an acceptable barrier between safety classes per ANSI/ANS 51 .1, thus providing the boundary from Class 1 (reactor vessel) to Class 2 (leak-off lines). Per IWC-5221, Class 2 components are tested at the system pressure obtained while the system is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability. The lines are isolated (no pressure) during normal operation. Applying pressure to the line for testing is not required. During each refueling outage prior to startup, a visual inspection (VT-2) is performed on these lines at normal operating pressure and temperat ure; 2-PT-R075, "RCS Integrity Inspection" for IP2, and 3-PT-R131, "RCS Integrity Leak Test" for IP3. During normal plant operation when the VT-2 inspection is performed, the flange leak-off line is not pressurized provided the 0-rings are holding pressure. However, if the inner 0-ring is leaking, there will be some pressure in the line and the VT-2 inspection will be effective in detecting leakage if a through wall defect is present.

The requirement of monitoring and inspecting for leakage past the reactor head flange 0-rings is satisfied by having a leak collection system and verifying that the collection system is operative. This satisfies a VT-2 visual examination of the 0-ring per ASME Section XI, IWA-5243. The reactor vessel leak detection instruments are maintained by calibration; IP2 calibration procedure 2-IP-I-T-401 , IP3 calibration procedure IC-PC-I-T-401.

2. How much of the reactor vessel and closure head flanges are clad? Are there any detailed drawings, which show the extent of cladding on these components?

The exposed mating surfaces of the reactor vessel and closure head flange are cladded. Drawing 234-042 for the reactor vessel show a nominal clad thickness of 7/32". However, due to the proprietary nature of the closure head forming and welding under drawing 233-046-4 and 234-046-1, exact detail of the cladding cannot be determined.

The best estimate that is available is from drawing 234-047.

3. What is the sensitivity level for detecting the presence of Boron in the weekly VC atmosphere samples?

What is the sensitivity level for detecting RCS leakage via the R-11 and R-12 radiation monitors?

The existing gross comparative* methodology for checking Boron concentration in VC atmospheric samples is by using the filter media from the rad monitor samples, by "digesting a portion of the filter paper and analyzing it using the Ion Coupled Plasma instrument" for Boron. However this methodology is intended to detect/ track gross changes in atmospheric boron concentration over long periods. We have visually inspected the filters, but do not 3

currently have the instrument required; i.e., "the acid digester" for performing the Boron analysis. We have saved the filters.

As for sensitivity of R-11 and R-12, based on current RCS activity level and a review of calculations IP3-CALC-RAD-03514 and IP3-CALC-RAD 03414, it is anticipated that R-11 would be able to detect a 1 GPM leak into containment within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while R-12 would be able to detect a 3- 10 GPM into containment within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4. During the most recent VC visual inspections, how much area around the reactor head flange was available for inspection (i.e. full circumference or partial)? And, are there any interferences that could mask potential flange leaks? Is it possible to visually examine the reactor vessel for boric acid accumulation while at power?

The entire periphery of the vessel is able to be observed from the east and west points over the cavity, looking down. The actual flange to vessel interface and the closure studs/bolts are not visible with the unit at power and the mirror insulation applied . However, the insulation extends down from the flange to the cavity floor and there is a gap between the insulation and the floor. Additionally, there is a vertical gap between the vessel to the cavity floor such that liquid could drop, down the side of the vessel to the Reactor Cavity Sump. There have been no level alarms or pump-outs of the cavity sump since the RFO. Leakage past the flange to head interface would start as steam and then condense. To this point, as noted on the weekly inspections, there has been no evidence of external leakage past the flange interface as noted in the form of steam, pooling condensed steam, or dried boron extending along the cavity floor.

5. In response to ODMI limits, it was noted that both Robinson and Salem operated with both 0-rings leaking. Was there any evidence of flange leakage observed following these two events? And, was there any adverse impact as a result of operating in this configuration?

There was reportedly no damage to the Robinson 2 0-ring seat ing surfaces, but extensive clean-up was required for the effects of the Salem 1 0 -ring leak. There no report as to the extent of any indications on the Salem 1 0-ring seating surfaces that required repair. However, there is mention of boric acid and oxide deposits that had to be cleaned during the recovery from the Salem 1 0-ring leak.

6. What is the potential impact to the structural integrity of the reactor vessel and head if leakage beyond both 0-rings were to occur for a prolonged period of time? Are there any additional components that could be subject to damage from boric acid from a flange leak in addition to the vessel, head, and studs?

The concern as related to structural integrity of the reactor vessel would be the condition of the vessel studs. The other concern would be potential wash out of the seating surfaces on the reactor vessel and the closure head which may require substantial remediation.

7. The inner 0-ring has been placed back in-service. Are there any contingencies or other plans to return the outer 0-ring to service if unexpected leakage past the inner 0-ring is observed?

The ODMI was revised to have a trigger action# 4 where leakage exceeds either 1 GPM or a rate of 0.25 GPM /

head. When this trigger is reached, consideration may be given to place the outer 0-ring in service.

4

From: Lorson, Raymond Sent: Wednesday, September 09, 2015 3:12 PM To: Floyd, Niklas; Gray, Mel; Burritt, Arthur; Scott, Michael Cc: Setzer, Thomas; Suber, Gregory

Subject:

RE: IP3 Reactor Stud Corrosion Cale Nik - tlhanks for the follow-up. I don't have any further questions at this time; should the licensee elect to defer the shutdown currently planned for the 14th we will re-visit. In the meantime it would be interesting to follow-up after the shutdown to understand the as-found condition of the studs. May provide some interesting data to apply to future leaks of this nature.

Ray From: Floyd, Niklas Se nt: Wednesday, September 09, 2015 2:15 PM To: Gray, Mel; Burritt, Arthur; Lorson, Raymond; Scott, Michael Cc: Setzer, Thomas

Subject:

FW: IP3 Reactor Stud Corrosion Cale

All, I reached out to one of our materials experts in NRR in order to get a second, independent assessment of Indian Point's boric acid corrosion calculation of the vessel studs. We performed our own calculations with an even more conservative approach (assuming a 2 in/yr corrosion rate around the entire stud) and concluded that the licensee will still be within their ASME code limits, given a plant shutdown on September 14, 2015. I included below the email exchange between myself and John Tsao for your reference.

Please feel free to ask any questions regarding the details of the calculation.

Nik From: Tsao, John Sent: Wednesday, September 09, 2015 1:33 PM To: Floyd, Niklas <Niklas.Floyd@nrc.gov>

Subje ct: RE: IP3 Reactor Stud Corrosion Cale

Nik, I used 2 weeks (from September 1 to September 14) as the time duration and a corrosion rate of 2 in/yr for both the inside and outside surface of the stud. I came up with a ratio of 0.955 which exceeds the allowable ratio of 0.922.

I would not say that I agree or disagree with the licensee's calculations. The licensee has its own justification for the assumptions and analytical approaches.

I think that our assumption of using a corrosion rate of 2 in/yr for both the inside and outside surface is more conservative than the licensee's. The steam and coolant exiting the outer RPV head 0-ring will cover the inside surface and outside surface of the studs. The studs will be corroded by boric acid regardless the location of its surface. Therefore, the corrosion rate for the outer surface should be similar to the inside surface.

1

Based on the information provided in the licensee's corrosion calculation and the staffs (in your case inspector's) independent calculations, the staff concludes that the studs will still be within the allowable stresses in accordance with the ASME Code, Section Ill, given the plant will shut down on September 14, 2015. The licensee has demonstrated the reasonable assurance that the structural integrity of the studs will be maintained until September 14, 2015.

John From: Floyd, Niklas Se nt: Wednesday, September 09, 2015 12:47 PM To: Tsao, John <John.Tsao@nrc.gov>

Subject:

RE: IP3 Reactor St ud Corrosion Cale

John, I also feel that the licensee should have used 2.37 in/yr instead of 2; however, 2 in/yr is still a pretty conservative value in the realm of boric acid corrosion.

The flange leakage started on September 151 based on indication of unidentified leakage and visual indications of boron on the vessel. The 2.5 months is just a bounding calculation to show that the plant is ok to operate for that long beyond September 1. As a comparison, they plan to shut down on September 14th (2 weeks is much shorter than 2.5 months= plenty of margin).

I am not sure where they obtained the 0.2 in/yr corrosion rate for the outside surface. The data that I found for low-alloy steel immersed in high temperature solution ranged from 0.01 to 10 in/yr, where 10 is very extreme and the conditions were stated as likely not possible in a plant setting.

When applying a corrosion rate of 2 in/yr to both the inside and the outside surface of the studs, I get approximately 1 month (26 days). As a comparison with this estimation, 2 weeks is still much shorter than 1 month.

Nik From: Tsao, John Sent: Wednesday, Sept ember 09, 2015 11:44 AM To: Floyd, Niklas <Niklas.Floyd@nrc.gov>

Subje ct: RE: IP3 Reactor Stud Corrosion Cale

Nik, My initial comments.

The licensee used a corrosion rate of 2 in/yr for the inside surface of the studs even though the max corrosion rate I read was 2.37 in/yr. I think that the licensee should have used 2.37 in/yr. However, the difference between 2 and 2.37 is small so the end results should not change too much.

The licensee used a duration of 2.55 months. Is this the time between when the leakage was discovered and September 14, 2015 when the plant will be shutdown? When the leakage occurs did the coolant impinge on the studs right away? Or the studs was in contact with the coolant only after the outer 0-ring was installed recently? In other words is 2.55 month duration correct or too conservative?

2

3. I did not see where the licensee obtained a corrosion rate of 0.2 in/yr for the outside surface of the studs. I did a quick calculation assuming that the corrosion rate for the outside surface of the studs is 2.0 in/yr {the same as the inside surface) using the licensee's method on page 2 of 4. My calculation showed that the studs would exceed the allowable stress limits. Of course I was way conservative.

I will keep reviewing further.

John From: Floyd, Niklas Sent: Wednesday, September 09, 2015 10:23 AM To: Tsao, John <John.Tsao@nrc.gov>

Cc: Alley, David <David.Alley@nrc.gov>

Subject:

IP3 Reactor Stud Corrosion Cale

John, If you have some time today, would you mind taking a quick look at the attached calculation? The licensee used a similar approach as one example in the EPRI Boric Acid Corrosion Guidebook.

Please let me know what you think about their assumptions regarding the corrosion mechanism and corrosion rates. They assumed two different mechanisms (one for the bolt area facing the reactor and another one for the bolt area facing towards the containment).

I've made my conclusions, but I would like a second opinion.

Thank you much!

Niklas Floyd Reactor Inspector Division of Reactor Safety USN RC Region I (610) 337-5282 3

ott,c,al Use Only- Sensitive Internal and Security-Relat ed lnformat,on - Pre-Dec1s1onal Indian Point 2&3 ROP-16 End-of-Cycle Plant Performance Summary January 1, 2015 - December 31 , 2015 Performance Overview and Previous Assessment Results Unit 2 Current Action Matrix Column Licensee Response Basis All Green Findings and Pis Previous Action Matrix Column Licensee Response Previous Basis All Green Findings and Pis I Unit 3 I Current Action Matrix Column Requlatorv Response Basis White Pl in Unplanned Scrams per 7000 Critical Hours Previous Action Matrix Column Licensee Response Previous Basis All Green Findings and Pis I Cross-Cutting Summary I Cross-Cutting Themes N/A Cross-Cutting Issues N/A Prev. Cross-Cutting Themes N/A Prev. Cross-Cutting Issues N/A I Deviations None Assessment Letter Information Assessment Letter Si nature Division Director Unit 2 Power History and Event Declarations Control Room operators initiated a Manual Reactor Trip due to indications of multiple dropped Control Rods. The initiating event was a smoldering Motor Control Center (MCC) cubicle in the Turbine Building that supplies power to the December 5, Rod Control System. The smoldering MCC cubicle had power removed from it 2015 when 24 MCC breaker tripped on overcurrent. Repairs were completed, reactor went critical on December 8 and full power operation resumed December 10.

Pl Im act - Un lanned Scrams er 7000 Critical Hours Unit 3 Power History and Event Declarations Operators commenced a shutdown in accordance with Technical Specification (TS) 3.5.4 due to both refueling water storage tank (RWST) level alarms being inoperable. Unit 3 reached 45 percent power when one level channel was January 8, restored and the shutdown stopped. Operators restored both level channels, 2015 commenced power ascension, and returned Unit 3 to 100 percent power later the same day. (Pl Impact - Unplanned Power Changes per 7000 Critical Hours March 1, The lant was shut down for a lanned refuelin and maintenance outa e

Official Use Only sensitive Internal and security-Related Information - Pre-Decisional 2  ! I ndian Point 2& 3 2015 (3R18). Following refueling and maintenance activities, the reactor was critical on March 24, and returned to power operation on March 25, 2015. Full power operation resumed on April 1, 2015 after troubleshooting and maintenance on 31 main boiler feedwater um .

On May 7, 2015, Unit 3 initiated an unplanned shutdown to a planned trip to repair a weld crack on valve BFD-64-10. Valve BFD-64-10 is a low side isolation May 7, 2015 valve for feedwater flow transmitter FT-438B. The leak was repaired and the reactor was restarted on May 8, 2015 and returned to 100% power on May 9, 2015. Pl Im act - Un fanned Power Chan es er 7000 Critical Hours The unit experienced a reactor trip and fire associated with failure of the 31 main transformer. The unit remained shut down to replace the transformer.

May 9, 2015 Unit 3 reactor went critical on May 25, 2015, and returned to full power on May 26, 2015. Pl Im act - Un lanned Scrams er 7000 Critical Hours The unit tripped from full power on June 15, 2015 due to a switchyard June 15, disturbance. Unit 3 reactor went critical on June 16 and returned to full power 2015 later that da . Pl Im act - Un lanned Scrams er 7000 Critical Hours The unit was manually tripped from full power on July 8, 2015, as result of a feedwater transient. The unit was restarted on July 9, 2015, and operated at July 8, 2015 reduced power while repairs were made to secondary components. Unit 3 returned to full power operations on July 11, 2015. (Pl Impact - Unplanned Scrams er 7000 Critical Hours The unit was shut down for a planned maintenance outage to replace the September reactor vessel 0-rings. Following maintenance activities, the reactor was critical 15, 2015 on September 25, 2015, and returned to power operation on September 26, 2015.

After receipt of a Main Generator Lockout trip signal, the reactor automatically tripped. Site personnel reported seeing arcing on a 345kV output transmission December line tower. Tower insulators were replaced and/or cleaned. The reactor was 14, 2015 critical on December 17, 2015 and returned to full power operation later that da . Pl Im act - Un fanned Scrams er 7000 Critical Hours Safety-Significant Inspection and Performance Indicator Results Issue Green-White Threshold exceeded for Pl - Unplanned Scrams per 7000 Critical Hours Safety White Significance Date 4th Quarter 2015 Supplemental The licensee has not yet declared readiness; the 95001 is not scheduled yet.

Inspection Status Traditional Enforcement Summary

  • EA-14-180 - SLIII Problem (2 violations in a single NOV) - impacting the regulatory process.

issued March 16, 2015. The first violation involved Entergy's failure to notify the NRC within 30 days after learning, on October 25, 2012, of a change in a Unit 3 reactor operator's (RO's) medical condition that involved a permanent disability/illness (sleep apnea). Entergy also did not request an a mended license with a condition to account for the medical issue, resulting in the RO performing licensed operator duties without a properly restricted license.

The second violation involved Entergy's submittal of information to the NRC in a December 3, 2012, application for renewal of the RO's license that was not complete and accurate in Offa;ial Yse GAly SeAsiti¥e IAterAal aAa See1:1rity Related IF1fem,,11tie11 Pre Deeisioilal

Official Use 0, ,Iv - Se, ,sitive Ii ,te, , ,al a11d !lecur l ty-1\ela ted 111for rnatlon - Pre-Detlslonal Indian Point 2&3 3 all material respects. Specifically, the application did not specify that the RO had a medical condition that required a restriction (for use of a Continuous Positive Airway Pressure (CPAP) machine). Based, in part, on this inaccurate information, the NRC issued a license renewal that did not contain the necessary restriction.

  • IR 2015005- SL-IV NCV, impacting the regulatory process, issued August 7, 2015.

On April 27, 2015, IPEC issued an LER in accordance with 50.73 which reported three MSSV test failures (MS-46-2, 45-4, and 47-4) that occurred on February 27, 2015.

However, the LER did not discuss the failure of MS-46-3, which also failed its TS as-found lift setting test and was declared inoperable on March 22, 2015. MS-46-3 was inoperable for greater than its TS allowed outage time, which is a condition prohibited by TSs, and therefore is required to be reported to the NRC.

Inspection Followup: IP 92723 was successfully completed at Indian Point during the week of December 7, 2015, and is documented in IR 2015004. This inspection reviewed actions planned and taken in response to the above SL IV and SLIII violation, an additional SLIV violation from 2014 (involving a failure to report the loss of a moveable in-core detector within one hour, as requ ired); these violations spanned the timeframe from July 1, 2014, to July 1, 2015.

  • (b)(5)

Potential Safety-Significant Inspection Findings

  • None Performance Indicators Close to Threshold
  • None Open Unresolved Items Issue Entergy failed to identify all critical digital assets (CDAs) in accordance with the NRC approved cyber security plan Implementation Schedule for Milestone 2.

Specifically, Entergy evaluated that digital devices used for access authorization such as the site's Security Access Management System (SAMS) were not classified as CDAs.

Date 4Q2015 Opened Licensee Based on an NRC inspection at Arkansas Nuclear One, Entergy initiated Actions CR-HQN-2015-00349 on April 10, 2015, to track actions necessary to address Official Use 01119 Se11sitive l11 te11 ,al a11d Secu, i ty-Rela led Ii ,Fo11 11atio11 P, e Decisioilal

Offtcial Use Only 3e, ,sitive Ii ,te, 1,el t1Ra See1,rity Related IPfocroarioo - Pce-pecisional 4  ! Indian Point 2& 3 NRC clarifications in these letters and ins ector concerns.

NRC Next NSIR will provide additional guidance to the Region and the industry as to Steps whether the digital devices for the Access Authorization Program should be classified as CDAs.

Open 01 Investigations

  • None Third Party Reviews
  • Resident staff reviewed the results from the most recent WANO/IN PO plant performance assessment conducted during December 2015 (report not yet issued). There were no conflicts between the NRC assessment and the WANO draft assessment results.

The inspectors reviewed the final INPO plant assessment of Indian Point, Units 2 and 3, conducted in December 2013. The inspectors evaluated these reports to ensure that NRC perspectives of Entergy performance were consistent with any issues identified during the assessments. The inspectors also reviewed these reports to determine whether INPO identified any significant safety issues that required further NRC follow-up. The results of this review were documented in the 2nd quarter resident inspection report 2014003.

Official U5e Onl9 SeRsiti¥e IRterRal aREI ~ee1,rity Rela:teg IPt,m.iatigp Pce-DecisioPal

Offltlal Use 01119 - Se, ,sitive l11te11 1al a11d Seca, ity-Rela led Ii ifo1111atio11 P, e Decisio11al Ind ia n Po int 2&3 5 Analysis of Cross-Cutting Areas (for finding specifics, refer to the PIM)

Area Aspect 1015 2015 3015 4015 Total H.1 Resources 1 1 H.2 Field Presence H.3 Chanqe Manaqement (l)

H.4 Teamwork 1 1

(.)

C: H.5 Work Management (1J E H.6 Design Marains 2 2

.... H.7 Documentatio n 1 1 2

.g (l) H.8 Procedure Adherence 1 1 2 a.. H.9 Traininq C:

(1J H.10 Bases for Decisions E

, H.11 Challenqe the Unknown 2 2 I

H.12 Avoid Complacency 1 1 H.1 3 Consistent Process H.14 Conservative Bias 1 1

_Total @acks~

P. 1 Identification 4

- 4

-- 3

--- 1 12 P.2 Evaluation 1 1 2 0::: P.3 Resolution 1 1 1 3

~ P.4 Trendinq a: P.5 Ooeratina Experience 2 2 P.6 Self-Assessment Total (Backstop) 1 2 3 1 J 7 r S.1 SCWE Policy Alternate Process for Raising S.2 Concerns w S.3 Free Flow of Information

~ The licensee has received a chilling effect

(.)

Cl) letter The licensee has received an escalated I (SLIII or higher) enforcement action or order involving discrimination.

Human Performance and Pl&R A theme or CCI does not exist per IMC 0305.

On January 27, 2016, the most recent Pl&R biennial inspection was completed. Based on the samples selected for review, the inspection team concluded that Entergy was generally effective in identifying, evaluating, and resolving problems. Entergy personnel identified problems and entered them into the corrective action program. Entergy generally prioritized and evaluated issues commensurate with the safety significance of the problems and corrective actions were generally implemented in a timely manner. In addition to implementation of the corrective action program, the inspectors also reviewed Entergy's use of operating experience, conduct of self-assessments, and safety conscious work environment at the station. Based on the samples selected for review, the inspectors did not identify any issues with Entergy's use of industry operating experience. The inspectors concluded that the self-assessments reviewed were Of£icial J !se Oolv - Seosi1ive Jot ecoal aod Sern city-Belated lofacmatiao - Pre-Decjsjonal

Offltlal Use Only Sensit ive l11te1 nal and !lecur l ty-Rela ted Ii lfu, 111atiu11 - fl, e Decisio11al 6  ! Indian Point 2& 3 effective in identifying issues and improvement opportunities. Finally, the inspectors found no evidence of significant challenges to Indian Point's safety conscious work environment Based on the inspectors' observations, Indian Point staff are willing to raise nuclear safety concerns through at least one of the several means available. However, two security-related green NCVs were identified during the inspection. One cross-cutting aspect was assigned in the area of Human Performance, Teamwork, because Entergy work groups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear security is maintained [H.4]. Additionally, one cross-cutting aspect was assigned in the area of Problem Identification and Resolution, Resolution, because Entergy failed to take effective corrective actions in a timely manner commensurate with safety significance [P.3].

Safety Conscious Work Environment Concerns

  • None Planned Refueling Outages, Temporary Instructions, and Other Significant Activities Tl 2800/041 , 10 CFR Part 37 Physical Protection of Category 1 01 /11/2016 - 01/15/2016 and Category 2 Quantities of Radioactive Material at Facilities with a 10 CFR 73 Physical Protection Program Unit 2 Refueling Outage (2R22) 03/07/2016 - 04/06/2016 Unit 3 Refueling Outage (3R 19) 03/06/2017 - 03/31/2017 Tl-191 FLEX Order Implementation (dependent on SE issuance) 04/17/2017-04/21/2017 Units 2&3 Triennial Fire Protection 05/01 /2017 - 05/05/2017 05/15/2017 - 05/19/2017 Biennial Pl&R 11/27/2017 - 12/01/2017 12/11/2017 - 12/15/2017 Proposed Appendix C Inspections through December 31 , 2017 IP 60845, Operation of Inter-Unit Fuel Transfer Canister and Cask 07/25/2016- 07/29/2016 S stem 09/04/2017 - 09/08/2017 Planned 2016 Pl&R Samples Corrective actions for the 31 main transformer fire on Unit 3: Unit 3 tripped when the 31 main transformer faulted and caught on fire on May 9, 2015 (CR-IP3-2015- 02913) DRS DRS - J.

(Patel) is scheduled to review the root cause and corrective 102016 Patel actions when the RCE is approved by CARB in February 2016.

Also verify that previous transformer failures were included in the CAP products.

Unit 2 Main Boiler Feed Pump loss of trip-ability due to control oil problems: the Unit 2 main feedwater pump failed to trip automatically and manually from the control room during the Unit 2 reactor trip on December 5, 2015 (CR-IP2-2015-05459).

DRP 102016 The feedwater pump had to be tripped locally. Tech Spec 3.7.3 (Main Feedwater Isolation) requires the main feedwater pump to be operable. The resident office (B. Haagensen) will review the aooarent conflict between the Tech Spec requirement and the Offltlal Use Only Sensitive Ii 1te11 1al and ~ecu, ity-fl.ela ted l11fu1111atio11 fl, e-Decisio11al

Official Use 0 1119 Se113itioe 11,tw'lel BRB See1:1rity RelatQa IRfan:;:iation l?ce-Decisiooal Indian Po i nt 2&3 7 degraded condition in the 1st quarter of 2016 (in progress).

Maintenance Rule process for balancing reliability with availability: The residents have noted in the past that the process for balancing reliability and availability is not conducted in a systematic way. This problem was noted recently during the DRP 202016 Pl&R inspection by S. Rich. There is no CR for this issue of concern. The resident office (S. Rich) will review the balancing process during the in 2nd quarter of 2016.

Commercial dedication of non-qualified QA parts in safety systems: There have been several recent allegations regarding the installation of non-qualified parts in safety-related systems with commercial dedication occurring after the part has been placed in service. The individual allegations have been properly DRP 302016 dispositioned. The resident office (G. Newman) will review the effectiveness of the changes made in response to these allegations to the commercial dedication process during the 3rd quarter 2016.

Effectiveness of the corrective actions for the degraded conditions that caused the Unit 2 reactor trip on December 5: DRS will review the resolution of the various degraded conditions noted during the Unit 2 reactor trip. This included a DRS 202016 failed 480 volt MCC 24 (CR-IP2-2015-05464 ), a degraded alternate power supply for the rod control that overheated and caused control rods to drop as well as valves that failed to properly stroke (CR-IP2-2015-05460, 05461, 05466, 05467).

Review of the causes and actions being taken to address the January 2016 tritium groundwater leak. Review actions being taken to investigate the rad waste drain system and DRS 102016 causes of contaminated water exiting the building envelopes and entering the groundwater. (1st indication that the leak occurred was on 1/16; see CR-IP2-2016-00264 / CR-IP2-2016-00226.)

Service water system piping leaks: There have been numerous leaks in ASME class Ill service water pipes over the past year. This has included the 21 fan cooler unit motor (CR-IP2-2015-05755), 21 CCW heat exchanger supply line (CR-IP2-2015-05358), 24 fan cooler unit, and the 32 CCW heat DRS 2 or .3Q2016 exchanger supply line. Inspection should include verificat ion this is captured in Maintenance Rule appropriately. Staffed by DRS in 2nd or 3rd quarter of 2016 after NRR has completed the license renewal audit of the service water system.

MT&E Program equipment calibration and operability issues: There have been a number of issues identified with MT&E that was not properly calibrated. A review will be DRP 302016 conducted to assure that issues identified in multiple condition reports are beina adeauatelv identified and resolved.

Extent of condition and root cause analysis for the Unit 3 reactor head flange gasket leak: The Unit 3 reactor vessel head flange gasket leaked following the refueling outage in DRP 302016 March. This leak resulted in a plant shutdown to repair the leak in September. Unit 3 has had several problems with leakina Official Use 0111 v Se11siti oe li1le11 ,al a11d Seculit9-Rela led Ii ,fu1111atio11 fl, e-E>ecisio11al

Gffaiial lJse Gr:tly l!ier:tsitive lr:tt ernal ar:te ~e1;1,1rit~* Relat ee lr:tferr:t:1atier:t

  • Pre i;>e1;isier:tal 8  ! I ndian Point 2&3 reactor vessel head flanges over the past 10 years. Unit 2 had not has any problems of this nature. This sample will review the differences between Unit 2 and Unit 3 as well as the root causes of the head gasket leak and determine if they have been adequately corrected.

Cyber Security. Follow up on corrective actions from T l- TBD 2201/104 as requested by NSIR DRS (CY2016 or CY2017)

Other Items of Interest

  • License Renewal: Eleve*n contentions have been heard: one was settled, one was resolved in favor of New York, and nine were resolved in favor of the Entergy and the Staff. Appeals from the Board's resolution of two contentions are pending before the Commission.

Hearings on the three remaining safety contentions were held in Tarrytown, NY, on November 16-19, 2015; the Board is not likely to issue a decision before the latter half of 2016.

The NRC Staff issued its. second supplement to the S.afety Evaluation Report (SER) for license renewal of IP2 and IP3 in November 2014, which it published in July 2015. The NRC staff has scheduled an onsite audit of Entergy's service water integrity and fire water system aging management programs during the week of February 22, 2016. The audit will focus on how Entergy manages recurring internal corrosion in the service water integrity and fire water system, as well as aging effects managed by the Fire Water System Program.

The staff's review of these issues was prompted by the issuance of interim staff guidance document LR-ISG-2012-02, "Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation."

On December 22, 2015, the Staff issued a draft for public comment of the second supplement to the Final Supplemental Environmental Impact Statement (FSEIS Vol. 5) for license renewal of IP2/IP3. This supplement addresses new information received by the staff since preparation and publication of the previous FSEIS Supplement (FSEIS Vol. 4) in June 2013. The public comment period closes on March 4, 2016. The staff expects to issue the final FSEIS Supplement in September 2016. It is possible that the draft and/or final FSEIS supplements. may trigger the filing of new environmental contentions.

  • Leak Rate Test Interval Extension: In December 2014, Entergy filed a license amendment request for Indian Point Unit 2 that would permanently extend the frequency of the containment integrated leak rate test (ILRT) from once every 10 years to once every 15 years. The last ILRT was performed in 2006 and will need to be performed during the upcoming March 2016 refueling outage unless the staff approves the amendment request.

In May 2015, the State of NY Attorney General's Office objected to the amendment request and filed a petition to intervene and request for hearing. The ASLB denied the request in their decision dated September 25, 2015. The State of NY AG Office filed an appeal on October 20, 2015, and the final decision is now before the Commission. The NRR technical staff recommends approval of the request and NRR/DORL is preparing a Notification of Significant Licensing Action (NSLA) informing the Commission that the staff intends to take the licensing action on or about February 12, 2016. If the Commission does not object within 5 working days following receipt of the NSLA, the staff can take the proposed action.

In early 2015, the staff granted Entergy's similar request to extend the ILRT interval to 15 years for Indian Point Unit 3.

Offieisl U5e ORiy SeAsiti¥e lr:ttern;.I aAd Secwi!y-Bela!ed lofocroatiao Pre-Pecisjonal

Offltlal Use 01119 - Se, ,sitive l11te11 1al a11d Seca, ity-Rela led Ii ifo1111atio11 P, e Decisio11al Indian Point 2&3 9

  • Spectra Energy's Proposed Natural Gas Pipeline (AIM Project): Spectra Energy approached Entergy during the summer of 2013 about plans to expand their natural gas pipeline capacity across the Hudson River with a new 42-inch diameter pipeline. On February 28, 2014, Spectra filed an application with the Federal Energy Regulatory Commission for a certificate to build a new 42-inch natural gas pipeline along a southern route on Indian Point property.

Entergy performed a site hazards analysis to determine the impact of the new natural gas pipeline on the site. On August 21, 2014, Entergy submitted a 50.59 evaluation and blast analysis for NRC review. A Region I DRS security inspector and Headquarters expert on blast analysis performed an ROP baseline inspection (71111.18 - Plant Modifications) of the 50.59 and blast analysis. The results of the inspection were documented in the 3rd quarter 2014 resident inspection report. In summary, the inspectors determined Entergy had appropriately concluded that the proposed pipeline does not introduce significant additional risk to safety-related SSCs and SSCs important-to-safety at Indian Point Units 2 and 3; and, therefore, the change in t he design bases external hazards analysis associated with the proposed pipeline does not require prior NRC review and approval.

FERC approved Spectra Energy's application on March 3, 2015, and construction has begun in areas near the Indian Point owner controlled property. The pipeline continues to receive considerable stakeholder interest and opposition. FERC has received multiple Congressional and local requests to reopen hearings. On January 28, 2016, FERC issued an order denying a rehearing and dismissing the stay request regarding the proposed Algonquin Incremental Market (AIM) pipeline project at Indian Point.

  • Indian Point 2 Spent Fuel Pool Criticality Analysis: The existing Unit 2 technical specifications for SFP criticality are non-conservative and compensatory measures have been implemented. The licensee initiated a SFP management improvement program that was originally scheduled for completion in 2016. A SFP criticality analysis, prepared by NETCO (a business segment of Scientech Curtiss-Wright), was submitted for NRC review and approval in November 2014. On November 23, 2015, the staff provided conditional approval of the NETCO report for reference at Indian Point 2. The criticality analysis takes credit for new boron inse*rts currently being designed by Holtec. Boron impregnated inserts will be installed in two phases pursuant to 10 CFR 50.59. Phase 1, originally scheduled for late 2015, will include regions of the SFP that have experienced the most degradation of the boroflex inserts. Phase 2 , originally scheduled for late 2016, will complete the process. The schedules for both Phase 1 and 2 have been delayed.

Entergy plans to submit a license amendment request that references the NETCO criticality analysis and seeks NRC approval of the new neutron absorbing inserts and revised technical specifications.

  • Weapons Preemption: The Indian Point security force currently uses weapons and large-capacity magazines that are banned by state and local laws. On January 5, 2016, NRR issued the confirmatory order, conforming amendments, and supporting safety evaluation that will permit security p,ersonnel at Indian Point to transfer, receive, possess, transport, import, and use certain firearms and large capacity ammunition feeding devices not previously permitted to be owned or possessed under Commission authority, notwithstanding certain local, state, or federal firearms laws, including regulations that prohibit such actions.

Offa;i3l 'IH 1 Only Seositive lo! eroal aod Sernrity-Belated Information Pre-Decisional

OHieial Use 0 11ly Semiti oe II 1te.i ,al a11d Secu1ity Rel a led Ii ,Fu, 11 ,aliu1, - fl, e-Elecisiu11al 10  ! Indian Point 2&3

  • NY Department of Public Service Investigation: On December 16, 2015, Governor Cuomo directed the NYS Department of Public Service (DPS) to launch an investigation into the operations and safety protocols of Indian Point. The direction was in response to the Unit 3 trip caused by the failure of an insulator on a high voltage transmission line. The trip was uncomplicated, but did result in the performance indictor for unplanned scrams crossing over the white threshold , and follows a trip of Indian Point Unit 2 which occurred the prior week. The NRG continues to monitor the State's investigation. OGG has been involved since the investigation is potentially pre-empted by NRG authority. The NRG staff continues to maintain an open dialogue with the State and respond to their questions.
  • State 401 Water Quality Certificate: On April 2, 2010, New York State denied Entergy's request for a Clean Water Act (CWA) section 401 Water Quality Certificate (WQC) for the period of extended operation unless cooling towers were constructed. Entergy has appealed the ruling to the State. The hearings were split by topic and began in October 2011 . Hearings will take place in 2016 on the two remaining topics. The DEC's current best estimate for a final decision on Entergy's appeal is early 2017, although that could be extended further.

Entergy is continuing to operate under its existing WQC and SPDES permit, which is in effect under the timely renewal doctrine and will continue in effect until Entergy's pending application for a SPDES permit under the CWA is resolved in the ongoing New York State adjudication. The NRC has not determined whether it will issue a renewed license for operation of IP2/ IP3 prior to the completion of New York's administrative proceedings.

  • Coastal Zone Management Act and Consistency Review: On December 17, 2012, pursuant to the Coastal Zone Management Act, Entergy provided a consistency certification to the NRG and furnished a copy to NYSDOS. However, by letter to the NRC dated November 5, 2014, Entergy withdrew this consistency certification pending issuance of the NRC Staffs 2016 supplement to its Final Supplemental Environmental Impact Statement ("FSEIS Supplement") concerning IPEC license renewal. NYDOS stated that Entergy did not have the right to withdraw its application. On November 6, 2015, NYSDOS denied Entergy's certification. On November 10, 2015, Entergy appealed to the Department of Commerce, National Oceanic and Atmospheric Administration (NOAA), asking that NOAA find the NYSDOS decision invalid and based on Entergy's withdrawal. On November 25, 2015, NOAA deferred its decision pending the outcome of the ruling by the New York Court of Appeals (discussed below).

On December 11, 2014, the NY State Supreme Court Appellate Division declared Indian Point Units 2 and 3 exempt from a coastal zone management consistency review under the New York Coastal Management Program. A five judge panel unanimously agreed to reverse an earlier ruling by a lower court that had upheld a New York State Department of State (NYSDOS) decision denying the two Indian Point units from being grandfathered under, and therefore exempt from, the state Coastal Management Program.

This appellate court ruling affirmed Entergy's position that Indian Point does not need to seek a new consistency certification from NYSDOS in order for the plants to have their operating licenses renewed by the NRC. The court ruled that because Indian Point began operating in the mid-1970s and an environmental impact statement for each unit was prepared prior to the effective date of New York State's coastal zone management regulations, the consistency certification for Units 2 and 3 is automatically grandfathered for Offa;i3l l lgg GAly SeAsiti¥e IAterAal aAEI ~eel:lrity Relateel IAfeFFl'latieA Pre DeeisieAal

Gffaiial lJse Gr:tly l!ier:tsitive lr:tt ernal ar:te ~e1;1,1rit~* Relat ee lr:tferr:t:1atier:t

  • Pre i;>e1;isier:tal Indian Po int 2&3 11 the plants' entire lifetime under state law. New York has appealed the decision to the New York Court of Appeals; briefing of that appeal is scheduled to be completed in early 2016.

On January 14, 2016, Entergy Nuclear filed suit against Cesar Perales, in his official capacity as the Secretary of the New York State Department of State ("NYDOS"). The suit is related to the licensee renewal application for Indian Point currently under review by the NRG- specifically Coastal Zone Management Act ("CZMA") compliance. Absent an exemption, the CZMA allows a state agency to object to the granting of a federal license on the ground that the facility is inconsistent with the state's coastal management program

("CMP"). Such an objection, unless overturned, prevents the federal agency from issuing the license. On November 6, 2015, NYSDOS issued an objection concerning Indian Point's license renewal application. According to Entergy, the objection repeatedly relies on nuclear safety concerns and, further, the objection's purportedly nonnuclear safety rationale (i.e.,

that Indian Point harms aquatic species in the Hudson River) is but merely a pretext for NYSDOS's nuclear safety concerns. Thus, Entergy argues that the objection thus intrudes on the field of the NRC's exclusive regulatory authority over nuclear safety concerns, and is as such preempted under well-settled case law. Entergy seeks a declaratory judgment that the objection is invalid and an injunction ordering the objection to be withdrawn. While the suit mentions the New York Department of Public Service ("NYDPS") investigation (initiated at the direction of Governor Cuomo in a letter dated Nov. 16, 2015) as further evidence of a broad "effort by [NY] to undertake preempted regulation of Indian Point on nuclear safety grounds," Entergy does not request as part of the suit any declaratory judgment or injunctive relief with respect to said NYDPS investigation.

  • Groundwater Tritium: Due to previously identified tritium groundwater contamination events during Unit 2 refueling outages 2R19 (2010) and 2R20 (2012), prior to the start of 2R21 (February 2014), IPEC increased the frequency of sampling in monitoring wells around the Unit from quarterly to monthly. In March 2014, monthly well samples from monitoring wells (MW) 31 and 32 showed significant increased tritium activity (up to 660,000 pCi/1). IPEC initiated a Condition Report (CR) and began conducting an investigation as to the source of the tritium contamination using the Kepner-Tregoe (KT) Analytic process. IPEC investigation activities also included performance of an underwater remote video examination to VT-1 inspection standards of the cask loading area of the Unit 2 spent fuel pool. IPEC ultimately attributed the tritium to a mid-February 2014 drain down activity of the containment spray header, when reactor coolant spilled on the floor of the 51' elevation of the pipe penetration building, rather than flowing down a floor drain at that location.

A Problem Identification and Resolution team inspection (Pl&R) was conducted on May 4-6, 2015. Results of the inspection included one licensee-identified Green inspection finding for a violation of Plant Technical Specification 5.4.1.a, in that Entergy did not evaluate work practices involving changing the mode of operation to drain the containment spray system to a floor drain in the Unit 2 Auxiliary Building to assure that the drainage operation did not reach groundwater.

Subsequent to the Pl&R inspection, in July 2015, the NRC recognized that groundwater tritium levels at MW-30 had spiked to 938,500 pCi/1 in February 2015, and remained above 400,000 pCi/1, some eleven months after the event identified by IPEC as the root cause had been terminated . This value was the highest tritium levels ever measured at IPEC since the start of the groundwater monitoring program in the mid-2000s. A telephone conference with IPEC was held on July 7, 2015, and based on that meeting, together with further review by the NRC staff, it was concluded that the February 2015 tritium spike was not part of the Official Use 0 1119 Se11sitive li1 te11 1al a11d ~ecarlty-Rela ted Information Pre-Dec1s1ona l

OHieial Use ORiy SeRsiti¥e IAt eFAal aAel See1:1 Fity Relateel IAieFFAatiaA PFe i;>eGi&ier:tal 12  ! Indian Point 2&3 2014 event, and represented a new leak, which had not been reviewed by the licensee. The NRC identified a Green Finding for a violation of 10CFR20.1406(c), in that Entergy did not conduct operations to minimize the introduction of residual radioactivity into the subsurface due to its failure to identify a leak of tritium seen in February 2015 as unrelated to the previously identified leak of tritium in March 2014. This was documented in the 3rd quarter integrated inspection rep,ort.

In mid-September 2015, IPEC called the NRC (D. Mayer to J. Noggle) to inform the NRC of IPEC's plans to complete installation and operation of a recovery well previously installed (RW-1 ), to collect the subject tritiated groundwater and return it to Unit 2 for normal liquid radioactive effluent processing. IPEC had previously identified this as a potential corrective action for the March 2014 spill. While the cause of the currently elevated groundwater tritium concentrations is unclear, the installation of the Unit 2 groundwater recovery well and its planned operation beginning in the summer of 2016, is expected to address the current groundwater contamination issue. As of December 7, 2015, the current tritium value was 457,400 pCi/1 at MW-30-69.

  • 2.206 Petition: By letter dated October 15, 2014, Paul Blanch submitted a 2.206 petition critical of Entergy's 50.59 site hazards analysis report associated with the proposed Spectra Energy 42-inch diameter natural gas pipeline. Mr. Blanch asserted that Entergy's 50.59 analysis was inaccu rate and incomplete resulting in violations of 10 CFR 50.9, "Completeness and Accuracy of Information," and possibly 10 CFR 50.5, "Deliberate Misconduct." The petition gathered significant local stakeholder and political interest.

The petitioner made presentations before the PRB on January 28 and July 15, 2015. On September 9, 2015, the PRB rejected the petition on the basis that all issues identified in the petition and its supplements had been previously reviewed and resolved by the NRC staff.

Despite the rejection, the petitioner continues to pursue the issue with the staff.

  • 31 Main Transformer Fire: On May 9, 2015, at 5:50 p.m., IP3 experienced a main turbine-generator lockout, main turbine trip, and automatic reactor trip as a result of an explosion and fire on the 31 Main Transformer. The site fire brigade was activated at 5:52 p.m., and called for offsite resources at 5:53 p.m. A Notice of Unusual Event was declared at 6:01 p.m. due to an explosion in the Protected Area. The fire was initially extinguished at 6:15 p.m.; however, it reflashed at 6:37 p.m., at which time additional offsite resources were brought into the Protected Area. The fire brigade continued to fight the fire with foam and declared the fire extinguished at 8:05 p.m. During the event, there were reports of water in accumulation in the 480V switchgear room, totaling approximately 1 to 2 inches in depth.

The source of the water was determined to be deluge system water that had come from a deluge valve room that sits adjacent to the 480V switchgear room. The water was unable to completely drain through a floor drain due to its limited capacity, resulting in it making its way into the 480V switchgear room.

An IMC 0309 review was completed, and a Special Inspection Team (SIT ) was sent to IPEC on May 19 in order to review Entergy's response to the event, equipment performance and design, and the licensee causal analyses. The SIT inspection report (IR 2015010) was issued on July 23rd with one Green NCV associated with Entergy's failure to correct a degraded condition offire protection system solenoid valve SOV-230-1 after previous failures of the valve during testing in 2011 , 2014 and 2015.

Official f lse Only Sensitive Internal and Sec11city-Relatgd IAfgn::Ae1tign ~rg OQbiSiQAeil

Official Use 01119 Semitioe IRteFAal aREl ~gc**rity-Belated lofacroatioo - Pre-Decisional Indian Po int 2&3 13 A Maintenance Rule sample regarding the main transformer failure was documented in 2015003 with no findings. As a part of that sample, a system review, and a review of the 31 main transformer failure was performed to ensure the effectiveness of maintenance activities. Past planned and corrective maintenance were also reviewed to verify they had been performed in accordance with work instructions. An internal flooding sample was documented in 2015003, and identified a Green finding because Entergy allowed the Unit 3 480V switchgear room floor drains to become blocked such that they could not mitigate an internal flood if both SW relief valves in the switchgear room lifted.

A Pl&R sample will be performed once Entergy's RCE is complete.

Key Messages

  • (b)(5)

Official Use 01119 SeAsiti¥e IRterRal aAd Secwity Belated lofocroatiao Pre-Decisional

Ofttc1ai Use Only Sensitive Inter nal and Seca1 ity Related ii ,fent1atiaA Pre i;>eGi,ional 14 I Ind ia n Point 2 & 3 (b)(5)

Attachments

1. Plant Specific Action-Matrix History Chart
2. Draft End-of-Cycle Assessment Letter
3. Plant Issues Matrix (Previous Four Quarters, all cornerstones)
4. RPS Report 22 and RPS Report 24 (1!pcoming 24 Months)
5. Performance Indicator for Unplanned Scrams per 7000 Critical Hours Offieial Use ORiy SeAsiti¥e IAterAai and Sei;;*,rity ~Ql3:tQg lntocrmt;on Pre-Decisional

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Indian Point Unit 2 forced outage from December 9 through December 23'd.

Response to questions from Paul Blanch:

Indian Point Unit 2 was shut down on December gth to replace a reactor coolant pump shaft seal that was approaching the allowed leakage limit. Shortly after the reactor was shut down, a containment entry was made by licensee personnel. During the entry into containment, boron was discovered in the vicinity of the reactor pressure vessel flange. The leak was small enough that it was not detected by either the reactor pressure vessel outer seal drain line or the containment air particulate detector. Total unidentified leak rate for Unit 2 at the time of the shut down was less than ten percent of the allowed limit of one gallon per minute. This limit is specified in the operating license technical specifications.

There was no corrosion of the reactor pressure vessel or head due to the leakage from the vessel o-rings. An in-service inspection specialist from Region I was dispatched to the site to perform inspection of components exposed to the boron leakage. Several of the 64 reactor pressure vessel closure studs were exposed to boron and were cleaned, inspected, and evaluated by the licensee. One of the studs did not meet acceptance criteria for re-use and was replaced with a spare stud . Both o-ring seals were replaced during the reactor vessel re-assembly. The design of the seals was modified slightly based on engineering recommendations, in an attempt to eliminate the seal leakage. There has been no evidence of leakage of the Inner or outer seals since the unit came back on line on December 23rd. These o-ring seals will be replaced again during the Unit 2 spring refueling outage.

Licensees perform Root Cause Analyses to determine the root and contributing causes, specify corrective actions, determine the scope of extent of condition and extent of cause review, and determine if a poor safety culture played a role in the causal factors. Root cause analyses performed by the licensee are typically completed 30 to 45 days after initiation. The NRC reviews these root cause evaluations as part of the reactor oversight process. Violations of NRC regulations are sometimes discovered during the review of these root cause analyses.

Any violation that is determined to be more than minor is documented in the quarterly integrated inspection report. There were no events or notification thresholds met by Indian Point during the December 2017 forced outage that required a report to the NRC.

Indian Point Unit 3 Reactor Pressure Vessel 0-Ring Seal Leak Key Messages

  • The 0-ring leakage and resulting increase in containment airborne particulate activity levels are within the plant's Technical Specifications and license requirements and does not currently pose a threat to public health or safety.
  • The licensee has scheduled a near-term shutdown of Unit 3 to perform repairs and is monitoring the issue to ensure the plant can be operated safely despite the small amount of leakage through the 0-rings. Additionally, the licensee Is conducting engineering evaluations and failure modes analyses to implement repairs and prevent recurrence.
  • NRC Resident Inspector staff continue to closely monitor the situation and have no immediate safety concerns.

Facts

  • The reactor pressure vessel head is attached to the pressure vessel by 56 2-lnch closure studs and sealed using two concentric 0-rings: an inner and outer. Either 0-ring is designed to provide the sealing function.
  • Following startup from the last Unit 3 refueling outage (Spring 2017), the inner 0-ring was in service and exhibited leakage, which was followed by a steady increase in containment unidentified leakage. The Hcensee then placed the outer 0-ring in service, which also began to leak.
  • The 0-ring leakage and resulting increase in containment airborne particulate activity levels are within the plant's Technical Specifications and license requirements and currently pose no threat to public health and safety.
  • The licensee plans to conduct a shutdown of Indian Point Unit 3 on June 11 to repair the 0-ring leakage. The Resident inspector staff will closely monitor and inspect the issue.
  • There is a history of 0-ring leaks at Indian Point Units 2 and 3, dating back to the 1990s.

Indian Point is currently conducting a failure modes and affects analysis that includes reviewing operating experience and benchmarking other utilities. The cause of the leak is currently unknown; however, foreign material may have migrated to the vicinity of the 0-ring seating surfaces, during recent outage-related activities.

  • The Resident Inspector staff are conducting a Problem Identification and Resolution inspection of the current and previous reactor vessel head 0-ring leaks.

From: Haagensen, Brian Sent: Friday, June 09, 2017 8:17 AM To: Ambrosini, Josephine; Haagensen, Brian; Highley, Christopher; McKown, Louis; Rich, Sarah; Setzer, Thomas; Siwy, Andrew; Henrion, Mark

Subject:

RPV Head Seal Leakage assessment IP3 RPV Flange seal leakage may have increased over the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Total leak rate has increased from 0.23 to 0.27 gpm today. Theo-ring leakage is a combination of unidentified leakage coming out the flange seating surface and identified leakage flowing through the o-ring seal leakoff line Into the RCDT.

The current leak rate trend is shown below. It is hard to measure RPV o-ring seal leakage alone because it is a combination of UIL and IL. Flange leakage partially contributes to identified leakage if the flange leakoff line is flowing boric acid from the o-ring leak into the RCDT. We are seeing an increase in the area around the flange seating surface where the boric acid is leaking out. There is no other information to indicate why measured identified leakage would increase when the UIL decreased except for the o-ring leakage going into the RCDT.

The planned outage will start on June 11 at midnight. The power reduction will begin around 2200 this evening.

I P3 RCS Leak8ge (b)(4)

Brian C. Haagensen Senior Resident Inspector Indian Point Energy Center 914 739-9360 (Office)

(b)(6) (Cell~

Indian Point Reactor Vessel (RPV) 0-ring Seal Leaks Assessment: June 8, 2017

  • Safety perspective of the issue: There* is little immediate safety significance to small amounts of RCS coolant leakage through the RPVowring flange seating surfaces.

Increasing RCS leakage may increase containment airborne particulate activity levels and challenge Tech Spec limits for RCS leakage resulting in a plant shutdown. Hot boric acid may intrude onto the RPV head studs (made of carbon steel) which may experience boric acid degradation. Ultimately, the reactor vessel integrity could theoretically be challenged if sufficient stud wastage occurs but this is highly unlikely considering the large number of studs (32) and the limited corrosion rates that would be experienced over a refueling cycle.

  • How it was identified: Following startup from the last refueling outage, the inner owring failed followed by a steady increase in RCS unidentified leakage (0.01 gpm to 0.12 gpm) over a five day period. Containment airborne particulate activity levels showed a slight increase as expected. Operators entered containment at power and visually observed boric acid leakage seeping out around the RPV head flange area.
  • Previous history: In 2015 following the last previous refueling outage, IP3 was shutdown to repair o-ring leakage. IP2 also experienced a recent failure of an inner o-ring seal in March 2017. RPV head o-ring seal failure history:

o 1986 - 1989 there were a cluster of 4 o-ring leaks split between both units in the late o 1995 - the o-ring design was then changed to a different type of o-ring o 1996 -2014 fewer leaks occurred wonly 3 inner o-ring leaks in 10 years o 2015-2017 - frequency of a-ring leaks increased - 3 leaks in 2 years:

  • 3R18 in May 2015 - Unit 3 had a double owring leak and plant shutdown
  • 2R22 - Unit 2 had a single o-ring leak in March 2017 (current condition)
  • 3R19- May 2017 Unit 3 had a double a-ring leak (the current leak).
  • Licensee's immediate actions: Monitored RCS leakage trends and entered containment to visually observe RPV flange leakage patterns. Established ODMI for trigger levels and actions. IPEC is currently conducting a failure modes and affects (FMA) analysis that includes reviewing OpE and benchmarking other utilities. They appear to be an industry outlier for o-ring leakage problems.
  • Licensee's plan to address the issue during the forced outage: The IPEC FMA analysis has identified six potential causes for o-ring leaks. They narrowed this down to two likely causes:

o FME intrusion into the flange seating surfaces during decontamination operations or RPV head stud can removal - (Note: IPEC is one of only two sites that still use stud cans) o Plant operations after head set but prior to head torque that introduced water into the flange area and disrupted the Qwring seal seating For the upcoming planned outage to repair the o-ring seal (starts June 11 - planned for 11 days), IPEC will:

  • not use stud cans when installing the RPV head studs
  • revise their decontamination procedures to minimize FME intrusion, and
  • prohibit all activities that can change water level inside the RPV between the time that the RPV head is set and the time when the RPV studs are torqued.

From: Haagensen, Brian Sent: Wednesday, June 21, 2017 1041 AM To: Haagensen, Brian; Henrion, Mad<: Rich, S,irt1h; Setzer, Thomas; Siwy, Andrnw Cc: Pelton, David; Guzman, Richurd

Subject:

0-ring leak assessment 6/21 at 1030 IP3 experienced a leak through the inner o-ring on the RPV flange seal this morning at 0102. They had reached NOP/NOT at 2 115 last night.

The IPRO had reviewed the corrective actions associated with the o-ring leakage. We had the following observations:

  • There was no obvious cause for the o-ring leakage wl1en the RPV head was removed. The boric acid deposits obscured any obvious indications of o-ring failure or FME intrusion during the initial as-found inspection. They did see small black deposits that could have been related to FME in the boric acid deposits.
  • After cleaning off the boric acid deposits, IPEC determined that the probable cause of the leakage was that they had allowed RCS waler level to wet the o-ring surfaces during the time between landing the head and tensioning the studs. IPEC thought that the water could have either dislodged the o-ring seating surfaces or caused FME to be transported into the flange seating surfaces. Removal of the boric acid deposits would have also removed any FME.
  • We reviewed the as-found o-ring inspection videos. We noted that some of the o-ring retaining clips were skewed (turned so that they were not in the correct position to hold the o-ring) and that it appeared that the a-ring may have been rolled (partially extruded from the flange seating surfaces). In addition, the RPV flange seating surfaces appeared to have surface pitting and other surface defects. IPEC had analyzed this degradation from prior outages and determined (by visual inspection) that the degradation had not become worse.
  • IPEC did not take measurements of the pitting on the flange seating sutiaces or conduct any depth mic measurements during this planned outage . They concluded that the pits were not of sufficient depth to require further repairs. They had previously taken measurements of these defects during the o-ring outage in 2015 and determined that they had not degraded further and did not require repairs. Theo-ring should have compressed 10 fill these defects and seal the joint.
  • IPEC completed an FMEA that identified numerous (- 10) possible causes for the o-ring leak. They identified two likely causes; 1) the o-ring became dislodged during the time when RCS water impacted the flange seating surfaces prior to tensioning due to changes in RHR flow, and 2) FME had been introduced in the flange seating surfaces. They elected to start up with the new o-ring because, although they did not determine the cause of the leakage, they thought their corrective actions would be effective because they had been effective in 2015.
  • IPEC has two new 0-rings of the same design available on-site for replacement. However, the supply chain is investigating lead times for a different design a-ring that would have thicker silver deposits to facilitate filling deeper flange surface defects. This repair option will require at least an 8 week lead time to manufacture this new a-ring.
  • IPEC is preparing a critical decision paper for review during a fleet call later today (1230) that will determine a path forward. They are likely to most likely continue the startup on the outer o-ring. Following approval of this decision, IPEC will have another fleet ca ll to approve the restart.
  • They are leaving the ODMls in place for the a-ring leakage.
  • Next update to be provided after the 1230 Fleet call.

Brian C. Haagensen Senior Resident Inspector

Indian Point Energy Center 914 739-9360 (Office)

(b)(6) (Cell) 2

L SUSTAINABLE Event Free Operations Leadership Effectiveness Learning Organization Tailgate June 8, 2017 e~------i,i--------- - - - - - w E Pow ER Lt F E'M w E pow ER LIFE"'

L Agenda fl!Entergy_

  • Unit 3 0 Ring Update
  • NIOS Escalation
  • NRC End of Cycle Meeting
  • Speed Bumps
  • Calendar of Events (5) I WE POWER LIFE"' 2

L Unit 3 0-ring leakage fl!Entergy_

  • Unit 3 has had 3 inner 0-ring failures and 2 outer 0

-ring failures since 2009 (2009 RFO, 2015 RFO and 2017 RFO)

  • Unit 2 currently has an inner 0-ring failure (2016 RFO)
  • IPEC formed a team to attempt to determine the possible causes and what mitigating actions that need to be put in place to ensure the highest probability of success for the upcoming planned outage to replace the Rx Head 0-rings.

(5) I WE POWER LIFE"' 3

L Unit 3 0-ring leakage fl!Entergy_

  • Categorized the potential causes into several areas
  • Potential causes that are considered unlikely
  • Design
  • Weare one of several PWR's ~(b)(4) I l<b)(4) I) that use this design in the industry with success including last cycle after we performed the mid-cycle replacement.
  • Installation Practices
  • We reviewed our procedures against others in the industry and our procedure has the most robust instructions and controls and we have incorporated all the recommendations from Technetics and Westinghouse. We also have video evidence during the installation of the RX vessel head there were no issues.

(5) I WE POWER LIFE"' 4

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L End of Cycle Meeting fl!Entergy_

  • Nuclear Regulatory Commission staff has scheduled the 2016 End of Cycle meeting to discuss safety performance of the Indian Point nuclear power plant for June 14, in Tarrytown, N.Y.
  • The meeting will begin at 7 p.m., at the Double Tree by Hilton, 455 South Broadway.
  • N RC staff responsible for inspections and oversight of the plant, including the resident inspectors based at the site full

-time, will be on hand to discuss plant performance. Doors to the meeting room will open at 6 p.m.

  • Meeting attendees will be given time to ask questions of the NRC. Names of those who sign up to speak will be selected randomly.

(5) I WE POWER LIFE"' 13

L Speed Bumps fl!Entergy_

Security to Use Temporary Speed Bumps Security will be using temporary speed bumps, similar to the one pictured below, to help reduce vehicle speed on site. These strips will be placed in the vicinity of the front gate and inspection area.

(5) I WE POWER LIFE"' 14

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From: Siwy, Andrew Sent: Friday, June 02, 2017 743 AM To: Hemion, Mark

Subject:

FW: History of o-ring leaks ;it IP3 -FIRST LOOK INFORMATION Attachments: !External .Sender] The 3-page attachment has been withheld in its entirety under FOIA exemption 4.

C/*.,.d,,e,w 5',;,., , ['S.

1 R,,sidL,nt !n.,plxlor I ludi,111 Point llq ;ion JI ~:~){4.73'J.r);,61l From: Haagensen, Brian Sent: Thursd ay, June 01, 2017 8 :32 AM To: Setzer, Thomas <Thomas.Setzer@nrc.gov>; Pelton, David <David.Pclton@nrc.gov>

Cc: Patel, Jigar <Jigar.Patell@nrc.gov>; Siwy, Andrew <Andrew.Siwy(tilnrc.gov>; Rich, Sarah <Sarah.Rich@nrc.gov>

Subject:

RE: Histo ry of o -ring leaks at IP3 -FIRST LOOK IN FORM ATION Dave, T'lrn, As r(:qt11:d:od, here is our analysis of HK: <.Hinq le:,ik hislc1ry al IPF::C. _(Seethe attached photo of Samh':3 o-rin9 leak time lino).

This f.1;forn.ationJuretimi11c11y and subit~d ton~vision and valirlaricm ..

The inifa:11 lrends ,'ire:

" 1986 -- 19H9 the1*e were ct clusler of 4 o-ring leaks split b<::tween both unit:, in th0 lalo 19H(h; over a 4 year period.

() :3 leak s on Unit 2 includin~J one dout>le o-ring leak r:) *1 kiak on Unil '.1

., ~**199(:i double (Hinq leak on U11it 3.

<JI t he o,.. ring des ign was the,n c hanged to a d ifforent tyr~ie of o-d :l 91

.. '199G .... 20 *15 There worn fewer leak:, (only :; -- f,in9k) o-rinq le-:tk:; '.,plit belweon the units over *-10 yecmc:)

o 2003 Unit 2 ~;in[~ltJ o-,*ing leak u 2004 Unit 3 ~,ingle o-ring leak o 20'!0 Unit 3 single o-rinr,i leak

  • 2015*-2017 **** frequency of o-r*in£l leaks increas(~d . . :-3 le;;ik s in ;2 year:3 o 3H"l8 in May 2015
  • Unit 3 had another doublG o-ri119 l(Jak ,rnd ll~d to ~:hu!down in SEJplmnber for re;xiirn.

o :.:'.F~22 *- Unit 2 had a singln o-ring leak in l'vb rch 20*17 ... shiflEid to the oul~:r o-ring seal and continued op81*ation * - curront condilion o 31~HJ - May 2017 Unit 3 had a doub!e cHi11g le;::.I~ (the currnnt leal,;). R<0)pair plan being formu lated by IPEC .

We will liavo further analy::;is as w8 continue our i11spection efforts. We have started thl=J Pl&n sample 011 Fie C,;;H .ISe~; and corrod.ive sictions for O**rin9 lea l<s tha t w a~1 sr;ecified st the COC ror this quart er.

Brian C. Hfwgonser1 Senior n esic1011t 11*1spector

Indian Point Energy Center 914 739--9360 (Office)

(b)(6) (Celi) .

From: Setzer, Thomas Sent: Thursday, June 01, 2017 7:31 AM To: Haagensen, Brian <.5.rian.Ha§f,&!l.Sen(a)nrc.:&QY>

Subject:

FW: orings at IP3 - inspections Can u pis help get this?

From: Pelton, David Sent: Thursday, June 01, 2017 7:31 AM To: Setzer, Thomas <Thornas.Setzer@n_~_,gQY>

Subject:

RE: orings at IP3 - inspections

Tom, Enforcement aside, can we get a history/summary of RV head seal leakage (by unit)? Front Office request Thanks.

dave p.

2

CR-AN0-1-2015-03240 11 of 169 CRG Agenda Meeting Date : 9/8/2015 Originator: Crosby.Patrick Phone: 4903 Site - Group: ANO - Eng Code Programs Staff ANO Discovered: 8/31/2015 15:50:48 Supe:rvlsor: Greeson,William C Initiated : 8/31/201516:01:55 Condition Desc:

CR-AN0-1-2015-02179 CA-15 requires the Boric Acid Corrosion Control Program (BACCP) to perform a Boric Acid Evaluation for the U1 RVCH inner 0 -ring leak documented in CR-AN0-1-2015-02967. The evaluation is required to be documented in an EC (EVAL), sub-type (SOR) IAW EN-DC-115 to evaluate acceptability of the leak until the next refueling outage (1 R26 - Fall 2016).

During the evaluation process it was determined that unpredictable steam cutting and impingement at the 0 -ring and reactor vessel flange area would neither allow for an acceptable evaluation nor provide justification for extended operation without mitigating actions. Per EN-DC-319 (BACCP), periodic monitoring of the leak is required to mitigate and determine the condition of the stainless steel cladding of the reactor vessel and closure head.

The carbon steel reactor vessel and closure head are cladded internally with stainless steel. Stainless Steel is resistant to boric acid corrosion but is not immune to steam cutting or impingement. Per industry tests documented in EPRI Report No.

1000975, steam cutting and impingement could degrade the cladding and expose carbon steel.

Based on the last visual inspection performed on 8/4/2015 and the lack of discoloration of boric acid crystals, the steel cladding is intact and performing its intended function.

The ODMI and Critical Decision document for CR-AN0-1-2015-021 79 reference SER 3-09, which is industry OE regarding 0 -ring leakage at Browns Ferry Unit 1 in 2008. This OE documents that Browns Ferry Unit 1 experienced 0 -ring leakage for 18 months across both the inner and outer 0 -rings, which resulted in an increase in unidentified Drywell leakage from O to 1.2 gpm over the course of the 18 month operating cycle. Following removal of the RV head, damage to both the RV and RVCH seating surfaces from steam cutting was observed. Regarding SER 3-09 applicability to AN0-1, it is noted that Browns Ferry Unit 1 is a BWR, which utilized non-borated water in the RCS and operates at a nominal RCS pressure of 1000 psig . As such, the observed damage at Browns Ferry Unit 1 resulted from a 1000 psig delta-P across two 0-rings, whereas AN0-1 is currently experiencing a 2155 psig delta-P across one 0 -ring. As such, degradation of the AN0-1 RV and RVCH flange 0-ring seating surfaces could accelerate at a much faster rate than the Browns Ferry Unit 1 scenario.

Immediate Action:

None Suggested Action Perform a containment entry every 30 days to verify the stainless steel cladding is intact by observing the drain piping for discoloration, such as, red or brown boric acid crystals indicative of carbon steel corrosion. The last visual observation was performed on 8/4/2015. Perform next inspection during the same power entry when adding oil to RCPs scheduled 9/15/15.

Consider updating current ODMI to take action upon discovery of discoloration in boric acid crystals.

Consider updating the current operabil ity to OP-DNC or OP-Comp Meas. with monitoring as the compensatory measure.

Flagged for SM: Y Flagged for Reportability Review: Y Assignments Category: C CR Owner:ANO - Eng Design & Programs Mgmt ANO - Butler. Paul Wayne Classification: ADV CORRECT Comments:

Butler: (RM Action Due Date 09/24/15)

Bring Back Department/Date:

Report Date: 09/03/201 5

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  • CARTER RODNEY A: WOODSON TIMOTHY R: .P.e.Lkin.s..

Keith; Harvey l ohn D; OLSON DONALD G; STEWART. IOSEPH; PARDI CHARLES A; FLOYD. VICTORIA B; Couch. Cheryl: SEITER. IEFFERY ALAN; Tobin. Margaret; TESSIER. ROBERT: HENDERSON. DOUGLAS: THINGER *

.fil'..RQN; Rhodes Kristie: SHORT BRADLEY W: SIG LE !ODY L: STALNAKER DALE EDWARD: McGee Francis; Epp erson Paula: HILL IESS!CA; Barrett Andy; HEFLIN IACK! E L; REMER CHARLES A" AN O Training Instructers; HELMS. STEVE: Sotomayor. Amy: '--'FR=I'X=:

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Subject:

[External_SenderJ 09110115 ANO CRG Report Note: Of two pages of the attached 401-page report are Date: Wednesday, September 09, 2015 3:14:55 PM responsive records.

Attachments: At.JO_CR G_Repru:t. pdf' Report -Attached Stephen Whitley Performance Improvement I CAA Adm in Arkansas Nuclear One External: 479-858-4082

CRG CR Summary CRG Agenda Meeting Date: 9/10/2015 Owner Oper CR Number Current Significance Site Owner Group Flag Assignment Description CR-AN0-1-2015-03210 B-ADV E-ACE CARB ANO Projects N Pace to Palmer: (E-ACE Action Construction-Supp Due Date 10101115)

Mgmt CRG Comments: TAKE BACK: CR-AN0-1-2015-3210: PCS requests CRG review/approve CRG Briefing Sheet (Attached at the end of this report)

Non Responsive Record 9/9/2015 14:54 Page 1 of 12

CR-AN0-1-2015-03210 1 of 327 CRG Agenda Meeting Date: 9/10/2015 Originator: Beaird.Robert K Phone: 3167 Site

  • Group: ANO - QA Audits Staff ANO Take Back Discovered: 8/27/2015 16:10:59 Supe:rvlsor: Blocker 111,Lenard Initiated: 8/ 27/2015 16:23:25 Condition Desc:

Problem Statement:

CR-AN0-1-2015-2967 documenting the confirmation of the Unit 1 Reactor flange inner o-ring leakage was classified by the CRG as a category "C" ADV correct CR.

Potential Consequences:

Potential outage delays or re-work Details:

During the review of site activities surrounding the potential Unit Reactor Flange leak ( CR-AN0-1-2015-2179) NIOS identified that CR-1-2015-2967(which confirms the leak) had been downgraded from Cat B E-CARB to Cat C Correct/Address by the 8/18/15 CRG. The Cat CCR classification will not ensure causal analysis is performed prior to any required shutdown to ensure issues such as poor installation practices or defective parts are identified and corrected prior to the installation of a new o-ring. This leaves the site vulnerable to outage delays or repeated leaks. No other CR was found that addresses the cause of the o-ri ng leakage.

Immediate Action:

Discussed issue with CAP/OE Manager Suggested Action Flagged for SM: N Flagged for Reportability Review: N Assignments Category: B CR Owner: ANO - Projects Construction-Supp Mgmt ANO -

Classification: ADV E-ACE CARB Comments:

Pace to Palmer: (E-ACE Action Due Date 10/01 /15)

Bring Back Department/Date:

TAKE BACK: CR-AN0-1-2015-3210: PCS requests CRG review/approve CRG Briefing Sheet (Attached at the end of this report)

Reference Items Type Description QA-ASSESSMENT CR QA-FA-LEADERSHIP Other Trends Type Code Code Description REPORT WEIGHT RW02 Oversight-identified WGMR CRG Condition Review Group WGMR PP10 Corrective Action Program.

WGMR EX10 Administrative Requirement not Met Report Date: 09/09/2015

From: SftORT, BRADLEY W To: Tobin Margaret: Barrett Andy: Tindell Brian

Subject:

[External_Sender) FW: AN0-1 RV o-ring leakage* 10/23/2015 Power Entry Date: Friday. October 30. 201 S 2:46:02 PM Folks Please look at t he response to CR-AN0-1-2015-2 179 CA 23. Let me know if t his addresses your q uestions or if there is something else you need . Thanks From: BARBOREK, W DOUGLAS Sent: Tuesday, October 27, 2015 2:41 PM To: margaret, tobin Cc: Brian.Tindell@nrc.gov; Andy.Barrett@nrc.gov; SHORT, BRADLEY W; WOODSON, TIMOTHY R; Conyers, Daniel

Subject:

RE: AN0-1 RV o-ring leakage

  • 10/23/2015 Power Entry Maggie, We have discussed t his internally and briefly w it h AREVA and have no definitive explanation fo r the leakage to plateau as it has since lat e July. Our issue has certainly exhibited different behavior than the Indian Point-3 leak and other previous o-ring leaks in the industry. It seems very likely th at o ur o-ring leakage has a different cause than the IP3 issue.

Theo-ring leakage temperature alarm is still locked in on t he Control Room annunciator, and t he leakage is still emanating from the header near t he sump, so we still believe the leakage is from the RV o-ring. The only confirming piece of information we don't have is thermography of the leak-off drain piping which we cannot obtain at power due to the location of the piping. That will be a piece of information we plan to obt ain if we have a planned or unplanned RX trip which allows general access to the RB basement.

During our next power entry in December, I have informed RP t hat my intent is to take another walk around elevation 354' to look at the basement floor (elevation 335') t hrough the grating and check for signs of any other leakage, as well as checking the temperature of the RCP intergasket leak-off lines aga in to ensure t hey are at ambient temperature and not leaking by. The last entry in which we performed this effort was t he July entry.

Please let me know if t his d id not answer your questions or if you have any additional quest ions.

Thanks, Doug Barborek Entergy Operations, Inc./ Arkansas Nuclear One System Engmeer- AN0-1 & AN02 Reactor Coolant Systems and ANO l Spent Fuel Cooling & Punficatoon System System Engineering Building/ N*SYE*4 wbarbol@eotersv rom 479-858-4337 (office)

!(b)(6)  !(pager)

From: Tobin, Margaret [mailto*Margaret Tobjn@nrc gov]

Sent: Tuesday, October 27, 2015 12:48 PM To: BARBOREK, W DOUGLAS Cc: 6ciao Iiodell@occ gov; Andy Barrett@nrc gov

Subject:

RE: AN0-1 RVo-ring leakage

  • 10/23/2015 Power Entry Doug, I have a follow-on question. I'm curious if you have a theory on why t he leak rate seems to have more or less settled out? Everything I heard about t his issue from t he start was that Op-E from other sites suggested t his leak should slowly get worse until you hit a trigger point and needed to shut down.

I'm concerned that somehow we may be missing some important piece of information somewhere because it isn't following t he expect ed physical phenomenon.

Thanks, Maggie From: BAR BOREK, w DOUG LAS [majltoWBARBOl@enteri;y com]

Sent: Tuesday, October 27, 2015 10:33 AM To: Barrett, Andy <Andy Barrett@nrc gov>

Cc: Tobin, M argaret <Margaret Tobjn@nrc gov>; Tindell, Brian <Brian Iiodell@orc gov>; SHORT, BRADLEY w

<bshort@entergy com>; WOODSON, TIMOTHY R <TWOODSO@entergy com>; Conyers, Daniel

<dconyer@entergy com>; EDGELL, DOUGLAS w <DEDGELL@entergy com>

Subject:

(Externa l_Sender] AN0-1 RV o-ring leakage - 10/23/2015 Power Entry

Andy, Regarding your questions on t he Unit 1 power entry last Friday.

The boric acid at the ta il pipe is still pristine white, indicating no wastage of carbon steel. The leak rate w as visually t ypical of previous entries. The rate of increase of the leak slowed down in late July, and is now on a very slow increasing t rend . Below is a snapshot fro m t he Shift Engineer's spreadsheet for RCS leakage which shows the Cycle 26 data. Also attached are some still shots from the video I took. You are more t han welcome to stop by my cube and w at ch t he entire video.

Please let me kn ow if you have any addition questions concerning this issue.

Thanks, Doug Barborek Entergy Operations. Inc. / Arkansas Nuclear One System Engineer - AN0 -1 & AN0 -2 Reactor Coolant Systems and AN0-1 Spent Fuel Cooling & Purification System System Engineering Building/ N-SYE-4 wbarbol@eotergy com

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From: SHORT, BRADLEY W To: Jobin Margaret

Subject:

[Externa l_Sender] RE : ODMI for RCS leakage Unit 1 Date: Wed nesday, August 19, 2015 7:36:17 AM Attachments: Reactor Vessel Flange Inner 0 -ring Leakage.ptm<,

See attached BRAD SHORT UCENS1NG. RE.6ULATORY S PPORT PHO £. 479-858-3271 Ct:LL j(b)(6) I From: Tobin, Margaret rgaret.Tobin@nrc.gov

Sent: Wednesday, August 19, 2015 7:24 AM To: SHORT, BRADLEY W

Subject:

ODMI for RCS leakage Unit 1 Good Morning Brad ,

Can I get a copy of the most updated ODMI for RCS leakage in Unit 1? It was updated several weeks ago, according to the control room .

Thanks, Maggie

Reactor Vessel Flange Inner 0-ring Leakage Trigger Point 1: More than three consecutive daily Total RCS leakrate determination exceeds 0.180 gpm while on Green Train, or 0.200 gpm while on Red Train .

Actions:

1. Notify Operations Management (notification is not required each night after the initial notification).
2. Revisit Critical Decision document.
3. Consider additional RB Power Entry at reduced power or with reactor off-line.
4. Commence OP-1103.013 Attachment A "RCS Leakage Investigation" to determine if there are other sources of leakage.

Reactor Vessel Flange Inner 0-ring Leakage (Cont.)

Trigger Point 2: The 5 Day Average Unidentified leakrate exceeds 0.090 gpm.

Actions:

1. Notify Operations Management (notification is not required each night after the initial notification).
2. Revisit Critical Decision document.
3. Consider additional RB Power Entry at reduced power or with reactor off-line.
4. Commence OP-1103.013 Attachment A "RCS Leakage Investigation" to determine if there are other sources of leakage.

Reactor Vessel Flange Inner 0-ring Leakage (Cont.)

Trigger Point 3: The 5 Day Average RB Sump leakrate exceeds 0.060 gpm.

Actions:

1. Notify Operations Management (notification is not required each night after the initial notification).
2. Revisit Critical Decision document.
3. Consider additional RB Power Entry at reduced power or with reactor off-line.
4. Commence OP-1103.013 Attachment A "RCS Leakage Investigation" to determine if there are other sources of leakage.

Reactor Vessel Flange Inner 0-ring Leakage (Cont.)

Trigger Point 4: Difference between current 5 Day Average Unidentified Leakrate and the value from 30 days prior exceeds 0.036 gpm .

Actions:

1. Notify Operations Management (notification is not required each night after the initial notification).
2. Revisit Critical Decision document.
3. Consider additional RB Power Entry at reduced power or with reactor off-line.
4. Commence OP-1103.013 Attachment A "RCS Leakage Investigation" to determine if there are other sources of leakage.

Reactor Vessel Flange Inner 0-ring Leakage (Cont.)

Trigger Point 5: RCS Total Leakage > 0.25 gpm Actions:

1. Notify Operations Management.
2. Convene Bridge Call to determine the best time to take the Unit off-line.
3. At the agreed upon time to take the Unit offline, commence Plant Shutdown utilizing OP-1102.016 "Power Reduction and Plant Shutdown".
4. Following reactor shutdown, perform immediate RB Entry to validate source(s) of RCS leakage and to perform thermography of RV flange leak-off piping at "A" Cold Leg opening in the primary shield wall to verify TE-1052 readings of 210 F. This will fully confirm the presence of RV flange inner 0-ring leakage.

Branch 2 I BRIDGE: 9-1-866-751-2529, passcodel(b)(6) I# I 01/ 16/2018 Dally Status I New Info In bold DRP Status Board Items [Dlac:uued at Morning MNllng I Proiect Manaaers I IP/MS: R. Guzman I I Branch Items Non Responsive Record INDIAN POINT UNIT 2 I Rx Power: 100% I Risk: Green I Unidentified Leakage: 0.05 gpm I AL: 1 Weekend Cover.age: Brian I RFO: 3/18 - 518118 I AL1: AL2: AL3:

Unplanned/Significant TSAs 12/3 Swap essential and non-essential SW header: 8-hr LCO 12/3 Monthly test of 23 EOG I Major Activities Onsite None Other Items: (Security, EP, Stakeholder Interest)

Union contract expiration date is 1/ 18/18.

. 21 RCP seal maintenance outage being planned for 1/4/18. Expected to last 8-14 days.

Soent fU1el camoaian in oroaress. Movina one drv cask to ISFSI 11/9-1 1/22 .

I Rx Power: 100% I Risk: Yellow I Unidentified Leakage: 0.01 gpm I AL: N/A INDIAN POINT UNIT 3 Weekend Coverage: Brian I 3R20: 3111 - 511119 I AL1 : AL2: AL3:

Non Responsive Record

MILLSTONE UNIT 2 I Rx Power: 100% I Risk: Green I Unidentified Leakage: 0.121 gpm IAL: 0 Non Responsive Record MILLSTONE UNIT 3 I Rx Power: 100% I Risk: Green I Unidentified Leakage: 0.124 gpm IAL: N/A Non Responsive Record

From: Tobin, Margaret Sent: Friday, January 05, 2018 9:20 AM To: Fredette, Thomas

Subject:

FW: AN0-1 Reactor Vessel Flange Inner 0 - ring Leakage - Video from 7/22/2016 Power Entry Attachments: Comparison 6 entries.pdf From: BARBOREK, W DOUGLAS [1]

Sent: Friday, July 22, 2016 11:16 AM To: Tindell, Brian <Brian.Tindell@nrc.gov>

Cc: Tobin, Margaret <Margaret.Tobin@nrc.gov>; Barrett, Andy <Andy.Barrett@nrc.gov>; PYLE, STEPHIENIE L

<SPYLE@entergy.com>; WOODSON, TIMOTHY R <TWOODSO@entergy .com>; SKARTVEDT, MARK EDM UND

<M SKARTV@entergy.com>

Subject:

[External_Sender) AN0-1 Reactor Vessel Flange Inner 0 -ring Leakage - Video from 7/22/2016 Power Entry Brian, For youir information .

The (RV inner o-ring) leakage at the HSD-15-2" header was observed to be approximately 48 drops/minute during this morning's power entry. Boric acid deposits remain white with no discoloration. Initiated CR-AN0-1-2016-02183 to document observations.

A still shot from the videos from the past 6 power entries is attached. Cycle 26 RCS/Sump leak rate info (as of today) is provided below.

Please llet me know if you have any questions regarding today's entry observations. If you would like to see the video from today's entry, please feel free to stop by my cube.

Thanks, Doug B.arborek Entergy Operations, Inc./ Arkansas Nuclear One System Engineer - AN0-1 Reactor Coolant System and AN0-1 Spent Fuel Cooling & Purificat ion System System Engineering Building/ N-SYE-4 wbarbol@entergy.com l(b)(6) I 479-858-4337 ~ ce)

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August 4th, 2015 September 15th, 2015 October 23rd, 2015 December 2nd, 2015 January 18th, 2016 July 22nd, 2016 OFFICIAL USE ONL'f SENSITIVE PRE-DECISIONAL INFORMATION VISIT BRIEFING PACKAGE Stephen G. Burns NRC Chairman United $ra res N uclear Regularory Co mmission Protecting People and the Environment INDIAN POINT NUCLEAR GENERATING UNIT NOS. 1, 2, & 3 ENTERGY NUCLEAR NORTHEAST ENTERGY NUCLEAR OPERATIONS, INC.

December 8, 2015 ADAMS Accession No. ML15334A442 OFFICIAL USE ONLY SENSITIVE PRE DECISIONAL INFORMATION 1

OFFICIAL USE ONLY-SENSITIVE INTERNAL INFORMATION Index for Tabs TAB I TOPIC OF INTEREST I PAGE NUMBER Non Responsive Record TAB6 I Current Issues I 15 - 21 Non Responsive Record 8FFICIAL USE 8NLY SENSITIVE PRE-DECISIONAL INFORMATION 2

OFFICIAL USE ONLY* SENSITIVE INTERNAL INFORMATION II CONTENTS II Indian Point Energy Center (IPEC)

Non Responsive Record Current Issues .............................................................. Tab 6 Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE DECISIONAL INFORMATION 3

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record

OFFICIAL USE ONLY* SENSITIVE INTERNAL INFORMATION December 8 1 2015 8:00 a.m. Chairman arrives at IPEC main gate 8:00 a.m. - 8:10 a.m. Present NRG badges to security at main gate and proceed to parking lot in front of Generation Support Building (GSB) 8:10 a.m. - 8:20 a.m. Meet Entergy personnel at MAC 4 entrance for access 8:20 a.m. - 9:00 a.m. Meet with resident inspectors in GSB conference room 132 9:00 a.m. - 9:10 a.m. Introductions with congressional staffers 9:10 a.m. - 9:20 a.m. Travel to Site VP conference room (Coffee, Danishes and Water provided) 9:20 a.m. - 10:20 a.m. Entergy Presentation Site History/Performance Fukushima Transformers Performance 10:20 a.m. -12:20 p.m. Plant tour Tower 12 BRE U2 Control Room U2 Turbine Central Alarm Station (GAS)

U3 Turbine 480V vital switchgear room U3 Emergency Diesel Generators (EDGs)

U3 Transformer yard 12:20 p.m. - 12:50 p.m. Lunch 12:50 p.m. - 1:20 p.m. Meeting with Tim Mitchell 1:20 p.m. - 2:00 p.m. Tour of FLEX building 2:00 p.m. Depart site Contact Information:

David Lew, Deputv Regional Administrator, Region I Cell:l(b)(6) ]

Brian Haagensen, Senior Resident Inspector Cell : l(b)(6) ]

Office: 914-739-9360 OFFICIAL USE ONLY - SENSITIVE PRE-DECISIONAL INFORMATION 5

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record 6

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE DECISIONAL INFORMATION 7

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE DECISIONAL INFORMATION 8

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE DECISIONAL INFORMATION 9

OFFICIAL USE ONLY-SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY - SENSITIVE PRE-DECISIONAL INFORMATION 10

OFFICIAL USE ONLY SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY - SENSITIVE PRE DECISIONAL INFORMATION 11

OFFICIAL USE ONLY* SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY - SENSITIVE PRE DECISIONAL INFORMATIObl 12

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE-DECISIONAL INFORMATION 13

OFFICIAL USE ONLY SENSI \/

Non Responsive Record

  • T l ~ E INTERNAL INFORMATION_

OFFICIAL USE ONLY SE NSITIVE PRE DECISIONAL INFORMATION 14

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE*DEGISIONAL INFORMATION 15

OFFICIAL USE ONLY-SENSITIVE INTERNAL INFORMATION TAB6 Current Issues Indian Point Nuclear Generating Unit Nos. 1, 2, and 3 A. EXPECTED DISCUSSION TOPICS Recent Licensee Performance Overall, Indian Point Units 2 and 3 operated in a manner that preserved public health and safety and met all cornerstone objectives. Performance at Indian Point Units 2 and 3 during the most recent quarter was within the Licensee Response Column of the NRC's ROP Action Matrix because all inspection findings had very low (i.e., Green) safety significance and all Pis indicated that performance was within the nominal, expected range (i.e., Green).

On May 9 2015, the IP3 31 main transformer experienced a failure, resulting in a fire and flooding from the transformer deluge system in the 480V safety-related switchgear room. As a result of the flooding, a Special Inspection Team (SIT) was launched on May 19, 2015 in order to review the circumstances surrounding the water intrusion into the switchgear room. The SIT report (2015010) documented one NRG-identified, Green NCV, for "failure to promptly identify, report, and correct a condition adverse to fire protection." The SIT did not review the cause of 31 main transformer failure, which is scheduled to be reviewed when Entergy concludes their Root Cause Evaluation (currently scheduled for 1Q2016).

History of Transformer Issues On May 9, 2015 IP3 experienced a main turbine-generator lockout, main turbine trip, and automatic reactor trip as a result of an explosion and fire on the 31 Main Transformer. The site fire brigade was activated and called for offsite resources . A Notice Of Unusual Event was declared due to an explosion in the Protected Area. The fire was initially extinguished at 6:15 p.m.; however, it reflashed at 6:37 p.m., at which time additional offsite resources were brought into the Protected Area. The fire brigade continued to fight the fire with foam and declared the fire extinguished at 8:05 p.m. During the event, there were reports of water in accumulation in the 480V switchgear room, totaling approximately 1 to 2 inches in depth. It was later determined that, based on the leak rate, the amount of water in the switchgear room could not have exceeded 0.4 inches. The source of the water was determined to be deluge system water that had come from a deluge valve room that sits adjacent to the 480V switchgear room. The water was unable to completely drain through a floor drain due to its limited capacity, resulting in it making its way into the 480V switchgear room.

In addition to the 31 main transformer failure on May 91h, there have been multiple previous transformer issues at IP2 and IP3. Since 2007, Indian Point has had the following transformer issues:

Indian Point Unit 3 - #31 Main Transformer explosion on April 6, 2007, due to an electrical fault in the 'B' phase high voltage bushing Indian Point Unit 2 - #21 Main Transformer explosion on November 7, 2010 OEEICIAI I ISE ONLY - SENSITIVE PRE DECISIONAL INFORMATION 16

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Indian Point Unit 3 - Unit Auxiliary Transformer, removed from service on February 29, 2012, due to high gassing The NRC's ROP inspection program includes inspections that review the impact that non-safety related systems, such as main transformers, have on the plant. This includes inspections of Problem Identification and Resolution (Pl&R), Maintenance Rule, and Performance Indicators.

The Resident Inspectors completed an Event-Follow-up inspection of the 31 main transformer following the failure, and the results were documented in the publically available quarterly report (2015002). Additionally, the NRC completed a Maintenance Rule inspection of the 31 main transformer, the results of which were documented in the publically available 3rct quarter report (2015003). A Pl&R inspection sample of the transformer issues since 2007 is scheduled to be completed once Entergy completes their Root Cause Evaluation, and is expected to be completed in the 1st quarter of 2016. The results will be published in publically available inspection reports. These inspections will provide insights into Entergy's transformer maintenance and monitoring programs, and will review the causes of the various transformer issues.

Status of Post-Fukushima Actions The post-Fukushima enhancements associated with EA-12-049, Mitigating Strategies for Beyond-Design-Basis External Events (FLEX), and EA-12-051 , Spent Fuel Pool Instrumentation, are ongoing at IPEC. The onsite audit for EA-12-049 and EA-12-051 was completed on October 27, 2014, and IP3 achieved compliance with the orders in the Spring of 2015. The onsite audit for IP2 regarding EA-12-049 is currently scheduled for the week of November 30.

License Renewal By letter dated April 23, 2007 (Agencywide Documents and Management System (ADAMS)

Accession Number ML071210512), Entergy applied for renewal of the Indian Point Units 2 and 3 operating licenses for an additional 20 years beyond the current expiration dates. The current expiration date of the operating license for Unit No. 2 was midnight on September 28, 2013, and the current expiration date of the operating license for Unit No. 3 is midnight on December 12, 2015.

The NRC staff conducted a safety review of the license renewal application and documented its findings in NUREG-1930, "Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3," (ADAMS Accession Nos. ML093170451 and ML093170671) issued in November 2009. On August 30, 2011 , the staff issued Supplement 1 to NUREG-1930 (ADAMS Accession No. ML11242A215) and on July 7, 2015, the staff issued Supplement 2 to NUREG-1930 (ADAMS Accession No. ML15188A383).

The NRC staff also conducted an environmental review of the license renewal application and documented its findings in NUREG-1437, Supplement 38, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Indian Point Nuclear Generating Unit Nos. 2 and 3" (ADAMS Accession Nos. ML103350438, ML103360209, and ML103360212) issued in December 2010. On June 20, 2013, the staff issued NUREG 1437, Supplement 38, Volume 4 Final Report (ADAMS Accession No. ML13162A616. The staff expects to issue a draft FSEIS Supplement in Jlanuary 2016 and the final FSEIS Supplement in September 2016.

OFFICIAL USE ONLY SENSITIVE PRE-DECISIONAL INFORMATION 17

OFFICIAL USE ONLY- SENSI I IVE IN I ERNAL INFORMATION It is possible that the draft and/or final FSEIS supplements may trigger the filing of new environmental contentions.

There are three unresolved safety contentions at present; additional contentions may be filed.

Eleven contentions have been resolved: one was settled, one was resolved in favor of New York, and nine were resolved in favor of the Entergy and the Staff. Appeals from the Board's resolution of two contentions are pending before the Commission. Hearings on the three remaining safety contentions were held in Tarrytown, NY, on November 16-20, 2015.

Entergy is pursuing resolution of issues related to the plant's Coastal Zone Management Act (CZMA) certification, Clean Water Act section 401 certification, and renewal of the plant's State Pollution Discharge Elimination System permit through the relevant New York State adjudicatory processes. The NRG staff is not directly involved in these processes. If, Entergy does not obtain the requisite certifications prior to completion of the agency's review of the license renewal application, the NRC may need to consider whether renewed licenses may be issued prior to completion of the State's adjudicatory processes.

State/Local Government Interest Indian Point generates significant interest amongst the State and Local government representatives. State and County agencies are highly engaged on all activities related to Indian Point. The licensee has an extensive outreach program with these offsite representatives and have developed a positive relationship with the four surrounding counties emergency organizations.

Notwithstanding the above, both the Governor of New York and the Attorney General's office are on record regarding their desire that the Indian Point plants cease operation. Following the Fukushima event, the Governor's office cited additional concerns in support of their position.

The three main topics were (1) the capability of evacuating New York City should a similar event occur, (2) the adequacy of a 10 mile EPZ (versus 50) and (3) the plant being located near a seismic fault. Most recently, the Governor's Office sent a letter to the NRG Commission dated November 16, 2015 opposing the relicensing of Indian Point. The letter can be found at:

https://www.governor.ny.gov/sites/governor.ny.gov/files/atoms/files/Entergy Letter.pdf The most controversial issue at Indian Point for the past 2 years has been the Algonquin Incremental Market (AIM) pipeline project. This project involves the installation of a 42-inch diameter natural gas pipeline in close proximity to Indian Point. There are existing natural gas pipelines that run even closer to the site that pre-dates the construction of Indian Point. Both the new pipeline and the existing pipelines have received rigorous review by NRG staff.

Entergy's site hazards analysis regarding the new pipeline was reviewed in 2014 with no safety issues identified. Questions regarding the existing pipeline are still being answered via an open allegation. Paul Blanch has reviewed the pipeline for the Town of Cortlandt and been in communication with elected officials voicing his concerns over the pipeline. We have heard from Assemblywoman Galef's Office that Paul Blanch's concerns are their concerns.

OFFICIAL USE ONLY SENSITIVE PRE DECISIONAL INFORMATION 18

OFFICIAL USE ONLY-SENSITIVE INTERNAL INFORMATION B. OTHER TOPICS OF INTEREST Emergency Preparedness Since the terrorist attacks of September 11 , 2001, significant stakeholder attention has been focused on security and emergency preparedness (EP) at Indian Point. In January 2003, James Lee Witt Associates, LLC, under contract from the State, issued a report critical of EP at Indian Point. Based on Federal Emergency Management Agency's (FEMA's) offsite and NRC's on-site EP assessments, the NRC's overall determination continues to be that EP at Indian Point is satisfactory and provides reasonable assurance of adequate protection of public health and safety. The last biennial full participation EP exercise at Indian Point was conducted on October 7, 2014. The next biennial EP exercise is scheduled for June 2016.

Additionally, since the events at Fukushima Daiichi Nuclear Station and the decision to recommend the evacuation of US citizens out to 50 miles from the facility, there has been renewed and significant stakeholder interest in emergency preparedness and emergency planning zones at Indian Point.

Open Allegations There are currently two open allegations at Indian Point:

  • Rl-2015-A-0074 involves a concern that Entergy did not adequately evaluate the effects of heat flux on SSCs around IPEC in the event of a ru ture of an existin as i eline
  • b 7 C there are parts in the 31 main trans armer that were not Open Investigations None Weapons Preemption The Indian Point security force currently uses weapons and large-capacity magazines that are banned by state and local laws. In October 2015, the NRR staff published a draft Environmental Assessment in the Federal Register that included a 30-day public comment period. NRR staff plans to issue a confirmatory order, conforming amendment, and supporting safety evaluation that will permit security personnel at Indian Point to transfer, receive, possess, transport, import, and use certain firearms and large capacity ammunition feeding devices not previously permitted to be owned or possessed under Commission authority, notwithstanding certain local, state, or federal firearms laws, including regulations that prohibit such actions.

These actions are anticipated to be complete by the end of this calendar year 2.206 Petitions Paul Blanch 2.206 Petition By letter dated October 15, 2014, Paul Blanch submitted a 2.206 petition critical of Entergy's OFFICIAL USE ONLY SENSITIVEi PREi DECISIONAL INFORMATION 19

OFFICIAL USE ONLY SENSITIVE INTERNAL INFORMATION 50.59 site hazards analysis report associated with the proposed Spectra Energy 42-inch diameter natural gas pipeline. Mr. Blanch asserted that Entergy's 50.59 analysis was inaccurate and incomplete resulting in violations of 10 CFR 50.59, "Changes, tests, and experiments, 10 CFR 50.9, "Completeness and Accuracy of Information," 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," and possibly 10 CFR 50.5, "Deliberate Misconduct." The petition, which gathered significant local stakeholder and political interest, was rejected by an NRC Petition Review Board by letter dated September 9, 2015 (ADAMS ML15251A023).

By letter dated November 6, 2015, NRR issued a response to 39 questions that Paul Blanch had presented during his July 15, 2015, presentation before the PRB regarding both the proposed and existing gas pipelines at IPEC.))

IP3 Reactor Vessel Head Seal Leakage/Repair Following the IPEC Unit 3 Spring 2015 3R18 refueling outage, leakage was identified past the inner o-ring reactor vessel head closure flange. In order to isolate the inner seal leakage, the leak off line was secured and the outer o-ring was placed in service allowing start-up to continue. On, July 14th, a high temperature leak off line alarmed and cleared intermittently. The high leak off line temperature alarm annunciated and did not clear on July 19th, indicating an outer o-ring seal failure.

Regional staff in DRS/DRP, Resident Inspectors, Nuclear Reactor Regulation staff comprised of a Senior Materials Engineer and the Senior Project Manager developed questions regarding the degraded condition and corrective action taken by the Licensee. On July 22, the licensee implemented an Operational Decision Making Issue (ODMI ) indicating different set points and associated triggers regarding the monitoring of containment temperature and humidity; telltale temp; and reactor coolant drain tank level (RCDT). In order to reduce the pressure between the inner and outer seal, both seals were placed in service.

On September 15, 2015, Unit 3 was shut down for a planned maintenance outage to replace the reactor vessel 0-rings. Following maintenance activities, the reactor became critical on September 25, 2015, and returned to full power operation on September 26, 2015.

((Gas Pipelines (Existing and Proposed)

Presently, there are 2 natural gas pipelines (26" and 30") that run through the Owner Controlled Area of IPEC, and have been the subject of great interest with various stakeholders over the past few years. Spectra Energy approached Entergy during the summer of 2013 about plans to expand their natural gas pipeline capacity across the Hudson River with a new 42-inch diameter pipeline. On February 28, 2014, Spectra filed an application with the Federal Energy Regulatory Commission for a certificate to build a new 42-inch natural gas pipeline along a southern route on Indian Point property.

In accordance with NRC requirements, Entergy performed a site hazards analysis to determine the impact of the new natural gas pipeline on the site. On August 21 , 2014, Entergy voluntarily submitted a 50.59 evaluation and blast analysis for NRC review. A Region I DRS security inspector and Headquarters expert on blast analysis performed an ROP baseline inspection (71111 .18 - Plant Modifications) of the 50.59 and blast analysis. The results of the inspection OFFICIAL USE ONLY SENSITIVE PRE-DECISIONAi INFORMATION 20

OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Non Responsive Record OFFICIAL USE ONLY SENSITIVE PRE-DECISIONAL INFORMAJIQN 21

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From: Tobin, Margaret To: farjna Thomas: OKeefe Neil Cc: Tindell. Brian

Subject:

CR on reactor vessel flange leakoff line Date: Thursday, August 06, 2015 4:15:00 PM Attachments: ~

All, Attached is a CR and corresponding documentation on this reactor vessel flange leakoff line. In short, there is a pressure test line between the inner and outer gasket that is vulnerable to corrosion induced pipe cracking due to water being left in the piping following refueling outages. The same issue may exist at the RCP inner gasket leakoff lines. These lines are normally isolated, which creates the same potential for cracking, and there may have been an NRC violation issued on these valves beinq closed years aqo.

(b)(5) 1nanKs, Maggie

Entergy CONDITION REPORT ICR-AN0-1-2008-025;

  • 1 O riginator: Barborek, W Douglas Originator Phone: 4337 Originator Group: Eng Systems NSSS StaffANO Operability Required: Y Reportability Required: Y Supervisor Name: Edgell,Douglas W Discovered Date: 12/08/2008 06:19 Initiated Date: 12/08/2008 06:22 Condition

Description:

1R2 l W0-00102463 Task OI was not able to be successfully performed as planned.

CR-AN0-1-2005-0 1140 was written to document OE from several plants which identifi ed corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/ pressure test connection piping resulting from water left in the piping following refueling outages. ANO- I was determined to be vulnerable since water has been trapped in this piping during previo us operating cycles. The scope of W0-00 I 02463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utilize a horoscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV a!"ld the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned.

Since CR-AN0-1-2005-0 11 40 was closed to W0-00 I 02463, this CR is being initiated to provide a means of tracking an alternate resolution for addressing the noted OE.

As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained during refueling outages prior to installing the RV head, tlhus mitigating the damage mechanism.

The operability statement in CR-AN0- 1-2005-01 140 remains applicable.

Immediate Action Descriptio n:

Initiated this CR.

Suggested Action

Description:

Develop new strategy for inspecting the RV flange leak-off/pressure test connection piping relative to the OE documented in CR-AN0-1-2005-0 I 140.

EQUIPMENT:

Tag Name Tag Suffix Name Compone:nt Code Process System Code NONE REFERENCE ITEMS:

l)pe Code Description CONDITION REPORT 1-2015-1950 CONDITION REPORT l-2005-1140 WORK ORDER 102463 TRENDING (For Reference Purposes Only):

Trend Type Trend Code REPORT WEIGHT l SEVERITY WEIGHT 1 HEP FACTOR p KEYWORDS KW-OPERA TrNG EXPERIENCE KEYWORDS KW-REACTOR VESSEL HEAD AJ MAPO

Entergy CONDITION REPORT ICR-AN0-1-2008-02560 Trend Type Trend Code AI ESSE AA ESSE lNPO BlNNlNG Pl2 LT-RFO/FO/SO CA-20 1R24

Entergy ADMIN I CR-AN0-1-2008-02560 Initiated Date: 12/8/2008 6:22 Owner Group : Eng Systems & Comps Mgmt ANO Current

Contact:

Current Significance: C Closed by:

Summary

Description:

I R2 I W0-00 I 02463 Task OI was not able to be successfully performed as planned.

CR-A N0- 1-2005-01140 was written to d ocument OE from several plants which identified corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/pressure test connection piping resulting from water left in the piping followi ng refueling outages. ANO- I was determined to be vulnerable since water has been trapped in this piping during previous operating cycles. The scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utilize a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV an d the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned.

Since C R-AN0 2005-01140 was closed to W0-00 102463, this CR is being initiated to provide a means of tracking an alternate resolution for addressing the noted OE.

As d ocumented in the 2005 CR, ANO procedures have been revised to ensure the subject p ip ing is drained during refueling outages prior to installing the RV head, thus mitigating the damage mechanism.

The operability statement in CR-AN0 2005-0 1140 remains applicable.

Remarks

Description:

OE C losure review required - lam 7/8/1 1 Pl Completed a Random Quality C losure Review of CA-22, in accordance with EN-LI-102 - Sat (RGT- PID) 4/2 1/20 15 Closure

Description:

Entergy OPERABILITY I CR-AN0-1-2008-02560 Ope rabilityVersion:

Operability Code: ADMlN - NA Immediate Report Code: NOT REPORTABLE Performed By: Akins,Danny W 12/08/2008 07:34 Approved By: Kinney,John W 12/08/2008 09:04 Operability

Description:

This condition report identifies the inability to successfully perform I R2 I W0-00102463 Task OI planned. As stated in the Condition Description, the scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flan ge leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utili ze a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV and the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned. Since CR-AN0- 1-2005-01140 was closed to W0-00102463, this CR is being initiated to provide a means of tracking an alternate resolution for addressing the noted OE.

As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained dm ing refueling outages prior to installing the RV head, thus mitigating the damage mechanism. Additionally, the operability statement in CR-AN0- 1-2005-01140 remains applicable.

The following excerpt was taken from the operability determination performed for CR-AN0- 1-2005-1140. "This condition report does not identify the presence ofTGSCC in the RV 0-ring Leak Detection Lines of ANO Unit I , but only the potential to develop a conducive environment. Hot shutdown walkdowns are performed at NOP/NOT conditions following each refueling. Gross leakage or external boron accumulation would be visible. Jfa leak on a detection line were to develop coincident with an Inner 0-ring leak, it would be detected as "unidentified leakage" during normal RCS Lcakrate surveillance testung. Any leakage via this mechanism would be bounded by Technical Specification 3.4. 13 (RCS Leakage) and appropriate corrective actions taken. This condition report does not specifically identify a degraded or potentially degraded plant system, structure, or component. The condition is considered to be administrative in nature. Licensing and System Engineering were contacted for concurrence with this Operability Determination."

Immediate NRC Rcportability is not required for the stated condition.

Approval Comments:

Entergy ASSIGNMENTS CR-ANO-l-2008-02560 Version :

Significance Code: C Classification Code: ADV CORRECT Owner Group: Eng Systems & Comps Mgmt ANO Performed By: zzANO CRG **lHEA use only** 12/10/2008 10:59 Assignment

Description:

(RM Action Due Date 12/23/08)

Edgell Converted from: Eng Sys Mgmt ANO

Entergy REPORTABILITY I CR-AN0-1-2008-02560 Repo rtability Version:

Report Number:

Report Code: NOT REPORTABLE Boilerplate Code: NO REPORT - ADMH Performed By : Van Buskirk,Fred P 12/09/2008 16: 10 Reportability

Description:

This condition involves administrative issues that are not reportable.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number:

G roup Name Assigned By: Eng Sys Mgmt ANO Miller Jr,John N Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborek,W Douglas Originated By: zz ANO CRG **lHEA use only** 12/10/2008 l0:59:5S Performed By: Edgell,Douglas W 12/18/2008 16:23:53 Subperformed By: Barborek,W Douglas 12/ 15/2008 11: I0:08 Approved By:

Closed By: Edgell,Douglas W 12/18/2008 17: 10:lE Current Due Date: 12/23/2008 Initial Due Date: 12/23/2008 CA Type: DTSP - CA CA Priority:

Plant Constraint: NONE CA

Description:

You have been assigned as the Responsible Manager for this Category "C", Non-Significant Condition Report by the CRG to ensure corrective actions are appropriately completed within the prescribed time frame. Develop an action plan and issue follow up action as needed.

Response

Tconcur with the fol lowing sub-response and closure of this corrective action.

Subrespo nse :

The following two CA's have been issued to further evaluate a course of action required to address the OE documented in CR-ANO- l-2005-0 l I 40. The issuance of these actions constitutes an acceptable corrective action plan for this CR.

Addit.ional CA's will be issued as deemed appropriate.

CA CR-AN0- 1-2005-0 1140 (CA-04) evaluat.ed the feasibility of perfonning a hydrostatic test of the RV flange leak-off/pressure test connection piping a:ad concluded that such a test was not feasible. Based on the inability to visually inspect the piping during IR21 (via boroscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity.

CA Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e. removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping.

WDB 12/15/2008 Closure Comments:

J concur with closure of this corrective action.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 2 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 12/17/2008 15:00: 1~

Performed By: Barborek, W Douglas 8/27/2009 10:37:49 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 8/27/200915:38:16 Current Due Date: 08/27/2009 Initial Due Date: 08/27/2009 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

CR-ANO-L-2005-01 140 (CA-04) evaluated the feas ibility of performing a hydrostatic test of the RY flange leaik-oWpressure test connection piping and concluded that such a test was not feasible. Based on the inability to visually inspect the piping during 1R21 (via horoscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity.

Response

The inspection/testing of the Reactor Vessel Flange Leak-off & Pressure Test Connection piping requires additional resources and engineering which transcends the role and responsibility of System Engineering. CA-07 has been created for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects organization. WDB 8/27/2009 Subresponse :

Closure Comments:

Tconcur with closure of this action. The scope of this inspection will required a project team given the complexities of the task, the expense involve, the potential dose involved and the required contingencies that must be in place at the time the inspection is performed.

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00002 Version: Approved: r,/

Requested Duedate: 06/04/2009 Previous Duedate: 02/ 18/2009 Requested By: Barborek,W Douglas 02/ 17/2009 Approved By: Edgell,Douglas W 02/17/2009 Request

Description:

This CA could not be completed by the assigned due date due to competing workload priorities, and recent manpower demands from forced outages 1F09-0 R, 1F09-02, and 2F09-02. Therefore, the due date is being extended to June 4, 2009.

The new date is <6 months beyond CR initiation and is therefore acceptable. WDB 2/16/2009 Approved

Description:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00002 Version: 2 Approved: r,/

Requested Duedate: 08/27/2009 Previous Duedate: 06/04/2009 Requested By: Barborek,W Douglas 06/03/2009 Approved By: Edgell,Douglas W 06/03/2009 Request

Description:

This CA could not be completed by the assigned due date due to competing workload priorities and the complexity of the CA request. Therefore, the due date is being extended to August 27, 2009, prior to the start of 2R20. The new proposed due date is > 6 months beyond CR initiation and this is the second due date extension request; however, the latest interim review (CA-05) approved by the Engineering Director (on 5/21/2009) did note that the CA would need to be extended past the current 6/3/2009 due date and is therefore acceptable. WDB 6/3/2009 Approved

Description:

I concur with this DOE as acting System Engineering Manager. This ODE will not drive the age of the CR beyond that of existing actions and the latest approved interim review which was reviewed by the Engineering Director.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 3 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 12/J7/2008 J5:01 :Of Performed By: Barborek, W Douglas 8/27/2009 10:37:24 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 8/27/200916:03:18 Current Due Date: 08/27/2009 Initial Due Date: 08/27/2009 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e. removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping.

Response

The inspection/testing ofthc Reactor Vessel Flange Leak-off & Pressure Test Connection piping requires additional resources and engineering which transcends the role and responsibility of System Engineering. CA-07 has been created for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects organization. WDB 8/27/2009 Subresponse :

Closure Comments:

Tconcur with closure of this corrective action.

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00003 Version: Approved: r,/

Requested Duedate: 06/04/2009 Previous Duedate: 02/ 18/2009 Requested By: Barborek,W Douglas 02/ 17/2009 Approved By: Edgell,Douglas W 02/17/2009 Request

Description:

This CA could not be completed by the assigned due date due to competing workload priorities, and recent manpower demands from forced outages 1F09-0 R, 1F09-02, and 2F09-02. Therefore, the due date is being extended to June 4, 2009.

The new date is <6 months beyond CR initiation and is therefore acceptable. WDB 2/16/2009 Approved

Description:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00003 Version: 2 Approved: r,/

Requested Duedate: 08/27/2009 Previous Duedate: 06/04/2009 Requested By: Barborek,W Douglas 06/03/2009 Approved By: Edgell,Douglas W 06/03/2009 Request

Description:

This CA could not be completed by the assigned due date due to competing workload priorities and the complexity of the CA request. Therefore, the due date is being extended to August 27, 2009, prior to the start of 2R20. The new proposed due date is > 6 months beyond CR initiation and this is the second due date extension request; however, the latest interim review (CA-05) approved by the Engineering Director (on 5/21/2009) did note that the CA would need to be extended past the current 6/3/2009 due date and is therefore acceptable. WDB 6/3/2009 Approved

Description:

I concur with this DOE as acting System Engineering Manager. This ODE will not drive the age of the CR beyond that of existing actions and the latest approved interim review which was reviewed by the Engineering Director.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 4 G roup Name Assigned By: Eng Sys Mgmt ANO Bond, Vincent S Asstgned To: Eng Sys Mgmt ANO Bond,Yincent S Subassig ned To :

Originated By: zz ANO CRG **lHEA use only** 4/29/2009 14:02:29 Performed By: Bond,Vincent S 5/5/2009 03:27:46 Subperformed By:

Approved By:

Closed By: Bond,Vincent S 5/5/2009 03:28:04 Current Due Date: 06/03/2009 Initial Due Date: 06/03/2009 CA Type: CR CLOSURE REVIEW CA Priority:

Plant Constraint: NONE CA

Description:

Closure Review or interim Review Required (NOTE - an Interim Review requires both "Responsible Manager" AND Director or Above" approval).

Conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist",

Attachment 9.8 of EN-LI- I 02 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section.

OR If all Corrective Actions in this Condition Report are closed, conduct a closure review. This action is being issued by CA&A, per LI-102, for you to review this Condition Report for closure per the guidelines provided in LI-102 and CA - CR Closure Checklist.

Response

Closed to CA-5.

S ubresponse :

Closure Comments:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 5 Group Name Assigned By: Eng Sys Mgmt ANO Bond, Vincent S Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: Bond,Vincent S 5/5/2009 03:27:23 Performed By: Edgell,Douglas W 6/2/2009 15:27:23 Subperformed By: Barborck,W Douglas 5/26/2009 08:06:29 Approved By:

Closed By: Edgell,Douglas W 6/2/2009 15:28:21 Current Due Date: 06/03/2009 Initial Due Date: 06/03/2009 CA Type: CR CLOSURE REVIEW CA Priority:

Plant Constraint: NONE CA

Description:

Closure Review or interim Review Required (NOTE - an Interim Review requires both "Responsible Manager" AND Director or Above" approval).

Conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist",

Attachment 9.8 of EN-LI- I 02 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section.

OR If all Corrective Actions in this Condition Report are closed, conduct a closure review. This action is being issued per LI-I 02 for you to review this Condition Report for closure per the guidelines provided in LI-102 and CA - CR Closure Checklist.

Response

l concur with the following sub-response and closure of this corrective action as acting System Engineering Manager. The next interim review action has been assigned.

Subrcsponse :

interim Review is attached. E-mail documenting Director approval is also attached. WDB 5/26/2009 Closure Comments:

T concur with closure of this corrective action which has been approved by the Engineering Director.

Attachments:

Subresponse Description Interim Review Subresponsc Description e-mail for Director Approval

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!interim Review

ATTACHM ENT 9.8 CR INTERIM AND PERIODIC REVIEW F ORM SHEET 1 O F 1 CR Interim and Periodic Review CR Number: CR-AN0-1-2008-02560 Category Level D AD B IZI C CR Owner Group: ENG SYS MGMT ANO CR

Description:

1R21 W0-00102463 Task 01 was not able to be successfully performed as planned.

CR-AN0-1-2005-011 40 was written to document OE from several plants which identified corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/pressure test connection piping resulting from water left in the piping following refueling outages. AN0-1 was determined to be vulnerable since water has been trapped in this piping during previous operating cycles. The scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. T he inspection approach was to utilize a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV and the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned. Since CR-AN0-1-2005-011 40 was closed to W0-00102463, this CR is being initiated to provide a means of tracking an alternate resolution for addressing the noted OE. As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained during refueling outages prior to installing the RV head, thus mitigating the damage mechanism . The operability statement in CR-AN0-1-2005-011 40 remains applicable.

CR Review: (All No responses require explanation be included.)

The following two CA's have been issued to further evaluate a course of action required to address the OE documented in CR-AN0-1-2005-01 140. The issuance of these actions constitutes an acceptable corrective action plan for this CR. Add itional CA's will be issued as deemed appropriate.

CA CR-AN0-1 -2005-01140 (CA-04) evaluated the feasibility of performing a hydrostatic test of the RV f lange leak-off/pressure test connection piping and concluded that such a test was not feasible. Based on the inability to visually inspect the piping during 1R21 (via boroscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity. This CA is currently due on 6/3/2009, but may need to be extended due to current workload.

CA Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e. removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping.

This CA is currently due on 6/3/2009, but may need to be extended due to current workload.

1. Wi ll the existing corrective actions documented in the condition report, when completed, correct the condition report issue? Yes IZI I No D The corrective actions, when completed, along with the Work Orders to be implemented for 1R22 (W0-00195437 for CA-02 and/or W0-00195438 for CA-03), will ensure the appropriate inspection is performed. Assuming no degradation is found , the overall issue identified by the industry OE will be fully resolved. Any degradation found during 1R22 would be addressed under a new Condition Report.
2. W hat is the expected CR Closure date based on remaining needed actions? DATE: 2/11/2010 EN-Ll-102 REV 13

It is anticipated that this CR will be closed prior to 1R22 Milestone P0-4 7 "All Work Orders "Ready to Work, which is due on February 11 , 2010.

3. Is the previously documented operability/functionality position still valid for the current condition and expected to remain valid until CR closure? Yes [gl / No D /N/A D If the answer is NO, then initiate a new CR to document the concern; CR# N/A
4. Are all Ll-102 requirements for corrective action administration and control being met, i.e.

justifications for Due Date Extensions valid, Long Term Corrective Actions identified, CARB approved CAPRs identified, and appropriate approvals obtained for all?

Yes [gi/ No D

5. What risk to plant operation is imposed by the condition identified and how is risk reduced to an acceptable level for the duration of the action plan?

This CR does not specifically identify a known degraded plant system, structure, or component at AN0-1, but does identify a potential degradation mechanism. Both the RV flange gasket leak detection line and pressure test co nnection lines were flushed/drained during 1R19, 1 R20, and 1R21, thus minimizing the potential for failure resulting from ID initiated, chloride induced, transgranular stress-corrosion cracking (TGSCC). No RV flange gasket leakage was identified during plant heat-up from 1R19, 1 R20, or 1R21 prior to isolating valves RBS-1 and RBS-2, and RCS leakage rates following 1R19, 1R20, and 1R21 have not indicated the presence of any concurrent leakage of the RV flange gaskets and leak detection/pressure test connection piping/components.

The RV flange gaskets are replaced during each refueling outage and valves RBS-1 and RBS-2 are not isolated until just prior to criticality after it has been verified that the inner gasket is not leaking.

Therefore, the likelihood of inner gasket leakage, while possible, is considered low.

In conclusion, the risk to plant operation is minimal and reduced to acceptable levels until the action plan for this CR can be implemented during 1 R22.

Review/ Approval Required:

Director/GM

Title:

Date:

(Print name & Position title)

NOTE: The expectation is to capture the discussion points of this form in a CA. The form itself need not be used, but all points applicable must be addressed.

EN-Ll-102 REV 13

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

~ -mail for Director Approval

BARBOREK, W DOUGLAS From: LAY, LI NDA S Sent: Thursday, May 21 , 2009 5:44 PM To: BARBOREK, W DOUGLAS Cc: EDGELL, DOUGLAS W

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Flange Piping" Interim approved per Cleve.

Linda From: BARBOREK, W DOUG.AS Sent: Thursday, May 21 , 2009 4:20 PM To: REASONER, a.EVELAND Cc: EDGELL, DCUGL.AS W; LAY, LI NDA S

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Range Piping" Cleve, 1 R22 W0-00195437 has been initiated for the pressure (hydro, pneumatic, or vacuum) test of the piping (if we determine we can do that), and 1 R22 W0-00195438 has been initiated for the intrusive inspections, if it comes to that. Keep in mind that the intrusive inspections, if that's the route we have to go, would be very costly, dose intensive, and have a major impact on critical path since it involves destructively removing the reactor cavity seal plate, the concrete shield blocks, and the RV insulation. It is my desire to find a way to perform a pressure test. I am trying to find out how Davis Besse, Oconee, and Calvert Cliffs did their testing.

Since this piping is inaccessible, this OE has proven difficult to resolve. I can bring some drawings to you if you wish to illustrate what we're up against on this.

Thanks, Doug Barborek Entergy Operations. Inc. / Arkansas Nuclear One System Engineer - Unit I Reactor Coolant System System Engineering Building/ N-SYS-4 wbarbo l @entcrgy.com 479-858-4337

!(b)(6) I pager Fro m: LAY, LINDA S Sent: Thursday, May 21, 2009 4: 12 PM To: BA.RBOREK, W DOUG.AS Cc: EDGELL, DOUGLAS W Subject : RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Range Piping"

Cleve's Question - Is inspection scoped in the outage?

From: BARBOREK, W DOUGLAS Sent: Thursday, May 21 , 2009 9:40 AM To: REASONER, QEVElAND; BOND, VINCENT S Cc: EDGELL, DOOGLAS W; EICHENBERGER, JOHN R

Subject:

REQUEST FOR REVIEW - Interim Review for CR-AN0- 1-2008-02560 "Resolve Industry OE on Ch loride Induced Stress Corrosion Oacking on RV Range Aping "

Cleve & Vince, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-05. The CA is ultimately due on Wednesday, June 3, 2009 (initially due by me on Monday, June 1).

Please let me know if you have any questions.

Thanks, Doug Barborek Entergy Operations. Inc. I Arkan.sas uclear One System Engi neer - Unit I Reactor Coolant System System Engi neering Building/ -SYS-4 wbarbo I @cntcrgy.com 479-858-4337

!(b)(6)  ! pager

<< File: CR-AN0-1-2008-02560 CA-05 Interim Review.doc>>

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 6 G roup Name Assigned By: Eng Sys Mgmt ANO Wi lliams,Patrick J Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: Edgell,Douglas W 6/2/2009 15:26: 16 Performed By: Williams,Patrick J 2/9/2010 17:10:37 Subperformed By: Barborck,W Douglas 2/8/20 10 10: 11:1 5 Approved By:

Closed By: Williams,Patrick J 2/9/2010 17:11:04 Current Due Date: 02/ 11 /2010 Initial Due Date: 02/11/20 I 0 CA Type: CR CLOSURE REVIEW CA Priority:

Plant Constraint: NONE CA

Description:

Closure Review or interim Review Required (NOTE - an Interim Review requires both "Responsible Manager" AND Director or Above" approval).

Conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist",

Attachment 9.8 of EN-LI- I 02 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section.

OR If all Corrective Actions in this Condition Report are closed, conduct a closure review. This action is being issued per LI-I 02 for you to review this Condition Report for closure per the guidelines provided in LI-102 and CA - CR Closure Checklist.

Response

Concur with closure.

Subresponse :

Interim Review attached. E-mail documenting Director approval is also attached. WDB 2/8/2010 Closure Comments:

Attachments:

Subresponse Description Interim Review Subresponse Description Director Approval e-mail

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!interim Review

ATTACHMENT 9.8 CR INTERIM AND PERIODIC REVIEW F ORM SHEET 1 O F 1 CR Interim and Periodic Review CR Number: CR-AN0-1 -2008-02560 Category Level DAD B IZI C CR Owner Group: ENG SYS MGMT ANO CR

Description:

1R21 W0-00102463 Task 01 was not able to be successfully performed as planned.

CR-AN0-1-2005-01140 was written to document OE from several plants which identified corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/pressure test connection piping resulting from water left in the piping following refueling outages. AN0-1 was determined to be vulnerable since water has been trapped in this piping during previous operating cycles. The scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utilize a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV and the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned. Since CR-AN0-1-2005-01140 was closed to W0-00102463, this CR is being initiated to provide a means of tracking an alternate resolution for addressing the noted OE. As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained during refueling outages prior to installing the RV head, thus mitigating the damage mechanism . The operability statement in CR-AN0-1-2005-011 40 remains applicable.

CR Review: (All No responses require explanation be included.)

The following CA's have been issued to further evaluate a course of action required to address the OE documented in CR-AN0-1-2005-01140. The issuance of these actions constitutes an acceptable corrective action plan for this CR. Add itional CA's will be issued as deemed appropriate.

CA CR-AN0-1-2005-01140 (CA-04) evaluated the feasibility of performing a hydrostatic test of the RV f lange leak-off/pressure test connection piping and concluded that such a test was not feasible. Based on the inability to visually inspect the piping during 1R21 (via boroscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity. This CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization.

CA Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e. removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping.

This CA was closed on 8/27/2009 to CA-07 wh ich was initiated for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization.

CA The inspection of the RV Flange Leak-off & Pressure Test Connection piping requires additional resources and engineering which transcends the role and responsibility of System Engineering. Initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization. SIPDB Record 4955 was initiated to transfer this scope of work to the Projects Organization and this CA was closed on 11 /19/2009. New CA-08 was issued to track the presentation of SIPDB Record 4955 to the URT.

EN-Ll-102 REV 13

CA Present SIPDB Record (Inspection of the RV Flange Leak-off & Pressure Test Connection Piping) to the URT/MPRC for this scope of work to be executed by the Projects Organization. SYE made the presentation at the 12/7/2009 URT meeting. The URT concurred that if the leak-off &

pressure test connection piping ultimately require physical inspection or replacement such that disassembly/reassembly of the reactor cavity seal plate, concrete shield blocks, and reactor vessel insulation is required , then that significant scope of work wou ld be transferred to the Project Organization. However, at this time, the URT decided that System Engineering should re-evaluate the pressure test option and come up with a plan to perform a pressure test on the leak-off and pressure test connection piping to verify the integrity of the piping. If the pressure test fails on one or both of the lines, the ensuing scope of work to access the piping for ultimate resolution of the issue would then be pursued by the Projects Organization. New CA-09 has been issued to SYE to determine a suitable pressure test method and to initiate the proper implementation documentation.

1 R22 will be utilized to perform scoping walkdowns since this piping is not accessible during power operations. This CA was closed on 1/21/2010.

CA Coordinate with EP&C personnel and determine a suitable pressure test method for the RV flange leak-off/pressure test connection piping, and initiate the proper implementation documentation.

Utilize 1 R22 to perform scoping walkdowns as required . Initiate other corrective actions as required.

Assigned to SYE with a due date of 5/6/2010.

1. Will the existing corrective actions documented in the condition report, when completed, correct the condition report issue? Yes ~ / No D Once CA-09 is completed, and any other actions which may be initiated as a result of CA-09, the CR should be able to be closed to the Work Order process for 1 R23 implementation of a pressure test.

Any actual piping degradation discovered (via failure of the pressure test) during 1R23 wou ld be addressed under a new Condition Report and the mitigation (i.e. pipe replacement) would likely be assigned to the Projects Organization due to the extensive scope required for mitigation. The goal is to have all CAs completed and Work Orders initiated prior to 1R23 Milestone P0-7 "Eng Systems Scope Identified', which is due on July 2, 2010.

2. What is the expected CR Closure date based on remaining needed actions? DATE: 712/2010 It is anticipated that this CR will be closed prior to 1R23 Milestone P0-7 "Eng Systems Scope Identified', which is due on July 2, 2010.
3. Is the previously documented operability/functionality position still valid for the current condition and expected to remain valid until CR closure? Yes ~ I No D /N/A D If the answer is NO, then initiate a new CR to document the concern; CR# N/A
4. Are all Ll-102 requirements for corrective action administration and control being met, i.e.

justifications for Due Date Extensions valid, Long Term Corrective Actions identified, CARB approved CAPRs identified, and appropriate approvals obtained for all?

Yes ~ / No D

5. What risk to plant operation is imposed by the condition identified and how is risk reduced to an acceptable level for the duration of the action plan?

This CR does not specifically identify a known degraded plant system, structure, or component at AN0-1, but does identify a potential degradation mechanism. Both the RV flange gasket leak detection line and pressure test connection lines were flushed/drained during 1R19, 1 R20, and 1R21 ,

thus minimizing the potential for failure resulting from ID initiated , chloride induced, transgranular stress-corrosion cracking (TGSCC). No RV flange gasket leakage was identified during plant heat-up EN-Ll-102 REV 13

from 1R19, 1 R20, or 1R21 prior to isolating valves RBS-1 and RBS-2, and RCS leakage rates following 1R19, 1R20, and 1R21 have not indicated the presence of any concurrent leakage of the RV flange gaskets and leak detection/pressure test connection piping/components.

The RV flange gaskets are replaced during each refueling outage and valves RBS-1 and RBS-2 are not isolated until just prior to criticality after it has been verified that the inner gasket is not leaking.

Therefore, the likelihood of inner gasket leakage, while possible, is considered low. It is noted that recent CR-AN0-1-2010-00056 and CR-AN0-1-2010-00091 have documented two instances where the Control Room received annunciator K09-F1 " Vessel Head Gasket Leak" for 5 minutes on 1/10/2010 and for 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on 1/13/2010, respectively, indicating a potential inner o-ring leak.

However, no increase in RCS leakage or sump fill rate has been observed, which indicates that if the inner o-ring is actually leaking, the leak-off piping is not degraded.

In conclusion, the risk to plant operation is minimal and reduced to acceptable levels until the action plan for this CR can be implemented during 1R23.

Review/ Approval Required:

Director/GM

Title:

Date:-

(Print name & Position title)

NOTE: The expectation is to capture the discussion points of this form in a CA. The form itself need not be used, but all points applicable must be addressed.

EN-Ll-1 02 REV 13

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!Director Approval e-mail

BARBOREK, W DOUGLAS From: MCCOY, JAIME H Sent: Saturday, February 06, 2010 1:33 PM To : BARBOREK, W DOUGLAS; WILLIAMS, PATRICK J Cc: EDGELL, DOUGLAS W

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Flange Piping" I approve the interim review as acting Engineering Director.

Jaime

Original Message-----

From : BAPBOREK, W DOUGLAS Sent: Thursday, February 04, 201 O4:32 Av1 To: MCXX>Y, JAIME H; WI LLIAMS, PATRICKJ Cc: EDGELL, DOUGLAS W; El CHENBERGER, JOHN R Subject : REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Indust ry OE on O,loride Induced Stress Corrosion Oacking on RV Range Piping" Jaime & Patrick, The interim review for the subject CR is attached for your review and concurrence. The draft interim rev iew has been attached to CA-06 of the subject CR. The CA is ultimately due on Thursday, February 11, 2010 (initially due by me on Tuesday, February 9).

Please let me know if you have any questions.

Thanks, Doug Barborek Entergy Operalions. lnc. / Arkansas Nuclear One System Engineer - Unit I Reactor Coolant System System Engineering Building/ N-SYS-4 wbarbo I @cntcrgy.com 479-858-4337

!(b )(6) I pager

<< File: CR-AN0-1-2008-02560 CA-06 Interim Review.doc>>

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 7 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 8/27/2009 16:01:06 Performed By: Barborek, W Douglas 11 /18/2009 13:39:20 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 11/18/2009 14:34:45 Current Due Date: 11 / 19/2009 Initial Due Date: 11 /19/2009 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

The inspection of the Reactor Vessel Flange Leak-off & Pressure Test Conrnection piping requires additional resources and engineering which transcends the role and responsibility of System Engineering.

Initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects orgaruization.

Response

SIPDB Record 4955 has been initiated to transfer the subject inspection from System Engineering (SYE) to the Projects organization as a new project for lR23. Per the attached e-mail, System Engineering has requested that the SIPI) Record be presented by SYE to the Unit Reliability Team (URT) in December 2009.

It is not necessary that this CA track the URT presentation itself since URT approval and MPRC assignment of funds could take several evolutions and is dictated by established site and corporate procedures. The creation of Record 495 5 sufficiently meets the intent of this CA.

WDB 11 / 18/2009 Subresponse :

Closure Comments:

I concur with closure of this corrective action. CA # 8 has been assigned to track the URT Presentation and decision.

Attachments:

Response Description e-mail to URT coordinator

Attachment Header Document Name:

untitled Document Location

!Response Description Attach

Title:

~ -mail to URT coordinator

Page 1 of 3 BARBOREK, W DOUGLAS From: BARBOREK, W DOUGLAS Sent: Wednesday, November 18, 200911 :14 AM To: HARTMAN, ANGELA M Cc: EDGELL, DOUGLAS W; WOODSON, TIMOTHY R

Subject:

SIPD 4955 - Hydro/Inspect AN0-1 RV Flange Leak-off/ Pressure Test Connection Piping due to Davis-Besse OE Angie, Can you put me on the agenda for a December URT to discuss handing this issue over to Projects?

This issue pertains to Davis-Besse OE from 2003 where they found their RV flange leak-off /pressure test connection piping degraded from PWSCC.

I can do more with this as a SYE. We need to hydro the RV flange leak-off I pressure test connection piping, which is buried under the reactor cavity seal plate, reactor shield blocks, and RV insulation. I tried to come up with a hydro method but couldn't, and I tried to inspect these lines externally with a boroscope during 1R21, but was unsuccessful due to the inaccessibility of the piping. Due to the magnitude of what it will take to resolve this issue, I need to turn it over to Projects for resolution.

Thanks, Doug Barborek Entergy Operations. lnc. / Arkansas Nuclear One Syste m Eng ineer - Unit I Reactor Coolant Syste m System Engineeri ng Bu ilding/ N-S YE-4 wbarbu l @emergy.com 479-858-4337

!(b)(6)  ! pager 11/18/2009

Page 2 of 3 l'J Microsoft Access - [SYSTEM PLAN~ING PROJECT]

~ Eile ~dit '.l!_iew I nsert f 0'mat ~ecords Iools ~ indow t!elp ProjectJD 11 49551 System:IJRCS *I ERGrade:11 l. ~ u ~ u Multi Site Initiative: ILI Project #: U 11 Project Tltle:! IAH0 Resolve OE on RV Flange Leak -off/Pres s Test Connection Pipe Cr, Preparer: IJBarborek ,W Douglas _:J Date ll1111a12009 j Project Manager: 11 *I Datell r Project Description IAttributes I Cost Cash Flow I References I Study Phase I Design Phase I Implementation Phase IJ Project Description I Operating Experience report OE-15417 "Cracking Identified in Class 1 Reactor Flange 0 on J anuary 27, 2003 to document through wall cracks on reactor vessel flange 0-ring (g Davis-Besse on October 29, 2002. CR-AN0-1 -2005-01140 documents attempts to resolv Engineering. However, as documented in that CR and CR-AN0-1-2008-02560, System E Recommended Solution I

!Project Management develop p lan to resolve potential issue identifed in OE-15417. Perform pressure tes t est connecti on pie!!!g_ and/or p erform visual inspection of piP.!!!g_. _

Key Benefits to be Validated I 1 1fa1sure integrity of piping to prevent potential boric acid corrosion of RCS components.

On-going Costs I Inone Alternatives Considered ~ 1 Inone I

I Schedule Considerations must be performed during refueling outage with refueling ca1nal drained.

Consequences of tlon Approval I

!The s ubject piping has been found to be degr aded at Davis -Besse andl othe r plants. Until the Atl0 -1 piping is ins pected, t head inner gas k et w ere to leak rem ains. Boric acid corros ion of the r eactor vessel could occur if this pi ping leak s.

Form View I. '-' Start! J ,I!) r~ I!!) [!) >> J ls:_j Inbox - Mier ... , S 4 Int ernet ... * ~ . . Paperless(.. . I!IJ P2E: Passpo ... j* Adobe RI 11/1 8/2009

Page 3 of 3 11/18/2009

Entergy I CA DUE DATE EXTENSION I CR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00007 Version: Approved: r,/

Requested Duedate: 11 / 19/2009 Previous Duedate: 10/29/2009 Requested By: Barborek,W Douglas 10/28/2009 Approved By: Edgell,Douglas W 10/28/2009 Request

Description:

This CA was not completed by the assigned due date due to competing workload priorities. Additional time is required to initiate the SIPDB record and present it to the URT/MPRC. The due date is being extended to 11/19/2009. The extension of the due date does not negatively impact any installed plant SSCs and is therefore acceptable. WDB l 0/28/2009 Approved

Description:

I concur with the following DDE which is the first extension on a CR that is greater that 6 month old. This DDE does not increase the age of the CR beyond the due date of other existing actions therefore additional approvals are not required.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 8 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS Staff ANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 11/ 18/2009 14:33:08 Performed By: Barborek, W Douglas 1/2 1/2010 16:43:57 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 1/2 1/2010 16:50:06 Current Due Date: Ol /2 1/20 I 0 Initial Due Date: 0 1/21 /20 10 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Present S1PD 4955 (Inspection oftbe Reactor Vessel Flange Leak-off & Pressure Test Connection piping) to the URT/MPRC for this this scope of work be executed by the Projects Organization.

Response

SYE proposed at the 12/7/2009 URT meeting (see attached presentation) to transfer the scope of work outlined in SlPD Record 4955 (Inspection of the Reactor Vessel Flange Leak-off & Pressure Test Connection piping) to the Projects Organization. The URT concurred that if the leak-off & pressure test connection piping ultimately require physical inspection or replacement such that disassembly/reassembly of the reactor cavity seal plate, concrete shield bloc ks, and reactor vessel insulation is required, then that significant scope of work would be transferred to the Project Organization.

However, at this time, the URT decided that System Engineering should re-evaluate the pressure test option and come up with a plan to perform a pressure test on the leak-off and pressure test connection piping to verify the integrity of the piping. If the pressure test fails on one or both of the lines, the ensuing scope of work to access the piping for ultimate resolution of the issue would then be pursued by the Projects Organization.

As such, new CA-9 has been issued to SYE to determine a suitable pressure test method and to initiate the proper implementation documentation. IR22 wi[I be utilized to perform scoping walkdowns since this piping is not accessible during power operations.

WDB 1/21/2010 Subresponsc :

Closure Comments:

I concur with closure of this corrective action.

Attachments:

Response Description URT Presentation Outline

Attachment Header Document Name:

untitled Document Location

!Response Description Attach

Title:

luRT Presentation Outline

SIPD 4955 "Resolve OE on RV Flange Leak-off/Pressure Test Connection Pipe Cracking"

  • Operating Experience report OE-1541 7 "Cracking Identified in Class 1 Reactor Flange 0-Ring Monitor Piping" was issued on January 27, 2003 to document through wall cracks on reactor vessel flange 0 -ring (gasket) monitor piping found at Davis-Besse on October 29, 2002.
  • Problems also identified at Calvert Cliffs 1 & 2 and other plants.
  • AN0-1 was determined to be susceptible since water was not drained in these lines for many years since leak-off isolation valves were closed since mid-1990's. Until 2005.
  • Procedure changes made to drain lines since 1Rl9; however, damage may have occurred before damage mechanism was mitigated. Must inspect to be sure.
  • If cracked, and we develop an inner gasket leak on head, we could cause BAC damage to exterior of RV.
  • System Engineering has been unable to develop successful plan for hydrostatic test. ....and contingencies if hydro fails is expensive (i.e. remove reactor cavity seal plate, shield blocks, and RV insulation).
  • Visual inspection attempted by horoscope in 1R2 l via reactor cavity shield plate openings, but not enough access for inspection.
  • Scope exceeds capability of SYE, need to tum over to Projects for resolution due to magnitude of scope and contingency.
  • 1R22 Work Orders W0-00195437 for pressure test. W0-00195438 for visual inspection. Most likely need to be deferred to 1R23. Risk of deferral is fairly low since RV head gasket leakage is rare, and lines are drained of water prior to operation.

Barborek 12/7/2009

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 9 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Doug las W Asstgned To: Eng Sys Mgmt ANO Barborek, W Douglas Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: Edgell,Douglas W 1/21/2010 16:51 :08 Performed By: Barborek, W Douglas I0/25/20 IO 17:2 1:O~

Subperformed By: Barborck,W Douglas 10/25/20 10 16:48:29 Approved By:

Closed By: Edgell,Douglas W 10/26/2010 16:45:30 Current Due Date: I0/28/20 I 0 Initial Due Date: I 0/28/20 I0 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Coordinate with EP&C personnel and determine a suitable pressure test method for the RV flange leak-off/pressure test connection piping, and initiate the proper implementation documentation. Utilize IR22 to perform scoping walkdowns as required. lnitiate other corrective actions as required.

Response

Sub-response is complete and satis factory. WDB I 0/25/20 I 0 Subrespo nse :

The attached RV Flange Piping pressure test plan has been developed with support and concurrence from EP&C (Rick Holman & Mike Paterak), Piping Design (Maqbool Bhatti), and RP (Dan Sto ltz). CA-10 has been initiated to assign followup corrective actions as required to implement the pressure test plan in I R23 via W0-0019543 7. WDB I 0/25/20 I 0 Closure Comments:

I conc ur with closure of this corrective action.

Attachments:

Subresponse Description I R23 Pressure Test Plan for RV Flange Piping

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

11R23 Pressure Test Plan for RV Flange Piping

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping Pressure Test Plan Page l of 6 OBJECTIVE:

Develop a plan to hydros.tatically test the Reactor Vessel leak-off and pressure test connection piping from the Reactor Vessel to the downstream iso lation valves (RBS-1 and RC-5) to resolve industry operating experience document OE-15417 "Cracking Identified in Class 1 Reactor Fla nge 0-ring Monitor Piping" relative to AN0-1.

REFERENCES:

l. CR-AN0-1-2005-01]40 "OE-15417 Appears to be Applicable to ANO- I"
2. CR-AN0-1-2008-02560 "External Inspection of RV Flange Leak-off Piping Not Successful during 1R21"
3. PEAR-93-0246 "Reactor Vessel Closure Head Leak-off Line Isolation"
4. OE-15417 "Cracking Identified in Class l Reactor Flange 0 -ring Monitor Piping"
5. P&ID M-230 "Reactor Coolant System"
6. ANO drawing M- 114-AC-l " 1/4 to 2 Inch Bolted Bonnet Gate Valve Forged"
7. ANO drawing MlB-223 "Reactor Vessel Upper Shell Assembly"
8. ANO isometric RC-201 "Reactor Vessel R-1 Flange Gasket Leak-off Drain"
9. ANO isometric RC-202 "Reactor Vessel R-1 F la nge Gasket Leak-off Drain"
10. CALC-93-E-5035-07 "Code Qualification of Lines CCC-6-1" & HSD-3-1" Piping &

Pipe Support on Isometrics RC-201 & RC-202"

11. 1R23 W0-00195437 "R-1 Perform Pressure Test of RV Intergasket Leak-off Lines"
12. ANO Procedure OP-5 120.247 "Pressure Test"
13. ASME B&PV Code Section XI Article IWA-5000 "System Pressure Tests (General requirements)"
14. ASME B&PV Code Section XI Article IWD-5000 "System Pressure Tests (Class 3 requirements)"
15. Entergy Procedure ClEP-PT-001 "ASME Section XI, Division l System Pressure Testing"
16. Website for Curtiss Wright GripTight High Pressure Test Plugs -

http://estgroup.cwfc.com/productsServices/spokes/O 1a HTplugs 04-S02 HiPressure.htm

17. http://www.cob-industries.co m/highpressurestopper. aspx BACKGROUND INFORMATION:

OE-15417 was issued on January 27, 2003 to document through wall cracks on reactor vessel flange 0 -ring (gasket) monitor piping found at Davis Besse on October 29, 2002.

The cracks were a result of chloride induced transgranular stress corrosion cracking due to water left in the piping during plant operation. The first several feet of piping which exit the RV flange is adjacent to the reactor vessel before exiting the RV insulation and experiences temperatures near reactor vessel operating temperatures (550-600°F). AN0-1 has two leak-off ports. One port is associated with the RV flange (intergasket) pressure test connection (a feature whic h is not used), and the other port is associated with the RV flange (intergasket) leak-off and leak detection connection Barborck l 0/25/20 I 0

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping Pressure Test Plan Page 2 of 6 AN0-1 has operated with both of these lines isolated during power operations since the early 1990 's, and the piping was not routine ly drained fo llowing refu eling outages until lR 19 (Fall 2005). As such, the AN0 - 1 piping was subject to the degradation mechanism described in OE- 15417 from the early 1990's to late 2005.

Visua l inspection of the piping (external) via boroscope was attempted during 1R2 l , but was unsuccessful due to the inaccessibi lity of the piping. As such, a pressure test is required to resolve this issue for AN0 -1 and verify the integrity of the subject piping.

Although the pressure test is not a Code requirement and is therefore subject to Owner-specified rules, Code rules will be utilized to verify integrity.

PLAN:

  • Cut l inch piping approximately 1 foot above (i.e. upstream) valves RBS- 1 and RC-5 (see photos 1 thru 4 below).

o Pip ing is l inch Sch 160, A-3 12 TP-3 16 per isometric RC-201 and RC-202.

RBS- 1 and RC-5 are Velan mode l W05-3054B-l3MS (l inch, socket welded, 1500# class, stainless steel F3 16 CF8M gate valves) shown on drawing M- 114-AC-l.

  • lnsert high pressure test plugs in the RV flange 0.5 inch ports (see Figure 1 & 2 be low) .

o FME controls must be in place to ensure plugs do not enter the RV in event of failure.

o Plugs should have venting capability (for pipe fill).

  • Insert high pressure test plugs in CCC-6 piping.

o Plugs to be used as hydro connection.

  • F ill piping (expected volume to be less than one gallon per connection).
  • Perform hydrostatic test at 101 to 110% of CCC-6 Design pressure (2500 psig), for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (4 hrs is Code requirement for insulated piping vs. 10 minutes for non-insulated).
  • Acceptance criteria is zero pressure decay.
  • Remove test equipme nt.
  • Reweld piping using 1 inch couplings, perform NDE as required for Class 3 piping.

OPEN ISSUES/ QUESTIONS / NOTES:

1. Will the test be performed before RVCH removal o r after RVCH installation?
2. ls there a possibil ity that TE-1052 could leak? Appears to be all welded construction.
3. Per discussions with P iping Design (Maqbool Bhatti), installation of l inch couplings would be acceptable for the Class 3 CCC-6 piping. EC Mark-up will be required to revise drawings RC-20 1 and RC-202, and CALC-93-D-5035-07.
4. Per ALARA (Dan Stolz), use of a 1 inch coupling in the vertical piping above RBS-1

& RC-5 would be acceptable re lative to creation of crud traps. Current dose rates are about 15 mrem for both of these inaccessible valves.

Barborck l 0/25/20 I 0

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping Pressure Test Plan Page 3 of 6 (b)(4)

Figure 1 - Excerpt from drawing MlB-223 showing 1/2 inch leak-off port Barborck l 0/25/20 I 0

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping Pressure Test Plan Page 4 of 6 (b)(4)

Figure 2 - Excerpt from drawing MlB-223 showing 1h inch leak-off port Barborck l 0/25/20 I 0

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping Pressure Test Plan Page 5 of 6 Photo 1- RV Leak-off piping (RBS-1 & 2), P-32A Cold Leg Opening Photo 2 - RV Leak-off piping (RBS-1 & 2), P-32A Cold Leg Opening Barborck l 0/25/20 I 0

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping Pressure Test Plan Page 6 of 6 Photo 3 - RV Leak-off piping (RC-5), P-32D Cold Leg Opening Photo 4- RV Leak-off' piping (RC-5), P-32D Cold Leg Opening Barborck l 0/25/20 I 0

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00009 Version: Approved: r,/

Requested Duedate: 08/ 12/20 I 0 Previous Duedate: 05/06/20 I 0 Requested By: Barborek,W Douglas 05/04/2010 Approved By: Edgell,Douglas W 05/05/2010 Request

Description:

Scoping walkdowns were performed during I R22; however, due to current workload, this CA cannot be completed at this time. CA is extended to 8/ 12/20 10 to allow additional time for completion. This CA involves development of activities to be performed in I R23 and the extension of the due date has no impact on installed plant SSCs. WDB 5/4/20 l 0 Approved

Description:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00009 Version: 2 Approved: r,/

Requested Duedate: I0/28/20 I 0 Previous Duedate: 08/ 12/20 10 Requested By: Barborek,W Douglas 08/11/20 I 0 Approved By: Williams,Patrick J 08/11 /20 10 Request

Description:

Scoping walkdowns were performed during JR22; however, due to current workload (including emergent work), this CA cannot be completed at this time. CA is extended to 10/28/2010 to allow additional time for completion. This CA involves development of activities to be performed in IR23 and the extension of tbe due date has no impact on installed plant SSCs.

WDB 8/ 11/2010 Approved

Description:

Approved.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 10 Group Name Assigned By: Eng Sys Mgmt ANO Barborek, W Douglas Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborek,W Douglas Originated By: Edgell,Douglas W 10/26/2010 16:41:1~

Performed By: Edgell,Douglas W 11 /17/2010 18:26:22 Subperformed By: Barborek,W Douglas 11/1 7/2010 15:07: 11 Approved By:

Closed By: Edgell,Douglas W 11/17/2010 18:26:48 Current Due Date: 11 / 18/20 I 0 Initial Due Date: 11 /18/2010 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Based on the pressure test plan developed by SYE and EP&C as documented in CA-09, initiate and assign the requisite CAs required to implement the test plan during I R23 (e.g. develop EC for installing couplings on CCC-6 piping, assist Planning in W0-00195437 development, initiate procurement of hydro plugs, etc.).

Response

Tconcur with the fo llowing sub-response and closure of this corrective action.

Subresponse :

The following two CA's have been initiated to support implementation of the pressure test plan documented in CA-09.

Additional CA's will be issued in the future, if warranted.

New CA Develop EC to support W0-001 95437. lfrequi_red, provide justification for installation of couplings on the affected CCC-6 piping. Also, if required, provide engineering requirements for FME controls associated with the pressure test equipment and components to be used in close proximity to the RV flange. See pressure test plan in CA-09 for additional details.

New CA Detennine the appropriate type & size of hydro plugs to be utilized for the pressure test (W0-00195437) outlined in CA-09. Initiate actions to procure the hydro plugs, as required.

WDB I J/17/2010 Closure Comments:

I concur with closure.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 11 G roup Name Assigned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys Mgmt ANO Barborek, W Douglas Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: Edgell,Douglas W 11/ 17/2010 18:27:5 1 Performed By: Barborek, W Douglas 7/20/2011 16:48:33 Subperformed By: Barborck,W Douglas 7/20/2011 16:46:54 Approved By:

Closed By: Edgell,Douglas W 7/2 1/2011 14:53:24 Current Due Date: 07/2 1/2011 Initial Due Date: 07/21 /2011 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Develop EC to support W0-001 95437. lfrequired, provide justification for installation of couplings on the affected CCC-6 piping. Also, if required, provide engineering requirements for FME controls associated with the pressure test equipment and components to be used in close proximity to the RV flange. See pressure test plan in CA-09 for additional details.

Response

Sub-response is acceptable. EC is no longer required since testing methodology which existed at the time of CA development has changed. WDB 7/20/2011 S ubresponse :

CA

Description:

Develop EC to support W0-00195437. If required, provide justification for installation of couplings on the affected CCC-6 piping. Also, if required, provide engineering requirements for FME controls associated with the pressure test equipment and components to be used in close proximity to the RV flange. See pressure test plan in CA-09 for additional details.

CA Response:

At the time this CA was written, the plan was to sever the one inch piping above the isolation valves (RBS- I and RC-5) and to install ported hydro plugs at this location and solid hydro plugs at the RV flange leak-off port locations above. Subsequent to the initiation of this CA, the plan has been revised following discussions with Outage Management (see attached plan) to use the isolation valves as the test boundary and to pressurize at ported hydro plugs installed in the RV flange ports. As such, the plan no longer relies on severing the piping, thus removing the installation of couplings to restore the CCC-6 piping.

While the severing and coupling of the piping would ultimately remain a contingency in the event the boundary valves leak, the severing of the piping and repeat of the pressure test would occur at the end of the outage before head set, thus allowing time to develop an EC to install the couplings.

Additionally, it is anticipated that this test and all contingency planning will be transferred from System Engineering to Project Management for 1R24 implementation (this work has been defen-ed from I R23 due lack of contingency plans).

Project Management may choose to develop contingency £Cs; however, these would not be anticipated to be developed by System Engineering as part of core business. The same philosophy applies to FME controls, which would likely not require an EC.

For the reasons discussed above, the noted EC is no longer required at this time to be developed by System Eng ineering as a prerequisite for perfonning the initial pressure test. Accordingly, this CA is being closed with no action taken. If a future need arises to develop EC's to support the pressure test, new Corrective Actions will be initiated as appropriate.

WDB 7/20/2011

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 Closure Comments:

l concur with closure of this corrective action.

Attachments:

Subresponse Description Hydro plan 20-201 1 version

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!Hydro plan 20-2011 vers ion

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page l of 8 OBJECTIVE:

Develop a plan to hydros.tatically test the Reactor Vessel leak-off and pressure test connection piping from the Reactor Vessel to the downstream iso lation valves (RBS-I and RC-5) to resolve industry operating experience document OE-15417 "Cracking Identified in Class 1 Reac tor Flange 0-ring Monitor Piping" relative to AN0-1.

REFERENCES:

l. CR-AN0-1-2005-01]40 "OE- 15417 Appears to be Applicable to ANO-I"
2. CR-AN0-1-2008-02560 "External Inspection of RV Flange Leak-off Piping Not Successful during 1R21"
3. PEAR-93-0246 "Reactor Vessel Closure Head Leak-off Line Isolation"
4. OE-15417 "Cracking Identified in Class l Reactor Flange 0-ring Monitor Piping"
5. P&ID M-230 "Reactor Coolant System"
6. ANO drawing M- 114-AC- l "1/4 to 2 Inch Bolted Bonnet Gate Valve Forged"
7. ANO drawing MlB-223 "Reactor Vessel Upper Shell Assembly"
8. ANO isometric RC-201 "Reactor Vessel R-1 Flange Gasket Leak-off Drain"
9. ANO isometric RC-202 "Reactor Vessel R-1 F lange Gasket Leak-off Drain"
10. CALC-93-E-5035-07 "Code Qualification of Lines CCC-6-1" & HSD-3-1" Piping &

Pipe Support on Isometrics RC-201 & RC-202"

11. 1R23 W0-00195437 "R-1 Perform Pressure Test of RV Intergasket Leak-off Lines"
12. ANO Procedure OP-5120.247 "Pressure Test"
13. Entergy Procedure CIEP-PT-001 "ASME Section XI, Division 1 System Pressure Testing"
14. Website for Curtiss Wright GripTight High Pressure Test Plugs -

http://estgroup.cwfc.com/productsServ ices/spokes/O la HTplugs 04-S02 HiPressure.htm BACKGROUND INFORMATION:

OE-15417 was issued on January 27, 2003 to document through wall cracks on reactor vessel flange 0-ring (gasket) monitor piping found at Davis Besse on October 29, 2002 (reference also Davis Besse OE-15061 and OE-20437). The cracks were a result of chloride induced transgranular stress corrosion cracking due to water left in the piping during plant operation. Similar failures have occurred at Oconee Unit 3 (B&W), Beaver Valley (Westinghouse) and Calvert Cliffs Units 1 & 2 (CE)[OE-6543]. This failure mechanism is not unique to B&W plants.

The first several feet of piping which exit the RV flange is adjacent to the reactor vessel before exiting the RV insulation and experiences te mperatures near reactor vessel operating temperatures (550-600°F). AN0-1 has two leak-off ports. One port is associated with the RV flange (intergasket) pressure test connection (a feature whic h is not used), and the other port is associated with the RV flange (intergasket) leak-off and leak detection connection.

Barborck 4/20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 2 of 8 AN0-1 has operated with both of these lines isolated d uring power operations since the early/mid 1990's to late 2005 for the leak-off piping, and fro m 1974 to the late 2005 for the pressure test connection piping. The piping was no t routine ly drained fo llowing refueling o utages until 1Rl9 (Fall 2005). As such, the ANO-I piping was subject to the degradation mechanism described in OE- 15417 fo r ma ny years. AN0 -2 has always operated with the iso latio n valves open, and is therefore not susceptible to this failure mechanism.

Visual inspectio n of the pip ing (external) via bo roscope was attempted during 1R2 1, but was unsuccessful due to the inaccessibility of the piping. As such, a pressure test is required to resolve this issue for AN0 -1 and verify the integrity of the subject piping.

The pressure test is not a Code require ment and is therefore subject to Owner-specified rules.

SYE presented a request to the URT in December 2009 to tmn this issue over to Project Management due to the efforts associated with the contingencies. URT directed SYE to develop and perform pressure test, and that contingenc ies would be handled by Projects.

TEST PLAN:

  • Will use 4 gpm hydro pump utilized in 2R21 for 2BCA- 14-3" (W0-00238467-01).
  • Operatio ns verify tha t valves RBS-1 & RBS-2 are closed on the leak-off piping, and that valve RC-5 is closed w ith the b.lind flange installed on the pressure test connection piping (see photos l thru 4 below).

o CCC-6 Piping is l inch Sch 160, A-3 12 TP-3 16 per isometric RC-201 and RC-202.

o RBS-1 and RC-5 are Ve.I an model W05-3054B-13MS (1 inch, socket welded, 1500# class, stainless steel F3 16 CF8M gate valves) shown on drawing M- 114-AC- l .

  • Fill pip.ing at RV flange 1/2" ports (expected volume to be less than o ne gallo n per connection).
  • Insert ported high pressure test plugs in the RV flange 0.5 inch ports (see Figure l &

2 below). Port will be connect ion to hydro pump.

o FME controls must be in place to ensure plugs do not enter the RV in event of hydro plug failure/ejection.

  • Perform venting of hydro pump test rig.
  • Perform hydrostatic test at 100% (+ I %, -0%) of CCC-6 Design pressure (2500 psig),

for 20 minutes (raise to test pressure, isolate pump from piping, secure pump).

  • Acceptance criteria is zero (i.e. negligible) pressure decay.
  • If test results are acceptable, remove test equipment and restore system.

Barborck 4/ 20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 3 of 8 CONTINGENCY SCENARIOS:

  • Test is successfu l, no pressure decay in e ither line.

Done, OE resolved.

  • M inor pressure decay in either or both ljnes. indicating either boundary valve leakage or actual hairline crack (must eliminate false negative).

o Cycle boundary valves and perform test again.

o If minor pressure decay persists, cut pipe above valves, insert hydro plug in pipe, and perform test again, most likely after refueling is complete (will need EC to sever pipe and reweld with 1" coupling).

o If minor pressure decay persists indicating a ha irline crack, then:

  • open RBS- 1 and RBS-2 for Cycle 24 operation to ensure lines remain depressurized (will require procedure changes to change valve configuration),
  • Perform Operability evaluation for 1 additio nal cycle of operation,
  • replace/cap piping during 1R24 (requires removal/reinstallation of reactor cavity shield plate, concrete shield blocks, RV insulation)
  • Significant pressure decay in one line indicating total loss of integrity, but other line is fully intact o insta ll rolled tube plug in affected port (will require AREVA analysis, equipment and EC, not yet confirmed as viable solution) o open valves (re move blind flange is applicable) on opposing line for pressure relief (will require procedure changes to change configuration) o perform Operability evaluation for 1 additional cycle of operation o replace/cap piping during lR24 (requires removal/reinstallation of reactor cavity shield plate, concrete shield blocks, RV insulation)
  • Significant pressure decay in both lines, indicating total loss of integrity of both lines.

o During l R23, replace leak-off piping, cut/cap pressure test connection. (requires removal/reinstallation of reactor cavity shield plate, concrete shield blocks, RV insulation)

Barborck 4/20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 4 of 8 (b)(4)

Figure 1 - Excerpt from drawing MlB-223 showing 112 inch leak-off port Barborck 4/20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 5 of 8 (b)(4)

Figure 2 - Excerpt from drawing MlB-223 showing 1h inch leak-off port Barborck 4/20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 6 of 8 Photo 1- RV Leak-off piping (RBS-1 & 2), P-32A Cold Leg Opening Photo 2 - RV Leak-off piping (RBS-1 & 2), P-32A Cold Leg Opening Barborck 4/20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 7 of 8 Photo 3 - RV Leak-off piping (RC-5), P-32D Cold Leg Opening Photo 4- RV Leak-off' piping (RC-5), P-32D Cold Leg Opening Barborck 4/20/20 I I

AN0-1 Reactor Vessel Leak-off/Pressure Test Connection Piping

      • DRAFT*** W0-00195437 Pressure Test Plan ***DRAFT***

Page 8 of 8 OPEN ISSUES/ QUESTIONS / NOTES:

1. Will the test be performed before RVCH removal or after RVCH installation?
2. Is there a possibility that TE-1052 could leak? Appears to be all welded construction.
3. Per discussions with Piping Design (Maqbool Bhatti), installation of 1 inch couplings would be acceptable for the Class 3 CCC-6 piping. EC Mark-up will be required to revise drawings RC-201 and RC-202, and CALC-93-D-5035-07.
4. Per ALARA (Dan Stolz), use of a 1 inch coupling in the vertical piping above RBS-1

& RC-5 would be acceptable relative to creation of crud traps. Current dose rates are about 15 mrem for both of these inaccessible valves.

Barborek 4/20/20 I I

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00011 Version: Approved: r,/

Requested Duedate: 07/21/2011 Previous Duedate: 02/20/2011 Requested By: Barborek,W Douglas 02/ 14/2011 Approved By: Edgell,Douglas W 02/15/201 1 Request

Description:

This CA could not be completed by the original due date due to competing work priorities. The current plan is to presurize against closed valves (RBS-I and RC-5) in lieu of severing the piping above the valves and installing hydro plugs. As such, the use of couplings is a contingency to be used only if pressurizing against the valves is unsuccessful. The d ue date is be ing changed to 7/21/2011 and is acceptable since it corresponds with Milestone 48 "All Contingency Plans Approved". WDB 2/14/2011 Approved

Description:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 12 G roup Name Assig ned By: Eng Sys Mg mt ANO Edgell,Douglas W Asstgned To: Eng Sys Mgmt ANO Barborek, W Douglas Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: Edgell,Douglas W 11/ 17/2010 18:28:58 Performed By: Barborek, W Douglas 6/7/20 11 14:44:00 Subperformed By: Barborck,W Doug las 617/2011 14:41 :04 Approved By:

Closed By: Edgell,Douglas W 6/9/20111 3:58:1 2 Current Due Date: 06/09/2011 Initial Due Date: 06/09/2011 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Determine the appropriate type & size of hydro plugs to be utilized for the pressure test (W0-00195437) outlined in CA-09.

Initiate actions to procure the hydro plugs, as required.

Response

Sub-response is appropriate. CA-18 assigned to Barborek. WDB 6/7/2011 Subrcsponse :

The 1/2 inch (ref. dwg. MIB-223) ports on the reactor vessel flange will be used as the hydro pump connections. The respective boundary valves (RBS-1/2 & RC-5/blind flange) will be used for isolation on the downstream end of the test boundaries.

The recommended type of hydro plug is the SQ2 High Pressure Test Plug manufactured by Curtiss-Wright Flow Control Company. These test plugs are designed for maximum operating pressures up to 6500 psig which bounds the expected test pressure of2500 psig. The test plugs have l/8 inch threaded ports for connection to the hydro pump skid.

Jt is recommended that both model SQ2-0047 and SQ2-0050 test plugs be procured for this test. The pipe (i.e. port) ID size range for the model SQ2-0047 test plug is 0.47 to 0.5 inch, and the pipe (i.e. port) ID size range for the model SQ2-0050 test plug is 0.5 to 0.53 inch. Since drawing MI B-223 shows that the leak-off ports are 1/2 inch diameter with no tolerance specified, procurement of both sizes is prudent to ensure proper fit.

Per the company website, the lead time for the plugs is typically on the order of severa l days. Information from the C urtis-Wright website is attached to this CA for information.

CA- 18 has been issued for SYE to develop CATIDs for the aforementioned test plugs. Additional actions will b e issued as appropriate during completion of CA- 18.

WDB 6/7/2011 Closure Comments:

I concur with closure of this corrective action.

Attachments:

Subresponse Description SQ2-0050 Test Plug Design Data Subresponse Description SQ2-0047 Test Plug Design Data Subresponse Description SQ2 High Pressure Test Plug S pecs

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 Attachments:

Subresponse Description SQ2 High Pressure Test Plug Info

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

~ Q2-0050 Test Plug Design Data

EST Group, a bus iness unit of Curtiss-Wright Flow Control Company 2701 Township Line Road, Hatfield, PA 19440 Telephone: (215) 721-1100 800-355-7044 Fax: (215) 721-1101 E-Mail: est-info@curtisswright.com Website: estgroup.cwfc.com Item # SQ2-0050, SQ2 High Pressure Test Plug QUOTE SQ2 High Pressure Test Plug To fit pipe and tube inside diameters from 0.48" (12.1 mm) to 0.92" (23.4mm). Working Pressures to 6,500 psi (466 Bar).

Notes:

1) Working Pressure determined for testing in ASTM A105 Grade B pipe. Actual working pressure may vary by pipe material.

2)For pipe or tube sizes larger that 0.92" (23.4mm) please see the GripTight' Test Plugs SQ2-0050 to S02-0060 Specifications subject to change without notice.

SPECIFICATIONS Pipe I.D. 0 .5 to 0 .53 Size Range Qn)

Pipe I.D. 12.7 to 13.5 Size Range (mm)

Plug O.D. 0.47 (in)

Plug O.D. 11 .9 (mm)

Undercut from 0.03 Minimum I.D. (in)

Undercut from 0 .8 Minimum I.D. (mm)

Std Seal Material Urethane Shaft O.D. (in) 1/2

  • 1/4 12.7
  • 6.4 Shaft O.D. (mm) 0.11 Shaft I.D. (in) 2.8 Shaft 1.0. (mm)

Urethane Shaft Seal Material 1/8 M Outlet Installation 2318 Depth (in)

Installation 60.3 Depth (mm) 6/712011 I Page 1 of 2

51 /16 0/A Length (in) 0 /A Length (mm) 128.6 Max Operating 6500 Pressure (psi)

Max Operating 446 Pressure (Bar) 3/8 Shaft Hex Wrench Size 314 Hex Nut Wrench Size A~proximate 0.4 S 1pping Weight {l bs)

Approximate 0.18 Shipping Weight (kgs) 6/7/2011 I Page 2 of 2

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

~ Q2-0047 Test Plug Design Data

EST Group, a bus iness unit of Curtiss-Wright Flow Control Company 2701 Township Line Road, Hatfield, PA 19440 Telephone: (215) 721-1100 800-355-7044 Fax: (215) 721-1101 E-Mail: est-info@curtisswright.com Website: estgroup.cwfc.com Item# SQ2-0047, SQ2 High Pressure Test Plug QUOTE SQ2 High Pressure Test Plug To fit pipe and tube inside diameters from 0.48" (12.1 mm) to 0.92" (23.4mm). Working Pressures to 6,500 psi (466 Bar).

Notes:

1) Working Pressure determined for testing in ASTM A105 Grade B pipe. Actual working pressure may vary by pipe material.

2)For pipe or tube sizes larger that 0.92" (23.4mm) please see the GripTight' Test Plugs SQ2-0050 to S02-0060 Specifications subject to change without notice.

SPECIFICATIONS Pipe I.D. 0.47 to 0.5 Size Range Qn)

Pipe I.D. 11 .91012.7 Size Range (mm)

Plug O.D. 0.44 (in)

Plug O.D. 11.2 (mm)

Undercut from 0.03 Minimum I.D. (in)

Undercut from 0 .8 Minimum I.D. (mm)

Std Seal Material Urethane Shaft O.D. (in) 1/2

  • 1/4 12.7
  • 6.4 Shaft O.D. (mm) 0.11 Shaft I.D. (in) 2.8 Shaft 1.0. (mm)

Urethane Shaft Seal Material 1/8 M Outlet Installation 2318 Depth (in)

Installation 60.3 Depth (mm) 6/712011 I Page 1 of 2

51 /16 0/A Length (in) 0 /A Length (mm) 128.6 Max Operating 6500 Pressure (psi)

Max Operating 446 Pressure (Bar) 3/8 Shaft Hex Wrench Size 314 Hex Nut Wrench Size A~proximate 0.4 S 1pping Weight {l bs)

Approximate 0.18 Shipping Weight (kgs) 6/7/2011 I Page 2 of 2

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

~ Q2 High Pressure Test Plug Specs

EXPANSION SEAL TECHNOLOGIES DC8050 05/98 REV 112/99 SQ2 High Pressure Test Plugs - Technical Specifications 1/8 1/8 NPT INSTA_LA TION DEPTH J INSTALLATION DEPT,

- -- - - - - - - 0/A LrnGTH - - - -~ - -- - - - - - - 0/A LENGTH - - - - --

SQ2-0050 to SQ2-0060 SQ2-0062 to SQ2-0090 Part Nuni>er I.D. Size Ranae Plug Undercut From Std Seal Shaft Shaft Installation 0 /A Length Max. Op. Shaft Hex Hex Nut Appro Xlmate Min. Max. Min. Max. O.D. Minimum I.D. Mafl O. D. (2) I.D. Do ofh Le* nth Pressure Wrench Size Wrench Size Shipping Weight (in) (in) (mm) (mm) ( in) (mm) (in) (mm) (1) ( in) ( mm) (in) (mm) (in) (mm) (in) ( mm) (psi) (Bar) (in) ( in) (lbs) (kgs)

SQ2-0047 0 47 0.50 119 12 7 0. 44 11 2 0 .03 08 u 1/2-114 127- 6.4 011 28 23/8 603 5 1116 1286 6500 446 3/8 3/4 0.40 0.18 SQ2-0050 0.50 0.53 127 13.5 0.47 11.9 003 0.8 u 1/2 - 1/4 127 - 6.4 0. 11 2.8 23/8 60.3 5 1/16 128.6 6500 446 3/8 3/4 0.40 0.18 S02-0053 0.53 0.56 135 14 2 0.50 127 0 .03 08 u 1/2

  • 1/4 12.7- 64 011 28 2 318 603 5 1/16 1286 6500 446 3/8 3/4 0.40 0.18 SQ2-0056 0.56 0.60 142 152 0.53 135 0 .03 08 u 112 - 1/4 12.7 - 6.4 011 28 2 318 60.3 5 1116 1286 6500 446 3/8 3/4 0.40 0.18 SQ2-0060 0.60 0.62 15.2 15.8 0.57 14.5 0 .03 0.8 u 1/2 - 1/4 12.7 - 6.4 0 .11 2.8 2318 60.3 5 1/16 128.6 6!D'.J 446 3/8 3/4 0.40 0.18

$02-0062 0.62 0.65 15.7 16.5 0.59 15.0 003 0.8 u 1/2 - 1/4 12.7 - 6.4 0 .11 2.8 2 318 60.3 5 1/16 128.6 6500 446 3/8 3/4 0.40 0.18 S02-0065 0.65 0.68 165 17 3 0.62 157 0 .03 08 u 1/2 - 1/4 12.7- 64 011 28 2 318 603 5 1116 1286 6500 446 3/8 3/4 0.40 0.18 SQ2-0068 068 0.72 17 3 183 0.65 165 0 .03 08 u 1/2 - 1/4 127- 64 011 28 2 318 603 5 1/16 1286 6500 446 3/8 3/4 0.50 0.23

$02-0072 0.72 0.75 18.3 19.1 0.69 17.5 0 .03 0.8 u 1/2 - 318 12.7 - 9.5 0 .13 3.3 2318 60.3 5 1/16 128.6 6500 446 3/8 3/4 0.50 0.23 SQ2-0075 0.75 0.78 19. 1 19.8 0.72 18.3 0 .03 0.8 u 1/2 - 318 12.7 - 9.5 0 .13 3. 3 2 5/16 58.7 5 1/16 128.6 6500 446 3/8 3/4 0.50 0.23 S02-0078 0.78 0.81 198 20 6 0.75 19 1 0 .03 08 u 1/2 - 318 12.7 - 9.5 013 33 2 5/16 58.7 5 1/16 1286 6500 446 3/8 3/4 0.50 0.23 SQ2-0081 0.81 0.83 206 21 1 0.78 198 0 .03 08 u 1/2 - 3/8 127- 9.5 013 33 2 5116 58 7 5 1/16 1286 6500 446 3/8 3/4 0.50 0.23 SQ2-0083 0.83 0.87 21.1 22. 1 0.80 20.3 0 .03 0.8 u 1/2 - 318 12.7 - 9.5 0. 13 3. 3 2 5116 58.7 5 1/16 128.6 6500 446 3/8 3/4 0.50 0.23 S02-0087 0.87 0.90 22.1 22.9 0.84 21 .3 0 .03 0.8 u 1/2 - 318 12.r - 9.5 0 .13 3. 3 2 5/16 58.7 5 1/16 128.6 6500 446 3/8 3/4 0.60 0.27 S02-0090 0.90 0.93 229 236 0.87 22 1 0 .03 08 u 1/2 - 318 12.7

  • 9.5 0 13 33 2 5/16 58 7 5 1/16 1286 6500 446 3/8 3/4 0.60 0.27 Notes:

(1) Standard Seal Materal* U = urethane (2) S02 shans are a stepped design The compressKJO nut is on the larger diameter, an other <XllTlponents are on the smaller diameter.

(3) Spec~icabons subject to chance Wllhout nOl>ce.

W orld H eadquarters:

Expansion Seal Technologies Expansion Seal Technologies EMEA Expansion Seal Technologin Asia Pte Ltd.

27 0 1 Township U na Road Hoom 3 12a

  • 2404 HL Alphen aan den Rijn 35 Tannery Rd, # 11-10 Tannery Block Hatfield, PA 19440-1770 USA The Netherlands Ruby Industrial Complex Tel: 1*215-7 21-110 0 Fax: 1 °215-721-110 1 Tel: +31-172 418841 Singapore 347740 Toll-Free: 1*80D-355-7044 Fax: +3 1-172 - 418 849 Tel: +65-6745-8560 Fax: +65-674 2-8700 SPECIALISTS IN TUBE TESTING. SLEEVING AND PWGGING TECHNOLOGY AN ISC,..9001 REGISTERED COMPANY info@expansionseal.com I www.expansionseal.com

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

~ Q2 High Pressure Test Plug Info

EST Gro up, a business unit of Curtiss-Wright Flow Control Company 2701 Township Line Road, Hatfield, PA 19440 Telephone: (215) 721-1100 800-355-7044 Fax: (215) 721-1101 E-Mail: est-info@curtisswright.com Website: estgroup.cwfc.com SQ2 High Pressure Test Plug I I: I: ,Ir I To fit pipe and tube inside diameters from 0.48" (12.1mm) to 0.92" (23.4mm). Working Pressures to 6,500 psi (466 Bar).

  • ~
  • i1 . l:**- * ~ li ',

~ Notes:

1) Working Pressure determined for testing in ASTM A105 Grade B pipe. Actual working pressure may vary by pipe material.

2)For pipe or tube sizes larger that 0.92" (23.4mm) please see the Grip Tight"* Test Plugs

  • .a ' ,; a .,,

-- = ~

Specifications subject to change without notice.

Results 1 - 15 of 15 Item # Pipe l.D. Pipe l.D. Plug O.D. Plug 0 .0 . Max Operating Max Operating List Price Size Range (in) Size Range (mm) (in) (mm) Pressure (psi) Pressure (Bar)

SQ2-0047 0.47 to 0.5 11.9to 12.7 0.44 11.2 6500 446 QUOTE SQ2-0050 0.5to 0.53 12.71013.5 0.47 11 .9 6500 446 QUOTE SQ2-0053 0.5310 0.56 13.5 to 14.2 0.5 12 .7 6500 446 QUOTE SQ2-0056 0.56 to 0.6 14.210 15.2 0.53 13.5 6500 446 QUOTE SQ2-0060 0.6010 0.62 15.2 10 15.8 0.57 14.5 6500 446 QUOTE SQ2-0062 0.62to 0.65 15.7to 16.5 0.59 15 6500 446 QUOTE S02-0065 0.65to 0.68 16.5to 17.3 0.62 15.7 6500 446 QUOTE SQ2-0068 0.68to 0.72 17.3to 18.3 0.65 16.5 6500 446 QUOTE SQ2-0072 0.72to 0.75 18.3to 19.1 0.69 17.5 6500 446 QUOTE SQ2-0075 0.7510 0.78 19.1 to 19.8 0.72 18.3 6500 446 QUOTE S02-0078 0.78 to 0.81 19.8to 20.6 0.75 19.1 6500 446 QUOTE SQ2-0081 0.81 to 0.83 20.6to 21 .1 0.78 19.8 6500 446 QUOTE SQ2-0083 0.83to 0.87 21.1 to 22.1 0.8 20.3 6500 446 QUOTE SQ2-0087 0.87 to 0.90 22.1 to 22.9 0.84 21 .3 6500 446 QUOTE SQ2-0090 0.90to 0.93 22.9to 23.6 0.87 22.1 6500 446 QUOTE Results 1 - 15 of 15 6/7/2011 I Page 1 of 1

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00012 Version: Approved: r,/

Requested Duedate: 06/09/201 1 Previous Duedate: 02/20/20 11 Requested By: Barborek,W Douglas 02/ 14/201 1 Approved By: Edgell,Douglas W 02/18/201 1 Request

Description:

Th is CA could not be completed by the original due date due to competing work load priorities. The due date is being extended to 6/9/201 1 which is acceptable since it corresponds with Milestone 30 "All Outage Material on Order". WDB 2/ l4/20J J Approved

Description:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 13 G roup Name Assigned By: Eng Sys Mgmt ANO Wi lliams,Patrick J Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: ZzANO CRG **lHEA use only** 11/ 17/2010 16:42:18 Performed By: Edgell,Douglas W 2/16/2011 16:56:41 Subperformed By: Barborck,W Douglas 2/ 16/2011 08:20: 13 Approved By:

Closed By: Williams,Patrick J 2/16/2011 19:42:51 Current Due Date: 02/ 17/2011 Initial Due Date: 02/17/2011 CA Type: PERIODIC REVIEW CA Priority:

Plant Constraint: NONE CA

Description:

Interim and Periodic Review Required (NOTE - an Interim Review requires both "Responsible Manager" AND a Director or Above" approval).

Conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist",

Attachment 9.8 of EN-LI- I 02 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section. Consider any open CAs for Long Term classification per Attachment 9.9 of EN-Ll-102.

Response

l concur with the attached interim review and closure of this corrective action.

S ubresponse :

Interim Review is attached. E-mail documenting Director/Manager approval is also attached. WDB 2/ 16/20 11 Closure Comments:

Attachments:

Subresponse Description Director & Manager Approval ofTR Subresponse Description interim review

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!Director & Manager Approval of IR

BARBOREK, W DOUGLAS From: WILLIAMS , PATRICK J Sent: Tuesday, February 15, 2011 10:08 PM To : BARBOREK, W DOUGLAS Cc: MCCOY, JAIME H; EDGELL, DOUGLAS W

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Flange Piping" I concur as Acting Engineering Director.

Patrick From : BAPBOREK, W DOUGLAS Sent : Tuesday, February 15, 2011 3:55 PM To: WILLIAMS, PATRICK J Cc: MOCOY, JAi ME H; EDGELL, DOUGLAS W Subje ct : RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Range Piping" Patrick, T hanks for the challenge on the L TCA designation. Yes, I believe LTCNLTCR status is appropriate for thjs 2008 CR.

There will be more CA's issued this yem* to implement this l R23 scope, and I think it best to wrap it all u p after the outage with a definitive statement that the OE has finally been addressed for AN0-1.

Issued two more CA's as follows. CA-15 is to assess contingency p.lans in case the pressure test results are not as expected. CA-16 will be the LTCA. I will pursue LTCA approval this week.

CA Per EN-OU-100, identify any required contingency plans for IR23 which are required to support the pressure test of the CCC-6 piping. Issue follow-up CA's as required. CA assigned to SYE. Original due date is 4/28/20 11, which is the finish date for Milestone 24 "Identify Contingency Plans".

CA Following implementation of W0-00195437 during IR23, document that the applicable OE has been adequately addressed for ANO- I . LTCA Classification is being pursued for this CA since RFO 1R23 is required to resolve this condition and since various additional CA's will likely be required to support 1R23 resolution of this issue. CA assigned to SYE. Original due date is 12/15/201 1 whjch should be after the completion of 1R23.

The updated IR is attached to this e-mai l and to CA-13 of the CR. Please let me know if you have any additional questions.

Thanks, Doug Barborek Entergy Operations, Inc. / Arkansas Nuclear One System Engineer - Unit 1 Reactor Coolant System System Engineering Building/ N-SYE-4 wbarbo I @entcrgy.com 479-858-4337

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<< File: C R-AN0-1-2008-02560 CA- 13 ln1crim Review.doc>>

From : WILLI AMS, PATRI CK J Sent: Monday, February 14, 2011 2:52 PM

To:BARBOREK, WDOUGLAS Cc: MOCDY, JAi ME H; EDGB._L, DOUGLAS W

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Olloride Induced Stress Corrosion Cracking on RV Range Piping" Approved as Acting Engineering Director, Is this a candidate for LTCA?

Patrick From : BARBOREK, W DOUGLAS Sent: Monday, February 14, 2011 1:56 PM To: MCCOY, JAIMEH; WILLIAMS, PATRICKJ Cc: EDGELL, DOUGLAS W; El Q-lENBERGER, JOHN R

Subject:

REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on O,loride Induced Stress Omosion Cracking on RV Range Piping" Jaime & Patrick, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-13 of the subject CR. The CA is ultimately due on Thursday, February 17, 2011 (initially due by me on Tuesday, February 15).

Please let me know if you have any questions.

Thanks, Doug Barborek Entergy Operations. Inc./ Arkansas Nuclear One System Engineer - Unit I Reactor Coolant System System Engineering Building/ N-SYS-4 wbarbo l @cmergy.com 479-858-4337

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<< File: CR-AN0-1-2008-02560 CA-13 Interim Review.doc>>

2

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!interim review

ATTACHMENT 9.8 CR INTERIM AND PERIODIC REVIEW FORM SHEET 1 OF 1 CR Interim and Periodic Review CR Number: CR-AN0-1-2008-02560 Category Level D AD B IZI C CR Owner Group: ENG SYS MGMT ANO CR

Description:

1R2 1 W0-001 02463 Task 01 was not able to be successfully performed as planned.

CR-AN0-1-2005-0 1140 was written to document OE from several plants which identified corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/pressure test connection piping resulting from water left in the piping following refueling outages. AN0-1 was determined to be vulnerable since water has been trapped in this piping during previous operating cycles. The scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utilize a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV and the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned. Since CR-AN0-1-2005-0 1140 was closed to W0-001 02463, this CR was initiated to provide a means of tracking an alternate resolution for addressing the noted OE. As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained during refueling outages prior to installing the RV head, thus mitigating the damage mechanism. The operability statement in CR-AN0-1-2005-01140 remains applicable.

CR Review: (All No responses require explanation be included.)

The following CA's have been issued to further evaluate a course of action required to address the OE documented in CR-AN0-1-2005-01 140. The issuance of these actions constitutes an acceptable corrective action plan for this CR. Additional CA's will be issued as deemed appropriate.

CA-02 [closed] - CR-AN0-1-2005-01140 (CA-04) evaluated the feasibility of performing a hydrostatic test of the RV flange leak-off/pressure test connection piping and concluded that such a test was not feasible. Based on the inability to visually inspect the piping during 1R21 (via boroscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity. This CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization.

CA-03 [closed] - Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e . removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping. This CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPDB Record and present to the URT/MP RC to recommend this scope of work be executed by the Projects Organization.

CA-07 [closed] - The inspection of the RV Flange Leak-off & Pressure Test Connection piping requires additiona l resources and engineering which transcends the role and responsibility of System Engineering.

Initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization. SIPDB Record 4955 was initiated to transfer this scope of work to the Projects Organization and this CA was closed on 11/19/2009. New CA-08 was issued to track the presentation of SIPDB Record 4955 to the URT.

CA-08 [closed] - Present SIPDB Record (Inspection of the RV Flange Leak-off & Pressure Test Connection Piping) to the URT/MPRC for this scope of work to be executed by the Projects Organization. SYE made the presentation at the 12/7/2009 URT meeting. The URT concurred t hat if the leak-off & pressure test connection piping ultimately require physical inspection or replacement such that disassembly/reassembly of the reactor cavity seal plate, concrete shield blocks, and reactor vessel insulation is required, then that significant scope of work would be transferred to the Project Organization. However, at this time, the URT decided that System EN-Ll-102 REV 13

Engineering should re-evaluate the pressure test option and come up with a plan to perform a pressure test on the leak-off and pressure test connection piping to verify the integrity of the piping. If the pressure test fails on one or both of the lines, the ensuing scope of work to access the piping for ultimate resolution of the issue would then be pursued by the Projects Organization. New CA-09 has been issued to SYE to determine a suitable pressure test method and to initiate the proper implementation documentation. 1R22 will be utilized to perform scoping walkdowns since this piping is not accessible during power operations. This CA was closed on 1/2 1/2010.

CA-09 [closed] - Coordinate with EP&C personnel and determine a suitable pressure test method for the RV flange leak-off/pressure test connection piping, and initiate the proper implementation documentation. Utilize 1R22 to perform scoping walkdowns as required. Initiate other corrective actions as required. Completed by SYE on 10/25/2010. CA-10 was initiated for SYE to assign followup corrective actions as required to implement the pressure test plan in 1R23 via W0-00195437. CA-10 was completed on 11/17/2010 by SYE and CA-11 and CA-12 were initiated as a result.

CA Develop EC to support W0-00195437. If required, provide justification for installation of couplings on the affected CCC-6 piping. Also, if required, provide engineering requirements for FME controls associated with the pressure test equipment and components to be used in close proximity to the RV flange. CA assigned to SYE, o riginal due date was 2/20/201 1. However, the current plan is to pressurize against closed valves (RBS-1 and RC-5) in lieu of severing the piping above the valves and installing hydro plugs. As such, the use of couplings is a contingency to be used only if pressurizing against the valves is unsuccessful. The due date is being changed to 7/21/201 1, corresponding with Milestone 48 "All Contingency Plans Approved'.

CA Determine the appropriate type & size of hydro plugs to be utilized for the pressure test (W0-00195437) outlined in CA-09. Initiate actions to procure the hydro plugs, as required. CA assigned to SYE, original due date was 2/20/201 1. Due date extended to 6/9/2011 to correspond with Milestone 30 "All Outage Material on Order'.

CA Provide input to Outage P&S to ensure 1R23 W0-00195437 is properly planned for performance of the CCC-6 piping pressure test. CA assigned to SYE. Original due date is 4/1 4/2011 , which is the finish date for Milestone 20 "All Work Order Tasks Planning Complete".

CA Per EN-OU-100, identify any required contingency plans for 1R23 which are required to support the pressure test of the CCC-6 piping. Issue follow-up CA's as required. CA assigned to SYE. Original due date is 4/28/201 1, which is the finish date for Milestone 24 "Identify Contingency Plans".

CA Following implementation of W0-00195437 during 1R23, document that the applicable OE has been adequately addressed for AN0-1. L TCA Classification is being pursued for this CA since RFO 1R23 is required to resolve this condition and since various additional CA's will likely be required to support 1R23 resolution of this issue. CA assigned to SYE. Original due date is 12/15/201 1 which should be after the completion of 1R23.

1. Will the existing corrective actions documented in the condition report, when completed, correct the condition report issue? Yes~ / No D Implementation of 1R23 W0-001 95437 will resolve this issue, assuming no degradation is discovered. Since additional CA's are expected between now and 1R23, the CR will remain open through 1R23 and will be closed following completion of CA-16.
2. What is the expected CR Closure date based on remaining needed actions? DATE: 12/15/2011 It is anticipated that this CR will be closed following 1R23 implementation of the pressure test and completion of CA-16. Target date is 12/15/201 1. If degradation is found, a new CR will be initiated during 1R23.
3. Is the previously documented operability/functionality position still valid for the current condition and expected to remain valid until CR closure? Yes ~ / No D /N/A D If the answer is NO, then initiate a new CR to document the concern; CR# N/A EN-Ll-102 REV 13
4. Are all Ll-102 requirements for corrective action administration and control being met, i.e.

justifications for Due Date Extensions valid, Long Term Corrective Actions identified, CARB approved CAPRs identified, and appropriate approvals obtained for all?

Yes [gj/ No D

5. What risk to plant operation is imposed by the condition identified and how is risk reduced to an acceptable level for the duration of the action plan?

This CR does not specifically identify a known degraded plant system, structure, or component at AN0-1 , but does identify a potential degradation mechanism. Both the RV flange gasket leak detection line and pressure test connection lines were flushed/drained during 1R19, 1R20, 1R21 and 1R22, thus minimizing the potential for failure resulting from ID initiated, chloride induced, transgranula r stress-corrosion cracking (TGSCC). No RV flange gasket leakage was identified during plant heat-up from 1R19, 1R20 , 1R21 or 1R22 prior to isolating valves R BS-1 and RBS-2, and RCS leakage rates following 1R19, 1R20, 1R21 and 1R22 have not indicated the presence of any concurrent leakage of the RV flange gaskets and leak detection/pressure test connection piping/components.

The RV f lange gaskets are replaced during each refueling outage and valves RBS-1 and RBS-2 are not isolated until just prior to criticality after it has been verified that the inner gasket is not leaking. Therefore, the likelihood of inner gasket leakage, wh ile possible, is considered low.

In conclusion, the risk to plant operation is minimal and reduced to acceptable levels until the action plan for this CR can be implemented during 1R23.

Review/ Approval Required:

Director/GM

Title:

Date:- - - - -

(Print name & Position title)

NOTE: The expectation is to capture the discussion points of this form in a CA. The form itself need not be used, but all points applicable must be addressed.

EN-Ll-102 REV 13

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00013 Version: Approved: r,/

Requested Duedate: 02/17/201 1 Previous Duedate: 02/02/20 11 Requested By: Barborek,W Douglas 02/01/2011 Approved By: Edgell,Douglas W 02/01 /201 1 Request

Description:

Due to competing work load assignments, no work has been performed on resolving the issue in this CR since early December 2010, and the CA plan needs revision in conjunction with perfonning an accurate interim review. The due date is being extended two weeks to 2/17/201 I. This CA is administrative in nature and extension of the due date by two weeks does not impact the operability or functionality of any installed plant SSCs; and is therefore acceptable. WDB 2/1/201 1 Approved

Description:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 14 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 2/15/201 1 17:54:12 Performed By: Barborek, W Douglas 5/5/201115:22:19 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 5/5/2011 15:52:08 Current Due Date: 05/05/2011 Initial Due Date: 05/05/2011 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Provide input to Outage P&S to ensure 1R23 W0-00195437 is properly planned for performance of the CCC-6 piping pressure test.

Response

Based on a meeting with Outage Management on Monday, May 2nd, 201 1, W0-00195437 to perform the RV flange piping pressure testing is going to be deferred from I R23 to IR24 due to the fact that inadequate time and resources exist to develop the necessary contingency and repair actions required in the event piping degradation is discovered by the testing.

As such, this CA is no longer required to support 1 R23 and is being closed.

Note that S IPD Record 5678 has been init iated to request that the URT again consider reassignment of this issue from System Engineering to Project Management for IR24 implementation. CA-17 has been issued to SYE to present this SIPD Record to the URI.

WDB 5/5/2011 S ubresponse :

Closure Comments:

1 concur with closure oftbis corrective action.

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00014 Version: Approved: r,/

Requested Duedate: 05/05/2011 Previous Duedate: 04/ 14/20 11 Requested By: Barborek,W Douglas 04/13/201 1 Approved By: Edgell,Douglas W 04/13/201 1 Request

Description:

Cannot complete by assigned date. Extend to 5/5/201 1 which is Milestone date for P0-20 "All Outage Work Order Task Planning Complete". Since the due date is before the milestone, the DDE is acceptable. WDB 4/13/20 11 Approved

Description:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 15 G roup Name Assigned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 2/15/201 1 17:54:56 Performed By: Barborek, W Douglas 5/18/20 11 15:25:30 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 5/19/2011 17 :03:03 Current Due Date: 05/ 19/2011 Initial Due Date: 05/19/201 1 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Per EN-OU-100, identify any required contingency plans for 1R23 which are required to support the pressure test of the CCC-6 piping. Issue follow-up CA?s as required.

Response

Based on a meeting with Outage Management on Monday, May 2nd, 201 1, W0-00195437 to perform the RV flange piping pressure testing is going to be deferred from I R23 to I R24 due to the fact that inadequate time and resources exist to develop the necessary contingency and repair actions required in the event piping degradation is discovered by the testing.

As such, this CA is no longer required to support 1R23 and is being closed.

Note that SIPD Record 5678 has been initiated to request that the URT again consider reassignment of this issue from System Engineering to Project Management for IR24 implementation. CA-17 has been issued to SYE to present this SIPD Record to the URI.

WDB 5/ 18/201 1 S ubresponse :

Closure Comments:

1 concur with closure oftbis corrective action.

Entergy I CA DUE DATE EXTENSION I CR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00015 Version: Approved: r,/

Requested Duedate: 05/19/2011 Previous Duedate: 04/28/2011 Requested By: Barborek,W Douglas 04/28/201 1 Approved By: Williams,Patrick J 04/28/201 1 Request

Description:

Due to incomplete and ongoing discussions with Outage Management, th is CA must be extended. A meeting is scheduled for Monday, May 2nd to discuss contingency actions associated with the 1R23 RV flange piping pressure test. This CA is being extended 3 weeks to May 19th, 20 1 J . Per Section 5.8[1] ofEN-OU- 100, JR23 Milestone P0-24 "Identify Contingency Plans" {due today) applies to Business Case Plans only. Since the pressure test is not associated with a Business Case Plan, extension of the due date to 5/ 19/201 1 is acceptable. WDB 4/28/201 1 Approved

Description:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 16 G roup Name Assigned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems NSSS Staff ANO Barborek, W Douglas Subassig ned To :

Originated By: Edgell,Douglas W 2/1 5/2011 17:53:14 Performed By:

Subperformed By:

Approved By:

Closed By:

Current Due Date: 11 / 17/2016 Initial Due Date: 11 /18/201 6 CA Type: GENERAL ACTION CA Priority: 4 Plant Constraint: NONE CA

Description:

Fol lowing implementation of W0-00 L95437 during IR23, document that the applicable OE has been adequately addressed for ANO-I.

Response

r Subrcsponse :

Closure Comments:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00016 Version: Approved: r,/

Requested Duedate: 05/30/2013 Previous Duedate: 12/ 15/20 11 Requested By: Barborek,W Douglas 12/12/2011 Approved By: Barborek,W Douglas 12/1 2/201 1 Request

Description:

The pressure test to be performed under W0-00 195437 was deferred from I R23 to IR24 via SCR-1 1049 (attached to DDE).

As such, this CA must be extended to beyond 1R24. As such, the CA is being extended to May 30, 2013. This CA is administrative in nature, so extended the date is both necessary and acceptable. WDB J2/ 12/20 l l Approved

Description:

Note that CA-20 has already been classified as an LTCA. Per discussion with Bob Eichenberger, only one CA has to be classified as an LTCA to properly classify the CR as a LTCR. ODE # l is necessary and acceptable. Doug Barborek (for Doug Edgell). WDB l 2/l 2/20 I l Attachments:

Request Description IR23 SCR l 1049

Attachment Header Document Name:

untitled Document Location

!Request Description Attach

Title:

11R23 SCR 11049

Scope Change Request Form Originator Section Scope Change#: 11049 Chan.ge Type: Deferral Date Initiated: 6/9/2011 Work Request: 0 Work Order: 195437 Component#: R-1 Component Noun Name: Reactor Vessel Reques tor: BAR BOREK, WILLIAM DOUGLAS Phone #: 4337 Discipline: MECH ER#: Req'd Mhrs: 0 Dose Est: 0 Est Cost: $0.00 Ta 1out Re 'd.: n Reference open CR-ANO- I -2008-02560. The pressure test of the reactor vessel leak-off and pressure test connection pi ping requires significant contingency planning which cannot be accomplished by System Engineering prior to I R23. Potential contingency plans include temporarily plugging (for one cycle) the leak-off port(s) associated with any piping found degraded, an involves a long term (i.e. next RFO) fix of removing the reactor cavity shield plate, concrete shield blocks, and RV insulation to repair/replace or cap the piping if found degraded by the pressure test. Request deferral to 1R24. SIPD Record 5678 has been developed to request transfering the scope of this pressure test and the development of the significant contingency planning action.

from S stem En ineerin to Pro'ect Mana ement for 1R24 im lementation.

Reason for Scope Change submittal: Future Outage- later determination made that work sh Justification for Scone Change:

Although the pressure test must ultimately be conducted to confirm that there is no cloride-induced stress corrosion cracking in tht RV leak-offYpressure test connection piping (as seen at other plants), it is noted that there is no known degradation in the subject ANO- L piping at this time. The leak-off & pressure test connection piping is currently (since IR 19) drained at the end of each refueling outage, thus mitigating the damage mechanism and eliminating any pressurization due to the heating of water trapped i1 the piping during power operations, assuming no inner o-ring leakage. Leakage of the RV nange inner o-ring is an infrequent occurance. CA-19 of CR-ANO-l-2008-02560 has been issued to SYE to justify and enact a procedure revision to change the configuration of leak-off isolation valves RBS- I and RBS-2 to "normally open" for Cycle 24 which will eliminate the possiblility of pressuring the CCC-6-1" piping in the event of an inner o-ring leak until such time the pressure test can resolve the industry OI The original design basis configuration for RBS-L and RBS-2 was "normally open" during power operationsuntil being changed ir the mid- I990's. With RBS-1 and RBS-2 open during Cycle 24, the CCC l " piping will not be pressurized in the event of an inner o-ring leak which will remove any pressure-induced challenges to the piping.

Criteria for approval of outage scope additions [after Scope Free1,e (T-11) and during outage execution]:

Check if this SCF has ANY Scaffold/Insulation work impact? D Originator is responsible for completing any forms required for progra1mnatic changes and submiuing those per the instructions on the specific forms (Forms available on IDEAS):

Is a PMDR required (EN-DC-324, Att 9.3)? 0 Is a ST Change (1000.009A) Required? 0 Change Review Section Radiation Protection Does RP approve this change request? Yes ~ No [J Name: STOLTZ, DANIEL CLIFF Date: 6/20/2011 Comments: No issue with deferral. Work Order still in PLAN status, dose estimate and work scope not defined at this time.

Prelim inar dose estimate is 250 mrem.

For Scope Deferrals from outage to on-line ma intenance; it is required that the OPS OWL has developed/reviewed the impact statement for the work to be performed and obtained concurrence from the OPS Manager, or his designee, that on-line maintenance risk is acceptable.

O erations: Does OPS approve this change request? Yes ~ No D Name: STUMBAUGH, STEVEN ARNOLD Date: 6/17/2011 Conunents: Deferral of the rerssure test is not an issue with O s.

For Scope Deferra ls thal require a PMDR, veri fy the PMDR has been approved prior to aulhorizing the scope change. Evaluate the results for ,

deferral, if a Condition Report been closed to this item?

En ineerin : Does Engineering approve this change request? Yes ~ No D Name: EDGELL, DOUGLAS WARREN Date: 6/9/2011 Comments: SYE recommends deferral of this WO. Although the pressure test must ultimately be conducted to confi rm that there is no cloride-induced stress corrosion cracking in the RV leak-oftYpressure test connection piping (as seen at other plants), it is noted that there is no known degradation in the subject ANO- I piping at this time. The leak-off pressure test connection piping is currently (since lRl 9) drained at the end of each refueling outage, thus mitigatin, the damage mechanism and elimjnating any pressurization due to the heating of water trapped in the piping during power operations, assuming no inner o-ring leakage. Leakage of the RV fl ange inner o-ring is an infrequent occurance. CA-19 of CR-AN0-1-2008-02560 has been issued to SYE to justify and enact a procedure revision to Monday, December 12, 2011 Page 1 of 2

Scope Change Request Form change the configuration of leak-off isolation valves RBS-I and RBS-2 to "normally open" for Cycle 24 which will eliminate the possiblility of pressuring the CCC-6-1" piping in the event of an inner o-ring leak until such time the pressure test can resolve the industry OE. The original design basis configuration for RBS- l and RBS-2 was "normally open" during power operationsuntil being changed in the mid- 1990's. With RBS-1 and RBS-2 open during Cycle 24, the CCC 1" piping will not be pressurized in the event of an inner o-ring leak which will remov an ressure-induced ch alien es to the i in"*

Maintenance: Does Maintenance approve this change request? Yes ~ No D Name: STUMBAUGH, STEVEN DALE Date: 6/20/2011 Comments: JAgree with deferral based on Eng. comments.

lf this SCF is dealing with a o n-line leak repair then verify the requirements of 1025.015 "On-Line Leak Repair" are satisfied prior to deterring work or disapproving addition.

Final Approval Section Note: Outage Manager may bypass review learn rccommcndalions and approve/reject for Minor Changes. Duri ng outage execution, the Shift Outage Manager may perfor m this function . The GMPO must approve any Major C hange (An Activi ty> $100,000, or which would result in Outage Extention).

Does Outage Mgt approve this change request? Yes PT No 0 Name: WALTERS, JOE RANDALL Date: 6/20/2011 Comments: !Move to 1R24 when engineering has developed suitable contingency actions. Moved WO rask to 1R24. DC Offical Date Approved: 6/20/201 1 Has Schedule been Updated? ~ PMDR approved and complete D Has MPC/OSG/RP been notified of addition/deferral O ST approved and complete? O Monday, December 12, 2011 Page 2 of 2

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00016 Version: 2 Approved: r,/

Requested Duedate: 12/04/2014 Previous Duedate: 05/30/20 13 Requested By: Barborek,W Douglas 05/21/2013 Approved By: Woodson P.E.,Timothy R 05/22/2013 Request

Description:

The pressure test of the RV leak-off lines was deferred to l.R25 by management due to lack of contingency planning (requires AREY A support, and funding for that support) and lack of resources to plan and execute the work. This CA is being extended to J2/4/20 14 to correspond to a date expected to be beyond 1R25 to allow time for WO implementation. This DOE is necessary since the work has been deferred by management. WDB 5/2 I/2013 Approved

Description:

ODE approved by Engineering Directo r per attached email.

Attachments:

Approved Description Director approval ofDDE

Attachment Header Document Name:

untitled Document Location f pproved Description Attach

Title:

!Director approval of DOE

BARBORB<, W DOUGLAS From: MCCDY, JA.IM EH Sent: Tuesday, May 21 , 2013 9:34 PM To : BARBOREK, W DOUGLAS Cc: B)GaL, DOUGLASW; WOODSJN , llMOTHYR; LAY, LINDA S SUbject: RE: PEQUEST FOR APPFOVAL - DDE#2 & DDE#1 for CA- 16 & CA-20 of CR-AN0- 1-2008-02560 (Pressure Test of RV Leak-off piping to resolve 2002 Davis Besse OE)

I approve both.

I would like to have a meeting with outage management, systems, and projects to make sure funding is arranged. Linda - please schedule a meeting next week or week after.

Thanks, Jaime Sent with Good (www.good.com)

Otiginal Message-----

From: BARBOREK, W DOUGLAS Sent: Tuesday, May 21, 2013 09:28 PM Central Standard Time To: MCCOY, JAIME H Cc: EDGELL, DOUGLAS W ; WOODSON, TIMOTHY R

Subject:

REQUEST FOR APPROVAL - DOE #2 & DOE #1 for CA-16 & CA-20 of CR-AN0-1-2008-02560 (Pressure Test of RV Leak-off piping to resolve 2002 Davis Besse OE)

Jlime, CA.-16 and CA.-20 of CRAN01-2008-02560 are due on 5/29/2013 and assumed we would perform the pressure test of the RV leak-off lines in 1R24. This work was once again deferred to 1R25due to the lack of rontingency planning, which requires funding for AFB/A support and requires det ermi nation of who wi11plan and perform the testing and rontingendesduring 1R25 (A'oject Management hasdedined to take t his work). I am submitting DDE#2 for CA.-16 and DDE:#1 for CA.-20for 12/4/ 2014 (beyond end of 1R25.. per ament cyde schedule). CA.-20 isan LTCA.and requires your approval.

CA. Following implementat ion of W0-00195437 during 1R23, dorument that the applicable OEhasbeen adequately addressed for AN0-1.

CA [Note: All LTCA. DDEs require G'v1F0'Dredor or Above approval. lam 12/03/ 12] 8lsure W0-00195437 is implemented in 1R24. This WO has been deferred to 1R24 asdorumented in the attached 1R23 g:R 11049. Issue additional rorrective actions as required to support 1R24 rerolution of this issue. This CA is being dassif ied as a LTCA..

DDE~uest for both CA's- lhe pressure test of the FN leak-off lines was deferred to 1FQ5 by management due to lack of contingency planning (requires AFB/A rupport, and funding for that rupport) and lack of reoources to plan and executethework. This CA is being extended to 12/4/2014 to correspond to a date expected to be beyond 1FQ5 to allow time for WO implementation. This DOE is necessary since the work has been deferred by management. WDBS/21/2013 Rease let me know if you have any questions regarding this request.

Thanks, Doug Barborek Entergy Q:,erations. Inc. / Arkansas Nudear One S,,stem Engineer - Unit 1 A:lactor ())olant S,,stem S,,stem Engineering Building / N-~

wbarbo1@entergy.com 479-858-4337

!(b )(6) Ipager From : woooroN, Tl MOTHY R Sent : Tuesday, May 21, 2013 5:49 Pfv1 To:BAA30REK WDOUG...AS Cc: <?ASTON, KERRY Subj ect: RE: CA's for RV leak-off line testing

~nd an email to J3mie requesting the DDE, procedure requires DDEapproval from a Drector/ ClvlA)for an LTCA. Cbpy l:x:>ug Edgel I.

Thanks, Tim From : BARBORB<, W DOUGLAS Sent: Saturday, May 18, 2013 11 :42 PM To: woooroN, Tl MOTHY R Cc: GA.SfON, KERRY

Subject:

CA's for RV leak-off line testing

Tim, CA-16 and CA-20 of CRAN0-1 -2008-02560 are due on 5/29/2013 and assumed we would perform the pressure test of the FN leak-off lines in 1FQ4. Cbviously, this has been deferred again to 1FQ5. Are we still 'dS9Jming 1FQ5will ocx::ur in Fall 2014 (after the cyde delay due to the stator drop) regarding due date extensions?

Not knowingtheoutagetimeframefor 1FQ5, I wasgoingto extend them to 12/1/ 2014 if you concur. If we are going to change the cyde schedule, should I move them out even further? CA-16 currently has one DDEand CA-20 has none.

Thanks ,

Doug Barborek 2

Entergy ~rat ions, Inc. / Arkan= Nudear Oie S,,stem Engineer - Unit 1 A3actor Coolant S,,stem S,,stem Engineering Building / N-SYE--4 wbarbo1@entergy.com

~ ~r 3

Entergy I CA DUE DATE EXTENSION I CR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00016 Version: 3 Approved: r,/

Requested Duedate: 11 / 18/2016 Previous Duedate: 12/04/20 14 Requested By: Barborek,W Douglas 12/03/2014 Approved By: Edgell,Douglas W 12/04/2014 Request

Description:

The pressure test of the RV leak-off lines was deferred to l.R26 by plant management due to a lack of contingency planning (rnquires AREY A support, and funding for that support) and a lack of resources to plan and execute the work. See attached IR25 SCR-15969 which approves the deferral from lR25 to I R26. T his CA is being extended to l J/J 8/20 16 to correspond to a date expected to be beyond I R26 to allow time for WO implementation. This DDE is necessary since the work has been deferred by management. WDB 12/3/ 2014 Approved

Description:

I concur with this DOE. The requested date support implementation during the next refueling outage.

Attachments:

Request Description l R25 SCR-15969 deferral

Attachment Header Document Name:

untitled Document Location

!Request Description Attach

Title:

11R25 SCR-15969 deferral

Scope Change Request Form Originator Section Scope Change #: .15.969 _ Change Type: __J)eferr.aL Date Initiated: 7/8/2014 Work Request: 0 Work Order: 195437 Component#: R-1 Component Noun Name: R-1 RX vessel leak off lines Requestor: SKARTVEDT, MARK Phone #: ___+/-6+/-2...._ Discipline: ____MNT_L ER#: Req'd. l\1hrs: _Q_ Dose Est: _Q_ ___ Est Cost: __$Q,OO Tagout Re 'd.: D This is to defer the reactor vessel head 0 -rin leak off line h dro that is scheduled for .1 R25 until IR26.

Reason for Scope Change submittal: Future Outage- later determination made that work sh Justification for Seo Chan e:

Criteria for approval of outage scope additions [after Scope Freeze (T-11) and during outage execution]:

Check if this SCF has ANY Scaffold/Insulation work impact? D Orig inat or is respo nsible for completin g a ny form~ required for progr a mmatic cha nges and submitting those per the instr uctions on the specific forms (Forms available on IDEAS) :

Is a PMDR required (EN-DC-324, Att. 9.3)? 0 Is a ST Change (1000.009A) Required? 0 Change Review Section Radiation Protection Does RP approve this change request? Yes ~ No [l Name: Reynolds, Ryan Date: 7/9/2014 Conunents: !RP approves this deferral.

For Scope Deferrals fro m outa ge to on-line maintenance; it is requ ired that the OPS O\.VL has developed/reviewed the impact sta tement for the work to be performed and obtained concurrence from the OPS Manager, or his desig nee, that on-li ne ma intena nce risk is accepta ble.

Operations: Does OPS approve this change request? Yes ~ No 0 Name: HILL. SJEYEN.n. Date: 7/10/2014 Conunents: ! not required, can be deferred For Scope Deferra ls tha t require a PMDR, verify the PMDR has been approved prio r to a uthori zing the scope cha nge. Evaluate the results for :

deferral , if a Condition Report been closed to t his item?

Engineering: Does Engineering approve this change request? Yes ~ No D Name: WOODSON. TIMOTHY ROY Date: 7/8/2014 Conunents: CR-AN0-1-2008-02560 is tracking OE generated issue where other B&W NSSS plants have had fa ilures of these lines. ANO Unit l has a reasonable p robability of one of these two lines failing. AJso reference SlPD 5678 for funding requests and requests to obtain a Project Manager to prepare the test and contengency plan to address failure. The plan would include plugging both lines at the reactor vessel for one cycle with rolled plugs and then correct the condition in the next refueling outage wh ich would require construction work to obta.in access to the lines. Resources do not exist to perform this work in IR25 and the risk still exists, but the work cannot be carr.i ed out in 1R25 .

Maintenance: I Does Maintenance approve this change request? Yes ~ No [1 Name: SKAR.TVEDT.~MARK Date: 7/1012014 Conunents: ~ln_o_c_o_m_m ~en_t~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~--

Ifthi s SCF is dealing with a on-line leak repa ir then verify the requirements o f 1025.0 15 "On-Line Leak Repa ir" are satisfied prior to deferring wo rk or disa pproving a ddition.

Final Approval Section Note: Outage Manager may bypass review team recommendations and approve/reject for Minor Changes. During outage execution, the Shift Outage Manager may perform this function. The GMPO must approve any Major Change (An Activity> $ 100,000, or which would r esult in Outage Extention).

Does Outage Mgt approve this change request? Y~ No Name: SKARTVEDT. MARK Date: ll0.120 I4 Comments: cannot meet mi lestone, Engei nnering resources wi ll not support completing this project for 1R25. **Moved to 1R26 OWL**

Wednesday, December 03, 2014 Page 1 of 2

Scope Change Request F onn Oflical Date Approved: 8/13/2014 Has Schedule been Updated? ~ PMDR approved and complete D Has MPC/OSG/RP been notified of addition/deferral D ST approved and complete? D Wednesday, December 03, 2014 Page2 of 2

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 17 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS Staff ANO Barborek, W Douglas Subassig ned To :

Originated By: Barborek, W Douglas 5/5/201 1 15:55:21 Performed By: Edgell,Douglas W 12/7/2011 11 :49:05 Subperformed By:

Approved By:

Closed By: Edgell,Douglas W 12/7/2011 11 :49:05 Current Due Date: 12/08/2011 Initial Due Date: 12/08/2011 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Present S1PD Record 5678 to the URT to request reassignment of the RV Flange leak-off/pressure test connection piping pressure test (W0-00195437) from System Engineering to Project Management for IR24 implementation. Issue follow-up CA's as required.

Response

This corrective action is no longer needed. This issue have been presented to the URT previously as documented in CA # 8.

The scope of work is approved outage scope for 1R24. Vent port inspection were completed during lR23 to support contingency plans for a I R24 execution. Re-assignment of this project execution and the associated contingency plans will be made by the site lead team. This corrective action is ready for closure.

Subresponse :

Closure Comments:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action: CR-AN0-1-2008-02560 CA-00017 Version: Approved: r,/

Requested Duedate: 07/21/2011 Previous Duedate: 06/09/20 11 Requested By: Barborek,W Douglas 06/08/201 1 Approved By: Edgell,Douglas W 06/08/201 1 Request

Description:

Additional time is required to schedule the URT presentation with the ORT coordinator. The due date is being extended to 7/21/2011. Since this work is being deferred from IR23 to IR24, extension of the due date is acceptable. WDB 6/8/201 1 Approved

Description:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action: CR-AN0-1-2008-02560 CA-00017 Version: 2 Approved: r,/

Requested Duedate: 12/08/2011 Previous Duedate: 07/21 /20 11 Requested By: Barborek,W Douglas 07/19/201 1 Approved By: Williams,Patrick J 07/19/201 1 Request

Description:

Due to higher URT priorities and URT coordinator turnover, this SIPD Record has not been scheduled for discussion.

Additional time is required to schedule the URT presentation with the URT coordinator. The due date is being extended to 12/8/201 J. Since this work is being deferred from lR23 to J R24, extensio n of the due date is acceptable and should provide adequate time to schedule and present. WDB 7/9/2011 Approved

Description:

Approved.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 18 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Barborek, W Douglas 6/7/2011 14:42:45 Performed By: Barborek, W Douglas 2/22/20 12 15:26: 15 Subperformed By:

Approved By:

Closed By: Barborek,W Douglas 2/23/2012 16:0 1:33 Current Due Date: 02/23/2012 Initial Due Date: 02/23/2012 C A Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Develop CATlDs for the test plugs specified in CA-12. lssue additional corrective actions as required to ensure material is pledged to W0-00 195437.

Response

CA REQUEST:

Develop CAT1Ds for the test plugs specified in CA- I 2. lssue additional corrective actions as required to ensure material is pledged to W0-00 195437.

CA RESPONSE:

The hydro test plugs are considered test equipment and do not require development of a CATID in order to be procured to support the IR24 pressure test. No action is required and this CA may be closed. WDB 2/22/2012 Subresponse :

Closure Comments:

CA is ready for closure. Doug Barborek 2/23/2012

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00018 Version: Approved: r,/

Requested Duedate: 02/23/2012 Previous Duedate: 08/04/20 11 Requested By: Barborek,W Douglas 08/03/201 1 Approved By: Edgell,Douglas W 08/03/201 1 Request

Description:

This CA could not be completed by the assigned due date and must be extended. ft is likely that this CA will not be required since the hydro test plugs may not require development of a CATID in order to be procured to support the I R24 pressure test.

However, the CA is be ing extended to 2/23/20 l2 in the event a CATIO is required. Extension of this CA to the new date does not affect the ability to perform the RV flange leak-off piping pressure test in I R24 and is therefore acceptable. WDB 8/3/2011 Approved

Description:

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 19 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Barborek, W Douglas 6/22/2011 08:35:19 Performed By: Barborek, W Douglas I0/20/201 1 21 :54:48 Subperformed By:

Approved By:

Closed By: Barborek,W Douglas 10/20/2011 21 :54:48 Current Due Date: I0/20/2011 Initial Due Date: I 0/20/2011 CA Type: ACTION CA Priority:

Plant Constraint: NONE CA

Description:

Submit Plf to Operations to revise the applicable plant operating procedures to change the normal operating configuration of RV Flange Leak-off Isolation Valves RBS-I and RBS-2 from "normally closed" to "nonnally open" prior to Cycle 24.

"Normally open" is the original design basis coafi!,'llration (ref. PEAR-93-0246). Issue corrective action(s) to Operations as required to enact procedure change prior to lR23. Reference lR23 SCR 11049 and SIPD Record 5678.

Response

CA REQUEST: Submit PIF to Operations to revise the applicable plant operating procedures to change the normal operating configuration of RV Flange Leak-off Isolation Valves RBS- I and RBS-2 from "normally closed" to "normally open" prior to Cycle 24. "Normally open" is the original design basis configuration (ref. PEAR-93-0246). Issue corrective action(s) to Operations as required to enact procedure change prior to I R23. Reference I. R23 SCR 11049 and SIPD Record 5678.

CA RESPONSE: PIF 1-11-062 1 to OP-1203.012H, PTF 1- 11 -0622 to OP-1015.036, PTF 1-1 1-0623 to OP-1102.015, and PIF 1-11-0624 to OP-1103.002 to change the configuration of valves RBS-I and RBS-2 from normally closed to normally open for Cycle 24 operation have been transmitted to Operations by CA-21 , which will track incorporation of th e PTFs into the applicable procedures. The PIFs were generated by and approved under EC-32271. A copy of each PIF is attached to CA-21. This action is complete. WDB 10/20/20 11 Subrcsponse :

Closure Comments:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00019 Version: Approved: r,/

Requested Duedate: I0/06/2011 Previous Duedate: 09/ 15/2011 Requested By: Barborek,W Douglas 09/15/2011 Approved By: Edgell,Douglas W 09/15/201 1 Request

Description:

In order to support changing RBS- I & RBS-2 from normally closed to normally open, an EC will be required since P&ID M-230 sh. I will need to be revised. Additional time (and resources) is required to develop an EC. Since the procedure change is not required until startup from JR23, extension of the due date to 10/6/2011 is acceptable. WDB 9/ 15/201 J Approved

Description:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00019 Version: 2 Approved: r,/

Requested Duedate: I0/20/201 1 Previous Duedate: 10/06/2011 Requested By: Barborek,W Douglas 10/05/2011 Approved By: Williams,Patrick J 10/05/201 1 Request

Description:

In order to support changing RBS- I & RBS-2 from normally closed to normally open, an EC will be required since P&ID M-230 sh. I will need to be revised. Additional time (and resources) is required to develop an EC. Since the procedure change is not required until startup from JR23, extension of the due date to 10/20/20 11 is acceptable. WDB 10/5/201 I Approved

Description:

Approved.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 20 Group Name Assigned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Barborek, W Douglas 7/27/201 1 16:24:23 Performed By:

Subperformed By:

Approved By:

Closed By:

Current Due Date: 11 / 17/2016 Initial Due Date: 11/18/2016 C A Type: CAT C-CORRECT CA Priority: 3 Plant Constraint: NONE CA

Description:

[Note: All LTCA DDEs require GMPO/Director or Above approval. lam 12/03/ 12]

Ensure W0-00195437 is implemented in I R24. This WO has been deferred to I R24 as documented in the attached I R23 SCR 11049. Issue additional corrective actions as required to support 1R24 resolution of this issue. This CA is being classified as a LTCA.

Response

r Subresponse :

Closure Comments:

Attachments:

CA Description 1R23 SCR 11 049 to defer pressure test to 1R24 CA Description LTCA Form

Attachment Header Document Name:

untitled Document Location FA Description Attach

Title:

11R23 SCR 11049 to defer pressure test to 1 R24

Scope Change Request Form Originator Section Scope Change#: 11049 Chan.ge Type: Deferral Date Initiated: 6/9/2011 Work Request: 0 Work Order: 195437 Component#: R-1 Component Noun Name: Reactor Vessel Reques tor: BAR BOREK, WILLIAM DOUGLAS Phone #: 4337 Discipline: MECH ER#: Req'd Mhrs: 0 Dose Est: 0 Est Cost: $0.00 Ta 1out Re 'd.: n Reference open CR-ANO- I -2008-02560. The pressure test of the reactor vessel leak-off and pressure test connection pi ping requires significant contingency planning which cannot be accomplished by System Engineering prior to I R23. Potential contingency plans include temporarily plugging (for one cycle) the leak-off port(s) associated with any piping found degraded, an involves a long term (i.e. next RFO) fix of removing the reactor cavity shield plate, concrete shield blocks, and RV insulation to repair/replace or cap the piping if found degraded by the pressure test. Request deferral to 1R24. SIPD Record 5678 has been developed to request transfering the scope of this pressure test and the development of the significant contingency planning action.

from S stem En ineerin to Pro'ect Mana ement for 1R24 im lementation.

Reason for Scope Change submittal: Future Outage- later determination made that work sh Justification for Scone Change:

Although the pressure test must ultimately be conducted to confirm that there is no cloride-induced stress corrosion cracking in tht RV leak-offYpressure test connection piping (as seen at other plants), it is noted that there is no known degradation in the subject ANO- L piping at this time. The leak-off & pressure test connection piping is currently (since IR 19) drained at the end of each refueling outage, thus mitigating the damage mechanism and eliminating any pressurization due to the heating of water trapped i1 the piping during power operations, assuming no inner o-ring leakage. Leakage of the RV nange inner o-ring is an infrequent occurance. CA-19 of CR-ANO-l-2008-02560 has been issued to SYE to justify and enact a procedure revision to change the configuration of leak-off isolation valves RBS- I and RBS-2 to "normally open" for Cycle 24 which will eliminate the possiblility of pressuring the CCC-6-1" piping in the event of an inner o-ring leak until such time the pressure test can resolve the industry OI The original design basis configuration for RBS-L and RBS-2 was "normally open" during power operationsuntil being changed ir the mid- I990's. With RBS-1 and RBS-2 open during Cycle 24, the CCC l " piping will not be pressurized in the event of an inner o-ring leak which will remove any pressure-induced challenges to the piping.

Criteria for approval of outage scope additions [after Scope Free1,e (T-11) and during outage execution]:

Check if this SCF has ANY Scaffold/Insulation work impact? D Originator is responsible for completing any forms required for progra1mnatic changes and submiuing those per the instructions on the specific forms (Forms available on IDEAS):

Is a PMDR required (EN-DC-324, Att 9.3)? 0 Is a ST Change (1000.009A) Required? 0 Change Review Section Radiation Protection Does RP approve this change request? Yes ~ No [J Name: STOLTZ, DANIEL CLIFF Date: 6/20/2011 Comments: No issue with deferral. Work Order still in PLAN status, dose estimate and work scope not defined at this time.

Prelim inar dose estimate is 250 mrem.

For Scope Deferrals from outage to on-line ma intenance; it is required that the OPS OWL has developed/reviewed the impact statement for the work to be performed and obtained concurrence from the OPS Manager, or his designee, that on-line maintenance risk is acceptable.

O erations: Does OPS approve this change request? Yes ~ No D Name: STUMBAUGH, STEVEN ARNOLD Date: 6/17/2011 Conunents: Deferral of the rerssure test is not an issue with O s.

For Scope Deferra ls thal require a PMDR, veri fy the PMDR has been approved prior to aulhorizing the scope change. Evaluate the results for ,

deferral, if a Condition Report been closed to this item?

En ineerin : Does Engineering approve this change request? Yes ~ No D Name: EDGELL, DOUGLAS WARREN Date: 6/9/2011 Comments: SYE recommends deferral of this WO. Although the pressure test must ultimately be conducted to confi rm that there is no cloride-induced stress corrosion cracking in the RV leak-oftYpressure test connection piping (as seen at other plants), it is noted that there is no known degradation in the subject ANO- I piping at this time. The leak-off pressure test connection piping is currently (since lRl 9) drained at the end of each refueling outage, thus mitigatin, the damage mechanism and elimjnating any pressurization due to the heating of water trapped in the piping during power operations, assuming no inner o-ring leakage. Leakage of the RV fl ange inner o-ring is an infrequent occurance. CA-19 of CR-AN0-1-2008-02560 has been issued to SYE to justify and enact a procedure revision to Wednesday, July 27, 2011 Page 1 of 2

Scope Change Request Form change the configuration of leak-off isolation valves RBS-I and RBS-2 to "normally open" for Cycle 24 which will eliminate the possiblility of pressuring the CCC-6-1" piping in the event of an inner o-ring leak until such time the pressure test can resolve the industry OE. The original design basis configuration for RBS- l and RBS-2 was "normally open" during power operationsuntil being changed in the mid- 1990's. With RBS-1 and RBS-2 open during Cycle 24, the CCC 1" piping will not be pressurized in the event of an inner o-ring leak which will remov an ressure-induced ch alien es to the i in"*

Maintenance: Does Maintenance approve this change request? Yes ~ No D Name: STUMBAUGH, STEVEN DALE Date: 6/20/2011 Comments: JAgree with deferral based on Eng. comments.

lf this SCF is dealing with a o n-line leak repair then verify the requirements of 1025.015 "On-Line Leak Repair" are satisfied prior to deterring work or disapproving addition.

Final Approval Section Note: Outage Manager may bypass review learn rccommcndalions and approve/reject for Minor Changes. Duri ng outage execution, the Shift Outage Manager may perfor m this function . The GMPO must approve any Major C hange (An Activi ty> $100,000, or which would result in Outage Extention).

Does Outage Mgt approve this change request? Yes PT No 0 Name: WALTERS, JOE RANDALL Date: 6/20/2011 Comments: !Move to 1R24 when engineering has developed suitable contingency actions. Moved WO rask to 1R24. DC Offical Date Approved: 6/20/201 1 Has Schedule been Updated? ~ PMDR approved and complete D Has MPC/OSG/RP been notified of addition/deferral O ST approved and complete? O Wednesday, July 27, 201 1 Page 2 of 2

Attachment Header Document Name:

untitled Document Location FA Description Attach

Title:

LTCA Form

ATTACHMENT 9.9 LTCA CLASSIFICATION FORM SHEET 1 OF 1 LTCA Classification Form Long Term CA Classification:

CR Number: CR-AN0-1-2008-02560 CR Owner Group: Eng Sys Mgmt ANO CA Number: 20 LTCA Assigned to Group: SYE-NSSS LTCA Classification (check ONLY one):

~ RFO/FO Req 'd 1 R24 D Mod/Design Change Req'd D NRC Resp. Req'd D Multi-cycle Training Req'd Provide specific details for LTCA classification selected above.

W0-00195437 to perform a pressure test of the RV flange leak-off and pressure test connection piping to resolve OE-15417 "Cracking Identified in Class 1 Reactor Vessel Flange 0-ring Monitor Piping [Davis Besse)"

has been deferred to 1R24 to allow additional time and resources to develop appropriate contingency measures in the event the pressure test fails on one or both leak-off lines. Since the condition described in this CR w ill not be resolved until 1 R24 (Spring 2013), this CR should be designated a LTCR via CA-20.

What risk to plant operation is imposed by the condition identified and how is risk reduced to an acceptable level for the duration of the action plan?

T his CR does not specifically identify a known degraded plant system, structure, or component at AN0-1, but does identify a potential degradation mechanism. Both the RV flange gasket leak detection line and pressu re test connection lines were flushed/drained during 1R19, 1 R20, 1R21 and 1R22, thus minimizing the potential for failure resulting from ID initiated, chloride induced, transgranular stress-corrosion cracking (TGSCC). No RV flange gasket leakage was identified during plant heat-up from 1R19, 1 R20, 1 R21 or 1 R22 prior to isolating valves RBS-1 and RBS-2, and RCS leakage rates following 1 R19, 1 R20 , 1R21 and 1R22 have not indicated the presence of any concurrent leakage of the RV flange gaskets and leak detection/pressure test connection piping/components.

The RV flange gaskets are replaced during each refueling outage and valves RBS-1 and RBS-2 are not isolated until just prior to criticality after it has been verified that the inner gasket is not leaking. Therefore, the likelihood of inner gasket leakage, whil e possible, is considered low.

In conclusion, the risk to plant operation is minimal and reduced to acceptable levels until the action plan for this CR can be implemented during 1 R24.

Explain impact to condition report timeliness.

This CR was anticipated to be closed following 1R23 . However, since the pressure test has been deferred to 1 R24, the deferral will result in an 18 month (i.e. one operating cycle) delay in closure of this CR.

EN-Ll-102 REV 16

Review/ Approval Required:

Director/GM

Title:

Jaime McCoy Director Engineering Date: 07/28/11 (Print name & Position title}

NOTE: The expectation is to capture the discussion points of this form in a CA, DOE request or initial CA assignment as appropriate. The form itself need not be used, but all points applicable must be addressed.

EN-Ll -102 REV 16

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00020 Version: Approved: r,/

Requested Duedate: 12/04/2014 Previous Duedate: 05/30/20 13 Requested By: Barborek,W Douglas 05/21/2013 Approved By: Woodson P.E.,Timothy R 05/22/2013 Request

Description:

The pressure test of the RV leak-off lines was deferred to l.R25 by management due to lack of contingency planning (requires AREY A support, and funding for that support) and lack of resources to plan and execute the work. This CA is being extended to 12/4/20 14 to correspond to a date expected to be beyond 1R25 to allow time for WO implementation. This DOE is necessary since the work has been deferred by management. WDB 5/2 1/2013 Approved

Description:

ODE for this LTCA designated action approved by the Engineering Director per the attached email.

Attachments:

Approved Description Director approval ofDDE

Attachment Header Document Name:

untitled Document Location f pproved Description Attach

Title:

!Director approval of DOE

Page 1 of 1 BARBOREK, W DOUGLAS From: LEA, PHILLIP B Sent: Thursday, October 20, 2011 10:01 AM To: WOODSON, TIMOTHY R; BARBOREK, W DOUGLAS Cc: GUILTY, RICHARD C

Subject:

RE: PIFs pertinent to EC-032271 Pl F nos.

1203.012H 1- 11 -0621 1015.036 1-1 1-0622 1102.015 1- 11-0623 1103.002 1- 11 -0624 Phill ip Lea UNIT 1 OPERATIONS SPECIALIST 479-858-5498 From : WOODSON, Tl MOTHY R Se111t: Thursday, October 20, 2011 9:22 AM To: LEA, PHILLIP B

Subject:

FW: PIFs pertinent to EC-032271 limothy R Wood9Jn, P.E Arkansas Nudear Ole Phone: (479) 858-5544 Pax: (479 858-5529 Pager: (b)(6)

Alpha e From : Kane, O,ris Se111t: Thursday, October 20, 2011 1:52 AM To: GUILTY, RICHARD C Cc: BARBOREK, W DOUGLAS; WOODSON, TIMOTHY R

Subject:

Pl Fs pertinent to EC-032271 Mr. Guilty, Attached are the PIFs associated with the 1 R23 conversion of Valves RBS-1 & RBS-2 from Normally Closed to Normally Open. Reference EC-032271 Please provide PIF numbers for Doug. He w ill develop a CA.

Thank you, Chris Kane 10/20/2011

Attachment Header Document Name:

untitled Document Location FA Description Attach

Title:

IPIF for 1103.002

ARKANSAS NUCLEAR ONE Paae 1 E-DOC TITLE:

PROCEDURE IMPROVEMENT FORM I

E-DOC NO.

1000.006-P I CHANGE NO.

053 This Document Contains 2 Page(s)

Page 1 of 2 TO: Karl Jones UNIT:_~1_ _ __ DATE: 10/11/11 Responsible Supervisor/Superintendent FROM: Chris Kane PHONE:_3_97_5_ __

0 R

I PROCEDURENUMBER:__ 11_0_

3_

.0_02~---- CHANGE N0:_0-3-6_ _

G I PROCEDURE TITLE: FILLING AND VENTING THE REACTOR COOLANT SYSTEM N

A Description of Improvement: Identify step number, page numbers, etc., where applicable or attach copy of T affected pages with suggested changes marked. Include any research that has been performed, such as drawing numbers, personnel contacted, etc.

0 R

Attachment C p.7 of 10 Note 1 corresponding to Valves RBS-1 and RBS-2.

These valves will NOT be closed by Containment Building Closeout (1015.036), Attachment L "Pre-Critical Inspection".

These valves will remain OPEN.

Reference:

EC-032271 Supervisor review should include the following:

1. Is an immediate Revision needed instead of a PIF? 0'9=S ~o
2. Is PIF needed? ~ES ONO

~

3. Does change require a software or database change? DYES
4. Should a condition report be initiated? DYES

ARKANSAS NUCLEAR ONE Paae 2 E-DOC TITLE: E-DOC NO. CHANGE NO.

PROCEDURE IMPROVEMENT FORM 1000.006-P 053 Page 2 of 2 R

E Assignee:_ _ _ _ _ _ _ __ _ _ _ _ __

s p

0 D Evaluate & revise immediately D Evaluate and discuss N

D Evaluate & include in next change s D Revise immediately I D Revise by_ _ _ _ _ _ __

B D Include in Next Procedure Change L

E Procedure No:

Remarks s

u p

V C U TRACKING SYSTEMENTRY PIF#_ _ _

  • _ _ _ _ _
  • _ _ _ __ Date:_ _ _ __

L O COPY DEPT. HEAD K

RESPONSE: Incorporated in Change No. _ _ __

A s

s I

G N

E E

Response By: Date:

s u

p V

REVIEWED BY: Date:

C D COPY ORIGINATOR L

K D TRACKING SYSTEM CLOSEOUT Date:

PROC.JWORK PLAN NO. PROCEDURE/WORK PLAN TITLE:

PAGE: 65 of 72 1103.002 FILLING AND VENTING THE REACTOR COOLANT SYSTEM CHANGE: 036 ATTACHMENT C Page 7 of 10 VALVE TAG NUMBER ( ,/) OPEN CLOSED DESCRI PTI ON RB BASEMENT - B STEAM GENERATOR AREA Capped RBD- lOA X PP- 1047 Isolation RBD- 9A X P32A & B cold leg drain RBD-8A X P32A cold leg drain RBD-8B X P32B cold leg drain Capped RBD-9B X P32A & B cold leg drain flush connection N 1 RBS-1 X

- RV gasket leak detection

~ 1 I RBS- 2 X RV gasket leak detection SOUTH CAVITY - RCP SEAL ELEVATION Hot Leg Level LT-1193 Lower Tap RC-1073 X Isolation LT-1195 Upper Tap Isolation Hot Leg Level LT- 1193 Lower Tap RC- 1074 X Isolation LT-1195 Upper Tap Isolation SOUTH CAVITY - RCP MOTOR ELEVATION Hot Leg Level LT- 1191 Lower Tap RC-1075 X Isolation LT-1193 Upper Tap Isolation Hot Leg Level LT- 1191 Lower Tap RC-1076 X Isolation LT- 1193 Upper Tap Isolation Root Valve to PDT-1034, 1035, RC- 1035A X 1036, 1037 and PDX-1 035 Root Valve to PDT-1034 , 1035, 1036, 1037 and RC- 1035B X PDX- 1035 Root Valve to PDT- 1034, 1035, RC-1036A X 1036, 1037 and PDX-1035 Root valve to PDT-1034, 1035, RC-1036B X 1036, 1037 and PDX-1035 Note 1: RBS-1 and RBS-2 will be closed by Containment Building Closeout (1015.036),

(_ Attachment L " Pre-Critical I.nspe ction " .

(Z.~N, f,/Of'fi I

Attachment Header Document Name:

untitled Document Location FA Description Attach

Title:

IPIF for 1102.0 15

ARKANSAS NUCLEAR ONE Paae 1 E-DOC TITLE: E-DOC NO. CHANGE NO.

PROCEDURE IMPROVEMENT FORM I 1000.006-P I 053 This Document Contains 2 Page(s}

Page 1 of 2 TO: Karl Jones UNIT:_ __.:.1_ _ __ DATE: 10/11/11 Responsible Supervisor/Superintendent FROM: Chris Kane PHONE:__,3......

9__75.___ __

0 R

I PROCEDURE NUMBER_~11~0=2= ~~15~---- CHANGE NO: 032 G

I PROCEDURE TITLE: FILLING AND DRAINING THE FUEL TRANSFER CANAL N

A Description of Improvement: Identify step number, page numbers, etc., where applicable or attach copy of T affected pages with suggested changes marked. Include any research that has been performed, such as drawing numbers, personnel contacted, etc.

0 R

Attachment H p.1 Section 3.4 When line is drained, Leave valves RBS-1 and RBS-2 OPEN.

Reference:

EC-032271 Supervisor review should include the following:

1. Is an immediate Revision needed instead of a PIF? DYES ONO
2. Is PIF needed? DYES ONO
3. Does change require a software or database change? DYES ONO
4. Should a condition report be initiated? DYES ONO

ARKANSAS NUCLEAR ONE Pace 2 E-DOC TITLE: E-DOC NO. CHANGE NO.

PROCEDURE IMPROVEMENT FORM 1000.006-P 053 Page2 of2 R

E Assignee:

s p

Evaluate & revise immediately D Evaluate and discuss 0

s N 8D Evaluate & include in next change Revise immediately I D Revise by B D Include in Next Procedure Change L

E Procedure No:

Remarks s

u p

V C LJ TRACKING SYSTEM ENTRY PIF# - - Date:

L K

D COPY DEPT. HEAD RESPONSE: Incorporated in Change No.

A s

s I

G N

E E

Response By: Date:

s u

p V

REVIEWED BY: Date:

C D COPY ORIGINATOR L

K D TRACKING SYSTEM CLOSEOUT Date:

PROC./WORK PLAN NO. PROCEDURE/WORK PLAN TITLE:

PAGE: 170 of 176 1102.015 FILLING AND DRAJNING THE FUEL TRANSFER CANAL CHANGE: 032 ATTACHMENT H Page 1 of 2 DRAINING REACTOR VESSEL GASKET LEAKOFF LINES This attachment is performed after the refueling canal has been drained below the vessel flange elevation in preparation for replacing the vessel head. This attachment must be completed prior to the head being set. Containment Building Closeout (1015.036) will ensure that RV Gasket Leak Detection (RBS-1 and RBS-2) and RV Gasket Pressure Test Connection (RC-5) are closed prior to power operation.

Reference CR-AN0- 1-2005-1140.

1.0 Verify RCS level ~376.3' elevation.

2.0 Maintenance has been contacted for removing flange at RV Gasket Pressure Test Connection (RC-5) per MWO 00094316.

NOTE If desired, sub-section 4.0 may be p'e"rio'rmed prior to sub-section 3.0.

3.0 Drain RV gasket leak detection line as follows:

CAUTION RV Gasket Leak Detection (RBS-1 and RBS-2) drain to the RB Sump. Avoid spread of contamination.

3.1 Verify one of the following conditions exists (~):

() No personnel and no tools are in the RB Sump, AND sump is available to receive RCS drainage.

() Verify appropriate catch basin provided at drain line .

3.2 Open RV gasket leak detection (RBS-1). Location: RB basement, south end, on primary shield wall, under A RCP discharge, 12 ft up.

3.3 Slowly throttle open RV gasket leak detection (RBS-2) 0 and drain line. Location: RB basement, south end, on primary shield wall, under C RCP discharge, 11 ft up.

3.4 WHEN line is drained, THEN close the following valves:

  • RV gasket leak detection (RBS-1)
  • RV gasket leak detection (RBS-2)

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE:

PAGE: 171 of 176 1102.015 FILLING AND DRAINING THE FUEL TRANSFER CANAL CHANGE: 032 ATTACHMENT H Page 2 of 2 4.0 Drain RV gasket pressure test connection line as follows:

4. 1 Verify Maintenance has removed flange at RV gasket pressure test connection (RC-5). Location : RB basement , above incore tunnel, by primary shield wall.

CAUTION RV Gasket Pressure Test Connection (RC-5) drainage must be captured or controlled to prevent spread of contamination.

4.2 Coordinate with RP as needed AND verify appropriate catch basin or hose rigged below RV gasket pressure test connection (RC-5).

4.3 Slowly throttle open RV gasket pressure test connection (RC-5).

4.4 WHEN line is drained, THEN close RV gasket pressure test connection (RC-5).

5.0 WHEN attachment complete, THEN make a station log entry s tating such.

Date Date

Attachment Header Document Name:

untitled Document Location FA Description Attach

Title:

IPIF for 1015.036

ARKANSAS NUCLEAR ONE Page 1 E-DOC NO. CHANGE NO.

E-DOC TITLE:

PROCEDURE IMPROVEMENT FORM I 1000.006-P I 053 This Document Contains 2 Page(s)

Page 1 of 2 TO: Karl Jones UNIT:._ __,1_ _ __ DATE: 10/11/11 Responsible Supervisor/Superintendent FROM: Chris Kane PHONE:~3~9~75::......._ __

0 R

I PROCEDURE NUMBER:_....:.; 10"""'1:...:5.:..:

.0=36=------ CHANGE NO: 034 G

I PROCEDURE TITLE: CONTAINMENT BUILDING CLOSEOUT N

A Description of Improvement: Identify step number, page numbers, etc., where applicable or attach copy of T affected pages with suggested changes marked. Include any research that has been performed, such as drawing numbers, personnel contacted, etc.

0 R

Attachment L "Pre-Critical Inspection" p.3 of 7 Step 2.11.1 Valves RBS-1 and RBS-2, RV Gasket Leak Detection, will be OPEN.

Step 2.11 "NOTE" section, first bulleted sentence Remove "RBS-1, RBS-2," from this sentence.

Reference:

EC-032271 Supervisor review should include the following:

1. Is an immediate Revision needed instead of a PIF? DYES ONO
2. Is PIF needed? DYES ONO
3. Does change require a software or database change? DYES ONO
4. Should a condition report be initiated? DYES ONO

ARKANSAS NUCLEAR ONE Page 2 E-DOC TITLE: E-DOC NO. CHANGE NO.

PROCEDURE IMPROVEMENT FORM 1000.006-P 053 Page 2 of 2 R

E Assignee:

s p

0 D Evaluate & revise immediately D Evaluate and discuss N D Evaluate & include in next change s D Revise immediately I D Revise by B D Include in Next Procedure Change L

E Procedure No:

Remarks s

u p

V C D TRACKING SYSTEM ENTRY PIF# - - Date:

L K

D COPY DEPT. HEAD RESPONSE: Incorporated in Change No.

A s

s I

G N

E E

Response By: Date:

s u

p V

REVIEWED BY: Date:

C D COPY ORIGINATOR L

K D TRACKING SYSTEM CLOSEOUT Date:

PROC.IWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 65 of 70 1015.036 CONTAINMENT BUILDING CLOSEOUT CHANGE: 034 ATTACHMENT L Page 3 of 7 2.11 IF Unit 1, THEN perform the following:

NOTE

  • If time permits, the closing of RBS-1, RBS- 2, RBS-11B, RBS - 12B, RBS- llC, RBS-12C, RBS - 11D and RBS-12D should be performed~ 4 hrs after reaching RCS normal operating temperature (NOT). This will allow sufficient time for thermal equilibrium to be reached across the RCP and reactor vessel head flange gaskets.
  • Once RCS NOT is reached, System Engineering may request early closure of particular RCP intergasket leakage isolation valves in order to check the integrity of other RCP's inner gasket.
  • RBS- llA and RBS-12A are normally closed.
  • Steps in this section may be performed in any order.

2 .11 . 1 Veri fy the following val ves closed:

  • RBS-1 , RV Gas ket Lea k Detection
  • RBS- 2, RV Gasket Leak Detec t ion
  • RBS - llA, RCP P32A Intergasket Leak !sol
  • RBS-12A, RCP P32A Intergasket Leak !sol
  • RBS-11B, RCP P32B Intergasket Leak !sol
  • RBS-12B, RCP P32B Intergasket Leak !sol
  • RBS-llC, RCP P32C Intergasket Leak !sol
  • RBS-12C, RCP P32C Intergasket Leak Isol
  • RBS- 11D, RCP P32D Intergasket Leak !sol
  • RBS-12D, RCP P32D Intergasket Lea k !sol 2.11.2 Verify the line downstream of RV Pressure Test Connection (RC- 5) has blind flange installed .

2.11.3 Secure Containment Building lighting per Attachment 2 of 1107.005.

2.11.4 Open breaker 5642A to RB Elevator (M- 6).

2 . 11.5 Open breaker B713, Power to RB Crane.

Attachment Header Document Name:

untitled Document Location FA Description Attach

Title:

IPIF for 1203.012H

ARKANSAS NUCLEAR ONE Paae 1 E-DOC TITLE: E-OOC NO. CHANGE NO.

PROCEDURE IMPROVEMENT FORM I 1000.006-P I 053 This Document Contains 2 Page(s)

Page 1 of 2 TO: Karl Jones UNIT:_ _ 1 _ __ DATE: 10/11/11 Responsible Supervisor/Superintendent FROM: Chris Kane PHONE:.__,3=9-'-75=-----

0 R

I PROCEDURE NUMBER: 1203.012H CHANGE NO: 040 G

I PROCEDURE TITLE: ANNUNCIATOR K09 CORRECTIVE ACTION N

A Description of Improvement: Identify step number, page numbers, etc., where applicable or attach copy of T affected pages with suggested changes mari<ed. Include any research that has been performed, such as drawing numbers, personnel contacted, etc.

0 R

p. 9 of 64 Vessel Head Gasket Leak Note Section First Bullet RV Gasket Leak Det (RBS-1 and RBS-2) will Not be Closed during performance of Containment Building Closeout ( 1015.036).

These valves will remain OPEN.

Reference:

EC-032271 Supervisor review should include the following:

1. Is an immediate Revision needed instead of a PIF? ~ES
2. Is PIF needed? (ZfYES
3. Does change require a software or database change? DYES
4. Should a condition report be initiated? DYES

ARKANSAS NUCLEAR ONE Page 2 E-DOC TITLE: E-DOC NO. CHANGE NO.

PROCEDURE IMPROVEMENT FORM 1000.006-P 053 Page 2 of 2 R

E Assignee:_ _ _ _ _ _ _ _ _ _ _ _ _ __

s p

0 D Evaluate & revise immediately D Evaluate and discuss N

D Evaluate & include in next change s D Revise immediately I

D Revise by_ _ _ _ _ _ __

B D Include in Next Procedure Change L

E Procedure No:

Remarks s

u p

V C 0 TRACKING SYSTEM ENTRY PIF #_ _ _ - - - -- - - - - - - Date:_ _ _ __

L 0 COPY DEPT. HEAD K

RESPONSE: Incorporated in Change No. _ _ __

A s

s I

G N

E E

Response By: Date:

s u

p V

REVIEWED BY: Date:

C D COPY ORIGINATOR L

K D TRACKING SYSTEM CLOSEOUT Date:

PROC./WORK PLAN NO. PROCEDURE/WORK PLAN TITLE:

PAGE: 9of64 1203.012H ANNUNCIATOR K09 CORRECTIVE ACTION CHANGE: 040 Location: C14 Device and Setpoint: VESSEL HEAD Reactor Gasket Drain Temp (TS- 1052) >150°F GASKET LEAK Alarm: K09-Fl NOTE Ga s ket Lea Det (RBS- 1 and RBS- 2) are closed during performa nce of Conta~nment Building Closeout (1015 . 036) .

  • RCS Leakage Safety Evaluation instructions are contained in RCS Leak Detection (1103.013) .

1 .0 OPERATOR ACTIONS

1. Calculate RCS leakage using RCS Leak Detection (1103.013).
2. IF leakage is <10 gpm, THEN contact System Engineering.
3. IF l eakage is >10 gpm, THEN perform the following:
  • Refer to Excess RCS Leakage (1203 . 039) .
  • Refer to TS 3 . 4 . 13 and enter Condition B, if mode is applicable .

2.0 PROBABLE CAUSES

  • React or head gasket leak-off line temperature high

3.0 REFERENCES

Window Arrangement Annunciator K09 (E- 459, sheets 1- 4)

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00021 Version: Approved: r,/

Requested Duedate: 11 / 11/2011 Previous Duedate: 11/04/20 l 1 Requested By: Cuilty,Richard C 11 /03/20 11 Approved By: Cuilty,Richard C 11 /03/20 11 Request

Description:

Operations Management and Engineering to determine direction. Questions have been raised which may require RBS- I and RBS-2 to remain closed during the cycle.

It is acceptable and necessary to extend this action because a final decision has not been made.

Approved

Description:

Approved.

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00021 Version: 2 Approved: r,/

Requested Duedate: 11 / 18/20 11 Previous Duedate: 11/ 11/2011 Requested By: Cuilty,Richard C 11 /1 1/201 1 Approved By: Ho rton,Jeffrey S 11 /1 1/201 1 Request

Description:

Management has determined procedure changes are required. This actiorn will be performed. Currently, Unit I is in cold shutdown, therefore it is acceptable to extend this action.

It necessary to extend this action due to limited resources and late decision to perform action.

Approved

Description:

approve DOE

Entergy I CORRECTIVE ACTION I CR-AN0-1-2008-02560 CA N umber: 22 G roup Name Assig ned By: Eng Sys Mgmt ANO Wi lliams,Patrick J Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: ZzANO CRG **lHEA use only** 12/21/201 1 13:07:38 Performed By: Edgell,Douglas W 2/15/20 12 17:00: 11 Subperformed By: Barborck,W Douglas 2/ 15/2012 13:43:36 Approved By:

Closed By: Williams,Patrick J 2/15/2012 17:50:15 Current Due Date: 02/29/2012 Initial Due Date: 02/29/2012 CA Type: PERIODIC REVIEW CA Priority:

Plant Constraint: NONE CA

Description:

Interim and Periodic Review Required - Complete one of the fo llowing:

l) lf this CR and all CRs closed to this CR are NOT associated with Safety Related Equipment tben, clearly document in this actiol!l that this CR is not associated with Safety Related Equipment. Include a brief discussion of the basis for that determination.

OR

2) (NOTE - an Interim and Periodic Review requires both "Responsible Manager" AND a "Director or Above" approval). If this CR or any CR closed to this CR are associated with safety related equipment then conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist", Attachment 9.8 of EN-Ll-102 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section. Consider any open CAs for Long Term classification per Attachment 9.9 of EN-Ll-102.

Response

Tconcur with the fol lowing sub-response and closure of this corrective action.

Subresponse :

Approved Interim Review attached. E-mail with Director approval also attached. WDB 2/ 15/2012 Closure Comments:

Attachments:

Subresponse Description Interim Review Subresponse Description e-mail with Director approval

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!interim Review

ATTACHMENT 9.8 CR INTERIM AND PERIODIC REVIEW FORM SHEET 1 OF 1 CR Interim and Periodic Review CR Number: CR-AN0-1-2008-02560 Category Level D AD B IZI C CR Owner Group: ENG SYS MGMT ANO CR

Description:

1R2 1 W0-001 02463 Task 01 was not able to be successfully performed as planned.

CR-AN0-1-2005-0 1140 was written to document OE from several plants which identified corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/pressure test connection piping resulting from water left in the piping following refueling outages. AN0-1 was determined to be vulnerable since water has been trapped in this piping during previous operating cycles. The scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utilize a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV and the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned. Since CR-AN0-1-2005-0 1140 was closed to W0-001 02463, this CR was initiated to provide a means of tracking an alternate resolution for addressing the noted OE. As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained during refueling outages prior to installing the RV head, thus mitigating the damage mechanism. The operability statement in CR-AN0-1-2005-01140 remains applicable.

CR Review: (All No responses require explanation be included.)

The following CA's have been issued to further evaluate a course of action required to address t he OE documented in CR-AN0-1-2005-01 140. The issuance of these actions constitutes an acceptable corrective action plan for this CR. Additional CA's will be issued as deemed appropriate. Based on past discussions with plant management, the pressure test scope needs to be transferred from the responsibility of System Engineering to Project Management to ensure successful 1R24 implementation. Otherwise, resolution of this OE in 1 R24 is in jeopardy.

CA-02 [closed] - CR-AN0-1 -2005-01140 (CA-04) evaluated the feasibility of performing a hydrostatic test of the RV flange leak-off/pressure test connection piping and concluded that such a test was not feasible. Based on the inability to visually inspect the piping during 1R21 (via boroscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity. This CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization.

CA-03 [closed] - Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e. removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping. This CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization.

CA-07 [closed] - The inspection of the RV Flange Leak-off & Pressure Test Connection piping requires additiona l resources and engineering which transcends the role and responsibility of System Engineering.

Initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization. SIPDB Record 4955 was initiated to transfer this scope of work to the Projects Organization and this CA was closed on 11/19/2009. New CA-08 was issued to track the presentation of SIPDB Record 4955 to the URT.

CA-08 [closed] - Present SIPDB Record (Inspection of the RV Flange Leak-off & Pressure Test Connection Piping) to the URT/MPRC for this scope of work to be executed by the Projects Organization. SYE made the presentation at the 12/7/2009 URT meeting. The URT concurred t hat if the leak-off & pressure test connection EN-Ll-102 REV 13

piping ultimately require physical inspection or replacement such that disassembly/reassembly of the reactor cavity seal plate, concrete shield blocks, and reactor vessel insulation is required, then that significant scope of work would be transferred to the Project Organization. However, at this time, the URT decided that System Engineering should re-evaluate the pressure test option and come up with a plan to perform a pressure test on the leak-off and pressure test connection piping to verify the integrity of the piping. If the pressure test fails on one or both of the lines, the ensuing scope of work to access the piping for ultimate resolution of the issue wou ld the n be pursued by the Projects Organization. New CA-09 has been issued to SYE to determine a suitable pressure test method and to initiate the proper implementation documentation. 1R22 will be utilized to perform scoping walkdowns since this piping is not accessible during power operations. This CA was closed on 1/2 1/2010.

CA-09 [closed] - Coordinate with EP&C personnel and determine a suitable pressure test method for the RV flange leak-off/pressure test connection piping, and initiate the proper implementation documentation. Utilize 1R22 to perform scoping walkdowns as required. Initiate other corrective actions as required . Completed by SYE on 10/25/2010. CA-10 was initiated for SYE to assign followup corrective actions as required to implement the pressure test plan in 1R23 via W0-001 95437. CA-10 was completed on 11/17/2010 by SYE and CA-11 and CA-12 were initiated as a result.

CA-1 1 [closed] - Develop EC to support W0-001 95437. If required, provide justification for installation of couplings on the affected CCC-6 piping. Also, if required, provide engineering requirements for FME controls associated with the pressure test equipment and components to be used in close proximity to the RV flange.

CA closed on 7/21/2011 with no action taken since pressure test plan no longer requires cutting/capping piping (isolation valves RBS-1 & RC-5 will be used as the pressure boundary).

CA-12 [closed] - Determine the appropriate type & size of hydro plugs to be utilized for the pressure test (W0-00195437) outlined in CA-09. Initiate actions to procure the hydro plugs. as required. CA completed by SYE on 6/9/201 1. CA-1 8 issued to SYE to develop CATID's (if required).

CA-14 [closed] - Provide input to Outage P&S to ensure 1R23 W0-00195437 is properly planned for performance of the CCC-6 piping pressure test. CA closed by SYE on 5/5/2011 since the WO was deferred to 1R24.

CA-15 [closed] - Per EN-OU-100, ide ntify any required contingency plans for 1R23 which are required to support the pressure test of the CCC-6 piping . Issue follow-up CA's as required. CA closed by SYE on 5/18/2012 since the WO was deferred to 1R24 .

CA Following implementation of W0-00195437 during 1 R23, document that the applicable OE has been adequately addressed for AN0-1. CA assigned to SYE. Current due date is 5/30/2013 which follows completion of 1 R24.

CA-17 [closed] - Present SIPD Record 5678 to the URT to request reassignment of the RV Flange leak-off/pressure test connection piping pressure test (W0-00195437) from System Engineering to Project Management for 1R24 implementation.

CA Develop CATIDs for the tes.t plugs specified in CA-12. Issue additional corrective actions as required to ensure material is pledged to W0-00195437. Current due date is 2/23/2012. Anticipate this action being closed to no action taken since the hydro plugs are a tool and not a stock item.

CA Ensure W0-00195437 is implemented in 1 R24. This WO has been deferred to 1 R24 as documented in the attached 1R23 SCR 11049. Issue additional corrective actions as required to support 1 R24 resolution of this issue. This CA is being classified as a LTCA. Current due date is 5/29/2013, based on anticipated completion of 1R24.

CA-21 [closed] - Incorporate PIF 1-11-0621 to OP-1203.012H, PIF 1-11 -0622 to OP-1015.036, PIF 1-11 -0623 to OP-1102.015, and PIF 1-11-0624 to OP-1 103.002 to change the configuration of valves RBS-1 and RBS-2 from normally closed to normally open for Cycle 24 operation. This change has been approved by EC-32271 .

EN-Ll-102 REV 13

CA closed by Operations on 11/16/2011. RBS-1 and RBS-2 are currently open for Cycle 24 operation which eliminates the potential for pressurizing the leak-off piping in the event of inner o-ring leakage.

1. Will the existing corrective actions documented in the condition report, when completed, correct the condition report issue? Yes 1Z! / No D Implementation of 1R24 W0-00195437 will resolve this issue, assuming no degradation is discovered. Since additional CA's are expected between now and 1R24, the CR will remain open through 1R24 and will be closed following completion of CA-16 & CA-20.

2 . What is the expected CR Closure date based on remaining needed actions? DATE: 5/30/2013 It is anticipated that this CR will be closed following 1R24 implementation of the pressure test and completion of CA- 16 & CA-20. Target date is 5/30/2013. If degradation is found, a new CR will be initiated during 1R24.

3. Is the previously documented operability/functionality position still valid for the current condition and expected to remain valid until CR closure? Yes IZ] / No D /N/A D If the answer is NO, then initiate a new CR to document the concern; CR# N/A
4. Are all Ll-102 requirements for corrective action administration and control being met, i.e.

justifications for Due Date Extensions valid, Long Term Corrective Actions identified, CARB approved CAPRs identified, and appropriate approvals obtained for all?

Yes IZ]! No D

5. What risk to plant operation is imposed by the condition identified and how is risk reduced to an acceptable level for the duration of the action plan?

This CR does not specifically identify a known degraded plant system, structure, or component at AN0-1 , but does identify a potential degradation mechanism. Both the RV flange gasket leak detection line and pressure test connection lines were flushed/drained during 1R19, 1R20, 1R2 1, 1R22 and 1R23, thus min imizing the potential for failure resulting from ID initiated, chloride induced, tran sgranular stress-corrosion cracking (TGSCC ). No RV flange gasket leakage was identified during plant heat-up from 1R19, 1R20, 1 R21, or 1R22 prior to isolating valves RBS-1 & RBS-2 for plant operation. RBS-1 & RBS-2 were left open for Cycle 24 to prevent pressurization of the piping in the event of an inner o-ring leak. RCS leakage rates following 1R19, 1R20, 1R21 , 1R22 and 1R23 have not indicated the presence of any concurrent leakage of the RV flange gaskets and leak detection/pressure test connection piping/components.

The RV f lange gaskets are replaced during each refueling outage and valves RBS-1 and RBS-2 were left open for Cycle 24 operation. The likelihood of inner gasket leakage, while possible, is considered low.

In conclusion, the risk to plant operation is minimal and reduced to acceptable levels until the action plan for this CR can be implemented during 1R24.

Review/ Approval Required:

Director/GM

Title:

Date:

(Print name & Position title)

NOTE: The expectation is to capture the discussion points of this form in a CA. The form itself need not be used, but all points applicable must be addressed.

EN-Ll-102 REV 13

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

~ -mail with Director approval

BARBOREK, W DOUGLAS From: MCCOY, JAIME H Sent: Wednesday, February 15, 2012 1:04 PM To: BARBOREK, W DOUGLAS; WILLIAMS, PATRICK J Cc: EDGELL, DOUGLAS W

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Flange Piping" I approve the interim review.

Linda - can you set up a meeting with Bauman, Edgell, W illiams, Barborek, and myself?

Jaime From : BAffiOREK, W DOUGLAS Sent: Monday, February 13, 2012 5:20 PM To: MCCOY, JAIMEH; WILLIAMS, PATRICKJ Cc: EDGELL, DOUGLAS W

Subject:

REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Oiloride Induced Stress Corrosion Cracking on RV Range Aping" Jaime & Patrick, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-22 of the subject CR. T he CA is ultimately due on Wednesday, February 15, 2012 (initially due by me today).

As we discussed last cycle, the pressure testing of the RV leak-off/pressure test connection piping needs to be turned over to the Project Management group. System Engineering does not have the ability or luxury to focus on a critical path activity like this to ensure it is coordinated and executed flawlessly during a refueling outage.

An example which highlights this is the 1 R23 activity which I drove to obtain diametrical data and boroscope inspections of the leak-off ports to support development of contingency efforts. We all know that did not go well after it fell to the opposite shift, who were not fully engaged in the activity and ultimately only obtained half of the information I wanted to obtain. Part of that was my fault due to the fact that I could not even think about this inspection during the first 6 nights of the outage due to my other required inspections.

I would like to meet with you, and perhaps Project Management (David Bauman}, to discuss this issue and the ultimate transfer of responsibility for the pressure test to the PM organization. Otherwise, I feel another outage will pass and we will still not have tested these lines to resolve the now 10 year old OE.

Please let me know if you have any questions.

Thanks, Doug Barborek Entergy Operations, Inc./ Arkansas Nuclear One System Engineer - Unit I Reactor Coolant System System Engineering Building/ N-SYS-4 wbarbo l @entergy.com 479-858-4337 479-858-0879 pager

<< File: CR-AN0-1-2008-02560 CA-22 lnterim Review.doc >>

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00022 Version: Approved: r,/

Requested Duedate: 02/29/2012 Previous Duedate: 02/ 15/20 12 Requested By: Barborek,W Douglas 02/15/2012 Approved By: Barborek,W Douglas 02/15/2012 Request

Description:

Due to an emergent plant issue and potential for a forced outage, this CA needs to be extended. TR is complete but awaiting director/manager approval. Extending two weeks to 2/29/2012. This CA is administrative in nature and extending it to 2/29/20 J2 is therefore acceptable. WDB 2/ l 5/20 J 2 Approved

Description:

Discussed this 1st DOE with Doug Edgell, who concurs with the DOE. Doug Barborek (for Doug Edgell)

Entergy I CORRECTIVE ACTION I CR-AN0-1-2008-02560 CA N umber: 23 G roup Name Assig ned By: Eng Sys Mgmt ANO Edgell,Douglas W Asstgned To: Eng Sys Mgmt ANO Edgell,Douglas W Subassig ned To: Eng Sys NSSS Staff ANO Barborck,W Douglas Originated By: Zz ANO CRG **lHEA use only** 12/20/201 2 l6:04:3S Performed By: Bond,Yincent S 2/12/201 3 11 :57:56 Subperformed By: Barborck,W Douglas 2/ 11 /201 3 15:43:3 1 Approved By:

Closed By: Oliver,Jason R 2/12/2013 17:23:49 Current Due Date: 02/ 13/2013 Initial Due Date: 02/1 3/201 3 CA Type: PERIODIC REVIEW CA Priority: 4 Plant Constraint: NONE CA

Description:

Interim and Periodic Review Required - Complete one of the fo llowing:

l) lf this CR and all CRs closed to this CR are NOT associated with Safety Related Equipment tben, clearly document in this actiol!l that this CR is not associated with Safety Related Equipment. Include a brief discussion of the basis for that determination.

OR

2) (NOTE - an Interim and Periodic Review requires both "Responsible Manager" AND a "Director or Above" approval). If this CR or any CR closed to this CR are associated with safety related equipment then conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist", Attachment 9.8 of EN-Ll-102 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section. Consider any open CAs for Long Terrn classification per Attachment 9.9 of EN-Ll-102.

Response

Sub-response is acceptblc.

Subresponse :

Interim Review and e-mail documenting Engineering Director approval are attached. WDB 2/ 11/2013 Closure Comments:

Interim review has been completed and approved by the engineering director. This action is complete and may be closed.

Jason Oliver acting for Doug Edgell. 2 13 Attachments:

Subresponse Description Director approval e-mail (McCoy)

Subresponse Description Interim Review

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!Director approval e-mail (McCoy)

BARBOREK, W DOUGLAS From: MCCOY, JAIME H Sent: Monday, February 11, 2013 1 :09 PM To: BARBOREK, W DOUGLAS; EDGELL, DOUGLAS W Cc: WOODSON, TIMOTHY R

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Cracking on RV Flange Piping" I approve the interim review.

I think we had put this on the P02 major maintenance list for 1R25. Please send an email to John Jacobs/Dustin Shoptaw/Randall Walters with the O&M estimate for developing the contingency plan including Areva cost -

they need to get that in the 2014 budget (unless we need to do some of it this year). I thought that Randal was going to ask for money to cover this. in his outage budget, but he continues to get squeezed so the site may have to eat it in O&M projects. If it ever turns into an actual installation, it might go capital at that time.

Jaime From : BAffiOREK, W DOUGLAS Sent: Monday, February 11, 201312:57 PM To: MCCOY, JAI ME H; EDGELL, DOUGLAS W Cc: WOODOON, Tl MOTHY R

Subject:

REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Olloride Induced Stress Omosion Cracking on RV Range Piping" Jaime & Doug, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-23 of the subject CR. The CA is ultimately due on Wednesday, February 13, 2013.

See e-mail string below from last year's IR. Since Corporate Project Management refused in 2012 (see attached e-mail) to accept the scope of this pressure test and contingency development, the pressure test will not occur in 1 R24. We will need to regroup and form a new plan if we are to pull this off for 1 R25.

Thanks, Doug Barborek Entergy Operations, Inc. / Arkansas Nuclear One System Engineer* Unit 1 Reactor Coolant System System Engineering Building/ N-SYE-4 wbarbo I @cntcrgy.com 479-858-4337 j(b)(6) Ipager

<< Messnge: f'W: /\NOi RV Seal Leak Off Contingency>> << File: CR-AN0- 1-2008.02560 CA-23 Interim Review.doc>>

From : MCCOY, JAi ME H Sent: Wednesday, February 15, 20121 :04 PM To: BARBOREK, W DOUGLAS; WILLI AMS, PATRI a< J Cc: EDGELL, DOUGLAS W

Subject:

RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Olloride Induced Stress Corrosion Cracking on RV Range Piping" I approve the interim review.

Linda - can you set up a meeting with Bauman, Edgell, Williams, Barborek, and myself?

Jaime From : BAPBOREK, W DOUGLAS Sent: Monday, February 13, 2012 5:20 PM To: MCCOY, JAIMEH; WILLIAMS, PATRIO<J Cc: EDGELL, DOUGLAS W

Subject:

REQUEST FOR REVIEW - Interim Review for CR-AN0-1 -2008-02560 "Resolve Industry OE on Olloride I nduced Stress Omosion Oacking on RV Range Aping" Jaime & Patrick, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-22 of the subject CR. The CA is ultimately due on Wednesday, February 15, 2012 (initially due by me today).

As we discussed last cycle, the pressure testing of the RV leak-off/pressure test connection piping needs to be turned over to the Project Management group. System Engineering does not have the ability or luxury to focus on a critical path activity like this to ensure it is coordinated and executed flawlessly during a refueling outage.

An example which highlights this is the 1 R23 activity which I drove to obtain diametrical data and horoscope inspections of the leak-off ports to support development of contingency efforts. We all know that did not go well after it fell to the opposite shift, who were not fully engaged in the activity and ultimately only obtained half of the information I wanted to obtain. Part of that was my fault due to the fact that I could not even think about this inspection during the first 6 nights of the outage due to my other required inspections.

I would like to meet with you, and perhaps Project Management (David Bauman}, to discuss this issue and the ultimate transfer of responsibility for the pressure test to the PM organization. Otherwise, I feel another outage will pass and we will still not have tested these lines to resolve the now 10 year old OE.

Please let me know if you have any questions.

Thanks, Doug Barborek Entergy Operations. Inc./ Arkansas Nuclear One System Engineer - Unit l Reactor Coolant System System Engineering Building / N-SYS-4 wbarbo I @entergy.com 479-858-4337

!(b )(6)

<< File: CR-AN0 2008-02560 CA-22 lnterim Review.doc>>

2

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!interim Review

ATTACHMENT 9.8 CR INTERIM AND PERIODIC REVIEW FORM SHEET 1 OF 1 CR Interim and Periodic Review CR Number: CR-AN0-1-2008-02560 Category Level D AD B IZI C CR Owner Group: ENG SYS MGMT ANO CR

Description:

1R21 W0-001 02463 Task 01 was not able to be successfully performed as planned.

CR-AN0-1-2005-01140 was written to document OE from several plants which identified corrosion (chloride) induced pipe cracking in reactor vessel (RV) flange leak-off/pressure test connection piping resulting from water left in the piping following refueling outages. AN0-1 was determined to be vulnerable since water has been trapped in this piping during previous operating cycles. The scope of W0-00102463 was for System Engineering to attempt to externally inspect the RV flange leak-off/pressure test connection piping for evidence of cracking. The inspection approach was to utilize a boroscope to externally inspect the piping via access through the reactor cavity seal plate openings and between the RV and the RV insulation. However, due to the tight clearances between the insulation and the RV, the inspection was not successful and the leak-off/pressure connection piping could not be visually inspected as planned. Since CR-AN0-1-2005-01140 was closed to W0-001 02463, this CR was initiated to provide a means of tracking an alternate resolution for addressing the noted OE. As documented in the 2005 CR, ANO procedures have been revised to ensure the subject piping is drained during refueling outages prior to installing the RV head, thus mitigating the damage mechanism. The operability statement in CR-AN0-1-2005-01140 remains applicable.

CR Review: (All No responses require explanation be included.)

The following CA's have been issued to further evaluate a course of action required to address the OE documented in CR-AN0-1-2005-01 140. The issuance of these act ions constitutes an acceptable corrective action plan for this CR. Additional CA's will be issued as deemed appropriate. Per past discussions with plant management, the pressure test scope was requested in 2012 to be transferred from the responsibility of System Engineering to Project Management to ensure successful 1R24 implementation. This was due to the coordination required to develop and implement the pressure test and due to the extensive contingency planning required, which includes AREVA support. This request was rejected by Corporate Project Management. As such, resolution of this OE in 1 R24 will not occur due to lack of adequate resources. Resolution by 1R25 is dependant solely on allocation of site resources, which at this time is unresolved.

CA-02 [closed] - CR-AN0-1 -2005-01140 (CA-04) evaluated the feasibility of performing a hydrostatic test of the RV flange leak-off/pressure test connection piping and concluded that such a test was not feasible. Based on the inability to visually inspect the piping during 1R21 (via boroscope), revisit the feasibility of performing a hydrostatic test on these lines to verify piping integrity. Th is CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPOS Record and present to the URT/MPRC to recommend this scope of work be ,executed by the Projects Organization.

CA-03 [closed] - Determine the scope required to perform a direct visual inspection of the RV flange leak-off/pressure test connection piping (i.e . removal of reactor cavity seal plate, concrete shield blocks, and RV insulation) in the event that no other option is identified to verify the integrity of the piping. This CA was closed on 8/27/2009 to CA-07 which was initiated for SYE to initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization.

CA-07 [closed] - The inspection of the RV Flange Leak-off & Pressure Test Connection piping requires additiona l resources and engineering which transcends the role and responsibility of System Engineering.

Initiate a new SIPDB Record and present to the URT/MPRC to recommend this scope of work be executed by the Projects Organization. SIPDB Record 4955 was initiated to transfer this scope of work to the Projects Organization and this CA was closed on 11/19/2009. New CA-08 was issued to track the presentation of SIPDB Record 4955 to the URT.

EN-Ll-102 REV 13

CA-08 [closed] - Present SIPDB Record (Inspection of the RV Flange Leak-off & Pressure Test Connection Piping) to the URT/MPRC for this scope of work to be executed by the Projects Organization. SYE made the presentation at the 12/7/2009 URT meeting. The URT concurred that if the leak-off & pressure test connection piping ultimately require physical inspection or replacement such that disassembly/reassembly of the reactor cavity seal plate, concrete shield blocks, and reactor vessel insulation is required, then that significant scope of work would be transferred to the Project Organization. However, at this time, the URT decided that System Engineering should re-evaluate the pressure test option and come up with a plan to perform a pressure test on the leak-off and pressure test connection piping to verify the integrity of the piping . If the pressure test fails on one or both of the lines, the ensuing scope of work to access the piping for ultimate resolution of the issue would then be pursued by the Projects Organization. New CA-09 has been issued to SYE to determine a suitable pressure test method and to initiate the proper implementation documentation. 1R22 will be utilized to perform scoping walkdowns since this piping is not accessible during power operations. This CA was closed on 1/21/2010.

CA-09 [closed] - Coordinate with EP&C personnel and determine a suitable pressure test method for the RV flange leak-off/pressure test connection piping, and initiate the proper implementation documentation. Utilize 1R22 to perform scoping walkdowns as required. Initiate other corrective actions as required. Completed by SYE on 10/25/2010. CA-10 was initiated for SYE to assign followup corrective actions as required to implement the pressure test plan in 1R23 via W0-00195437. CA-1 O was completed on 11/17/201 O by SYE and CA-1 1 and CA-1 2 were initiated as a result.

CA-11 [closed] - Develop EC to support W0-00195437. If required, provide justification for installation of couplings on the affected CCC-6 piping. Also, if required, provide engineering requirements for FME controls associated with the pressure test equipment and components to be used in close proximity to the RV flange.

CA closed on 7/21/2011 with no action taken since pressure test plan no longer requires cutting/capping piping (isolation valves RBS-1 & RC-5 will be used as the pressure boundary).

CA-12 [closed] - Determine the appropriate type & size of hydro plugs to be utilized for the pressure test (W0-00195437) outlined in CA-09. Initiate actions to procure the hydro plugs, as required. CA completed by SYE on 6/9/201 1. CA-18 issued to SYE to develop CATID's (if required).

CA-14 [closed] - Provide input to Outage P&S to ensure 1R23 W0-00195437 is properly planned for performance of the CCC-6 piping pressure test. CA closed by SYE on 5/5/201 1 since the WO was deferred to 1R24.

CA-15 [closed] - Per EN-OU-100, identify any required contingency plans for 1R23 which are required to support the pressure test of the CCC-6 piping. Issue follow-up CA's as required. CA closed by SYE on 5/18/2012 since the WO was deferred to 1R24.

CA Following implementation of W0-00195437 during 1 R23, document that the applicable OE has been adequately addressed for AN0-1 . CA assigned to SYE. Current due date is 5/30/2013 which follows completion of 1 R24. Will have to be extended again since this work is not going to be performed in 1R24 due to lack of resources and contingency planning.

CA-17 [closed] - Present SIPD Record 5678 to the URT to request reassignment of the RV Flange leak-off/pressure test connection piping pressure test (W0-00195437) from System Engineering to Project Management for 1R24 implementation.

CA-18 [closed] - Develop CATIDs for t he test plugs specified in CA-12. Issue additional corrective actions as required to ensure material is pledged to W0-00195437. CA closed on 2/22/2012 since the hy,dro plugs are a tool and not a stock item .

CA Ensure W0-00195437 is implemented in 1 R24. This WO has been deferred to 1 R24 as documented in the attached 1R23 SCR 11049. Issue additional corrective actions as required to support 1 R24 resolution of this issue. This CA is being classified as a LTCA. Current due date is EN-Ll-102 REV 13

5/29/2013, based on anticipated completion of 1R24. This CA will have to be extended again to 1R25 since this work will not be performed in 1R24 due to lack of resources and contingency planning.

CA-21 [closed] - Incorporate PIF 1-11-0621 to OP-1 203.012H, PIF 1-11-0622 to OP-1 015.036, PIF 1-11 -0623 to OP-1102.015, and Pl F 1-1 1-0624 to OP-1103.002 to change the configuration of valves RBS-1 and RBS-2 from normally closed to normally open for Cycle 24 operation. T his change has been approved by EC-32271 .

CA closed by Operations on 11/16/201 1. RBS-1 and RBS-2 are currently open for Cycle 24 operation which eliminates the potential for pressurizing the leak-off piping in the event of inner o-ring leakage.

1. Will the existing corrective actions documented in the condition report, when completed, correct the condition report issue? Yes ~ / No D Implementation of W0-00195437 will resolve this issue, assuming no degradation is discovered . Since additional CA's are expected between now and 1R25, the CR will remain open through 1R25 and will be closed following completion of CA-16 & CA-20.
2. What is the expected CR Closure date based on remaining needed actions? DATE: 11/30/2014 It is anticipated that this CR will be closed following 1R25 implementation of the pressure test and completion of CA- 16 & CA-20. Target date is 11/30/2014. If degradation is found, a new CR will be initiated during 1R25.
3. Is the previously documented operability/functionality position still valid for the current condition and expected to remain valid until CR closure? Yes ~ / No D /N/A D If the answer is NO, then initiate a new CR to document the concern; CR# NIA
4. Are all Ll-102 requirements for corrective action administration and control being met, i.e.

justifications for Due Date Extensions valid, Long Term Corrective Actions identified, CARB approved CAPRs identified, and appropriate approvals obtained for all?

Yes~/ No D

5. What risk to plant operation is imposed by the condition identified and how is risk reduced to an acceptable level for the duration of the action plan?

This CR does not specifically identify a known degraded plant system, structure, or component at AN0-1 , but does identify a potential degradation mechanism. Both the RV flange gasket leak detection line and pressure test connection lines were flushed/drained during 1R19, 1R20, 1R2 1, 1R22 and 1R23, thus min imizing the potential for failure resulting from ID initiated, chloride induced, tran sgranular stress-corrosion cracking (TGSCC). No RV flange gasket leakage was identified during plant heat-up from 1R1 9, 1R20, 1 R21 , or 1R22 prior to isolating valves RBS-1 & RBS-2 for plant operation. RBS-1 & RBS-2 were left open for Cycle 24 to prevent pressurization of the piping in the event of an inner a-ring leak. It is anticipated that these valves will be open for Cycle 25 operation as well. RCS leakage rates following 1R19, 1R20, 1R2 1, 1R22 and 1R23 have not indicated the presence of any concurrent leakage of the RV flange gaskets and leak detection/pressure test connection piping/components.

The RV flange gaskets are replaced during each refueling outage and valves RBS-1 and RBS-2 were left open for Cycle 24 operation. The likelihood of inner gasket leakage, while possible, is considered low.

In conclusion, the risk to plant operation is minimal and reduced to acceptable levels until the action plan for this CR can be implemented during 1R25.

Review/ Approval Required :

Director/GM Title : Date:

(Print name & Position title)

EN-Ll-102 REV 13

NOTE: The expectation is to capture the discussion points of this form in a CA. The form itself need not be used , but all points applicable must be addressed .

EN-Ll-1 02 REV 13

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 24 G roup Name Assigned By: Eng Systems & Comps Mgmt ANO Edgell,Douglas W Asstgned To: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Subassig ned To : Eng Systems NSSS Staff ANO Barborck,W Douglas Originated By: Zz ANO CRG **lHEA use only** l/27/2014 09:55:21 Performed By: Woodson P.E.,Timothy R 2/1 0/2014 21:0 I :24 Subperformed By: Barborck,W Douglas 2/ 10/2014 15:02:01 Approved By:

Closed By: Edgell,Douglas W 2/12/2014 07:10:59 Current Due Date: 02/ 12/2014 Initial Due Date: 02/1 2/2014 CA Type: PERIODIC REVIEW CA Priority: 4 Plant Constraint: NONE CA

Description:

Interim and Periodic Review Required - Complete one of the fo llowing:

I) If this CR and all CRs closed to this CR are NOT associated with Safety Related Equipment tben, clearly document in this actiol!l that this CR is not associated with Safety Related Equipment. Include a brief discussion of the basis for that determination.

OR

2) (NOTE - an Interim and Periodic Review requires both "Responsible Manager" AND a "Director or Above" approval). If this CR or any CR closed to this CR are associated with safety related equipment then conduct and document an interim review of this Condition Report using the "CR Interim and Periodic Review Checklist", Attachment 9.8 of EN-Ll-102 which is available via the Reference Library ECH Site in the Nuclear Management Manual Common Forms section. Consider any open CAs for Long Terrn classification per Attachment 9.9 of EN-Ll-102.

Response

Concur with sub-response Subresponse :

Interim Review is attached. E-mail documenting Director approval is also attached. WDB 2/10/2014 Closure Comments:

I concur with closure of this corrective action.

Attachments:

Subresponse Description Interim Review Subresponse Description e-mail documenting Director approval

Attachment Header Document Name:

untitled Document Location

~ ubresponse Description Attach

Title:

!interim Review

BARBORB<, W DOUGLAS From: MCCDY, JA.IM EH Sent: S.mday, February 09, 2014 10:16 PM To : BAROOREK, W DOUGLAS; EDGaL, DOUGLAS W Cc: WOODSJN , TIMOTHY R SUbject: RE: REQUEST FOR REVIEW - Interim Review for CR-AN0- 1-2008-02560 "Resolve Industry OE on O,loride Induced $ress C.orrosion Qacking on RV Range Piping" I approve the interim review.

Is this in a.ment 1R25 outage g;ope?

If it is not, then we need to have aO&M project added to 2015 budget to oontract Areva to develop theoontingency plan/rolled plug, oo we can do in 1R26. If it is, then it appears we are behind the funding 8ball because there is not an O&M project that I know of in 2014 budget for this item...

..aime From : BARBOREK, W DOUGLAS Sent: Thursday, February 06, 2014 4:39 Pfv1 To: MCCOY, JAJ ME H; EDGELL, DOUG.AS W Cc: WOODSJN, Tl MOTHY R Subject : REQUEST FOR REVIEW - I nterim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Chloride Induced Stress Corrosion Qacking on RV Flange Piping" Jaime & Doug, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-24 of the subject CR. The CA is ultimately due on Wednesday, February 12, 2014.

See e-mail string below from 2012's IR. Since Corporate Project Management refused in 2012 (see attached e-mail) to accept the scope of this pressure test and contingency development, the pressure test did not occur in 1 R24. We will need to regroup and form a new plan if we are to pull this off for 1R25. This effort is once again in jeopardy for 1R25 due to lack of available resources.

Thanks, Doug Barborek Entergy Operations, Inc. I Arkansas Nuclear One System Engineer - Unit 1 Reactor Coolant System System Engineering Building / N-SYE-4 wbarbo I @entergy.com 479-858-4337

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<< Message: FW : ANO ! RV Seal Leak Off Contingency>> << File: CR-ANO- l -2008-02560 CA-24 lnterim Review.doc >>

From : MCCOY, JAi ME H Sent: Wednesday, February 15, 2012 1:04 PM To: BARBOREK, W DOUG.AS; WILLIAMS, PATRIO<J Cc: EDGELL, DOUGLAS W

Subject : RE: REQUEST FOR REVIEW - Interim Review for CR-AN0-1-2008-02560 "Resolve Industry OE on Olloride Induced Stress Corrosion Oacking on RV Range Piping" I approve the interim review.

Linda - can you set up a meeting with Bauman, Edgell, Williams, Barborek, and myself?

Jaime From : BARBOREK, W DOUGLAS Sent: Monday, February 13, 2012 5:20 Pfv1 To: MCCOY, JAi ME H; WILLI AMS, PATRI O< J Cc: EDGELL, DOUGLAS W

Subject:

REQUEST FOR REVIEW - Interim Review for CR-AN0- 1-2008-02560 "Resolve Industry OE on Olloride Induced Stress Corrosion Oacking on RV Range Piping" Jaime & Patrick, The interim review for the subject CR is attached for your review and concurrence. The draft interim review has been attached to CA-22 of the subject CR. The CA is ultimately due on Wednesday, February 15, 2012 (initially due by me today).

As we discussed last cycle, the pressure testing of the RV leak-off/pressure test connection piping needs to be turned over to the Project Management group. System Engineering does not have the ability or luxury to focus on a critical path activity like this to ensure it is coordinated and executed flawlessly during a refueling outage.

An example which highlights this is the 1R23 activity which I drove to obtain diametrical data and horoscope inspections of the leak-off ports to support development of contingency efforts. We all know that did not go well after it fell to the opposite shift, who were not fully engaged in the activity and ultimately only obtained half of the information I wanted to obtain. Part of that was my fault due to the fact that I could not even think about this inspection during the first 6 nights of the outage due to my other required inspections.

I would like to meet with you, and per haps Project Management (David Bauman}, to discuss this issue and the ultimate transfer of responsibility for the pressure test to the PM organization. Otherwise, I feel another outage will pass and we will still not have tested these lines to resolve the now 10 year old OE.

Please let me know if you have any questions.

Thanks, Doug Barborek Entergy Operations, Inc./ Arkansas Nuclear One System Engineer - Unit I Reactor Coolant System System Engineering Building/ N-SYS-4 wbarbo I @cntcrgy.com 479-858-4337

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<< File: CR-AN0-1-2008-02560 CA-22 Interim Review.doc>>

2

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 25 G roup Name Assig ned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Woodson P.E.,Timothy R l/21/2015 16:23:34 Performed By:

Subperformed By:

Approved By:

Closed By:

Current Due Date: 09/03/2015 Initial Due Date: 09/04/2015 CA Type: CAT C-CORRECT CA Priority: 3 Plant Constraint: NONE CA

Description:

Based on the l/15/2015 ECRG meeting, SIPD-5678 "ANO! - Resolve OE on RV Flange Leak-off/Press Test Connection Pipe Cracking" was not approved for I R26 but was deferred to I R27 by the ECRG due to a lack of available Engineering resources. To support the 1R27 efforts to resolve this 2003 OE, obtain a cost estimate from AREY A to design and qualify a rolled tube plug which can be installed into the leak-off port(s) as a one cycle contingency if the pressure test reveals degradation in the leak-off piping. Also, obtain AREVA estimate to perform the pressure test and install the rolled tube plug(s) (if required) as a turnkey effort in 1R27. Once information is obtained, present the SlPD to the ECRG for fundi ng of the AREVA contract(s) and assignment/funding of the required engineering resources for EC development to support IR27 implementation.

Response

Subresponse :

Closure Comments:

Entergy I CORRECTIVE ACTION I CR-AN0-1-2008-02560 CA N umber: 26 G roup Name Assig ned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Barborek, W Douglas 5/13/2015 16:39:3 1 Performed By: Barborek, W Douglas 7/16/20 15 10:58: 10 Subperformed By:

Approved By:

Closed By: Woodson P.E.,Timothy R 7/16/2015 15:56:40 Current Due Date: 07/24/2015 Initial Due Date: 07/24/2015 CA Type: CAT C-CORRECT CA Priority: 3 Plant Constraint: NONE CA

Description:

Contact AREY A and determine if sufficient AREY A resources are avai lable to suppori qualification of the rolled tube plug contingency in time to support I R26 PO Milestone P0-16 (2/4/2016). lfnot, determine if AREVA can support a date after the P0-16 Milestone but prior to the start of 1R26.

Response

CA REQUEST:

Contact AREY A and determine if sufficient AREYA resources are available to support qualification of the rolled tube plug contingency in time to support LR26 PO Milestone P0-16(2/4/2016). If not, determine if AREY A can support a date after the P0-1 6 Milestone but prior to the start of l R26.

CA RESPONSE:

Per the attached e-mail from AREVA, AREY A believes the rolled tube plug option is viable for I R26. The I R26 PO Milestones may not be able to be met: however, due to the current RV flange inner o-ring leakage and the inability to isolate it due to the unresolved Davis Besse OE, System Engineering is going to pursue seeking approval for a I R26 pressure test and approval to develop the rolled tube plug contingency with AREYA. CA-27 will drive System Engineering to re-present SIPD Record 5678 to the ECRG once again for IR26 consideration. Based on the results of the ECRG meeting, additional Corrective Actions will be initiated as required to perfom1 the pressure test and contingency measures in I R26 (or IR27), as determined by the ECRG.

T?ve reviewed the CA response using the CA Quality WILL Sheet and EN-Ll-102 and conclude that it is appropriate for closure.

WDB 7/ 16/2015 Subresponse :

Closure Comments:

Concur with sub-response, action may be closed.

Attachments:

Response Description e-mail AREVA to Barborek dated 7-9-20 15

Attachment Header Document Name:

untitled Document Location

!Response Description Attach

Title:

~ -mail AREVA to Barborek dated 7-9-2015

BARBORB<, W DOUGLAS From: S<UUNA Dave (AREVA) <David.S<ulina@areva.com>

Sent: Thursday, JJly 09, 2015 8:49 AM To : BARBOREK, W DOUGLAS Subject : RE: ANO Leak off line - question I have not yet. Due to drrumstances, the critical personnel are not immediately available. I will get you a status of the proposal effort A'2:AP. I apologize for the delay.

However, I am confident that we can support the effort during your next roieduled outage.

Fegards, Dave David Skulina Project Manager, Business Development AREVA, Inc.

3315 Old Forest Road (OF53)

Lynchburg, VA 24501 Office: 434-832-2621 Mobile: !(b)(6) I

~ Please consider the environment before printing this e-mail o ert and is intended solely for the addressees. Re roduction and From : BARBOREK, W DOUGLAS [2]

Sent: Thursday, July 09, 2015 9:24 AM To: SKULi NA Dave (RSIIB)

Subject:

RE: ANO Leak off line - question Dave, Have you heard anything from AFEVA 81gineering & OJtage Sarvices regarding the feasibility of developing/qualifying the rolled tube plug design for our Sapt 2016 FFO? I would assJme that we need to get started on this effort pretty soon. In light of our current situation, I think I could serure the funds to make it happen.

Qment R:Sunidentified leakage has been holding fairly steady over the last week around 0.070 gpm. We are about 0.030 to 0.035 gpm higher than a month ago .... so about doubled.

Thanks ,

Doug Barborek 8'1tergy Q:ierations, Inc./ Arkansas Nudear Ole S,,stem Eiigineer - AN0-1 & AN0-2 ~actor ():)olant S,,stems and AN0-1 !:pent Fi.Jel ():)oling & Purification S,,stem S,,stem 8'1gineering Building / N-SY&4 wbarbo1@entergy.com 479-858-4337 (office)

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Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00026 Version: Approved: r,/

Requested Duedate: 06/25/2015 Previous Duedate: 06/ 12/20 15 Requested By: Barbo rek,W Douglas 06/ 11/20 15 Approved By: Woodson P.E.,Timothy R 06/1 1/201 5 Reque st Descriptio n:

AREVA (David Skuli na) has been contacted concerning the feasibility of developing the rolled tube plug contingency in time for I R26 execution. A response from AREYA E ngineering and AREY A Outage Services is still pending . As such, this CA is be ing extended two weeks to obtain a response from AREVA. T his CA is admin istrative in nature and the extension of the due date by two weeks to 6/25/201 5 has no impact on installed plant SSCs. As such, the DOE is necessary and acceptable.

WDB 6/ 11/2015 Approved

Description:

I st ODE approved for this co rrective action, work is neari ng completion.

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00026 Version: 2 Approved: r,/

Requested Duedate: 07/24/2015 Previous Duedate: 06/25/2015 Requested By: Barborek,W Douglas 06/19/2015 Approved By: Woodson P.E.,Timothy R 06/2 1/2015 Request

Description:

AREVA (David Skulina) was contacted again on 6/19/2015 concerning the feasibility of developing the rolled tube plug contingency in time for I R26 execution. A response from AREYA Engineering and AREYA Outage Services is still pending at this time, and the CA assignee will be offsite during the week this CA is actually due (due on 6/25/2015). As such, this CA is being extended to 7/24/2015 to allow additional time to obtain a response from AREYA and to account for the availability of the CA assignee, who will be out of the office for two of the next three weeks. This CA is administrative in nature and the extension of the due date to 7/24/2015 has no impact on installed plant SSCs. As such, the DDE is necessaty and acceptable.

WDB 6/19/2015 Approved

Description:

DDE approved.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 27 G roup Name Assigned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Barborek, W Douglas 5/13/2015 16:39:42 Performed By:

Subperformed By:

Approved By:

Closed By:

Current Due Date: 08/27/2015 Initial Due Date: 08/28/2015 CA Type: CAT C-CORRECT CA Priority: 3 Plant Constraint: NONE CA

Description:

Based on the results ofCA-26, revise and present SlPD Record 5678 to the ECRG once again for lR26 consideration. Based on the results of the ECRG meeting, initiate additional Corrective Actions as required to perform the pressure test and contingency measures in lR26 (or 1R27), as determined by the ECRG.

Response

Subresponse :

Closure Comments:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00027 Version: Approved: r,/

Requested Duedate: 08/28/2015 Previous Duedate: 06/26/20 15 Requested By: Barborek,W Douglas 06/ 19/2015 Approved By: Woodson P.E.,Timothy R 06/2 1/2015 Request Descriptio n:

This CA is dependant on the response to CA-26. CA-26 has been extended to 7/24/20 15. Therefore, this CA requires extension and is being extended to 8/28/2015 to allow time to update SIPD record 5678 and to make the presentation to the ECRG. The extension of the due date to 8/28/2015 does not have any deleterious impact on any installed plant SSCs and is therefore acceptable. WDB 6/ 19/2015 Approved

Description:

ODE approved.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA N umber: 28 G roup Name Assig ned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems NSSS StaffANO Barborek, W Douglas Subassig ned To :

Originated By: Woodson P.E.,Timothy R 5/21/2015 16:47:10 Performed By: Barborek, W Douglas 6/4/2015 12: I 0:34 Subperformed By:

Approved By:

Closed By: Woodson P.E.,Timothy R 6/4/2015 14:28:22 Current Due Date: 06/05/2015 Initial Due Date: 06/05/2015 CA Type: GENERAL ACTION CA Priority: 5 Plant Constraint: NONE CA

Description:

CR-ANO-L-2015-01950 was Administratively Closed to this CR. As Responsible Manager for this CR, ensure tbat the condition documented in that CR is appropriately addressed within the scope of this CR's Corrective Action Plan.

CR-AN0-1-20 15-0 1950 Condition Desc:

CR-ANO-l-20 15-0 1936 was written to document a K09-Fl (Reactor Vesse l Head Gasket Leak) alarm received at 0713 on 4/25/2015. If determined to be a valid afarm , this would indicate that leakage is occurring past the inner o-ring (gasket) on the Reactor Vessel Closure Head (R VCH). This CR is being initiated to make the site aware that in the event that actual RVCH o-ring leakage is confirmed to be occurring, the RV Gasket Leak Detection Isolation Valves RBS- I and RBS-2 cannot be closed to isolate the leak since the 2003 Davis Besse OE concerning potential chloride induced stress corrosion cracking (SCC) of the RVCH Leak-off lines has not been resolved for ANO-I (ref. open CR-AN0-1-2008-02560 and closed CR-AN0-1-2005-0 1140). Since this OE has not been resolved, the two RVCH leak-off lines should not be allowed to pressurize by closing RBS- I and RBS-2 until the integrity of the piping is confirmed via a refueling outage pressure test of the piping (CCC-6 line class) at its Design Pressure (2500 psig). Jfthe structural integrity of the piping is in fact degraded by SCC, failure of the piping due to pressurization could ultimately result in boric acid corrosion of the external surface of the Reactor Vessel. For this reason, inner o-ring leakage cannot be mitigated by closing RBS-I and RBS-2 and thus shifting the boundary to the outer o-ring and leak-off piping. The inability to isolate an inner o-ring leak could result in the leak becoming increasingly worse and result in a plant shutdown due to exceeding Tech Spec leakage limits and could potentially result in erosion of the o-ring/gasket seating surface.

System Engineering initiated W0-00195437 (WR-OOI64196) in May 2009 to perform a pressure test of the two RVCH Leak-off lines during IR22. The pressure test has been continually deferred and is currently not scheduled to occur until IR27 due to lack of allocation of funds and resources to develop the contingency measures required in the event degradation is discovered during the pressure test. The contingency measures include a (one operating cycle) qualification ofa rolled tube plug which would isolate one or both of the leak-off lines at the (ASME Class I) RV leak-off port(s) until repairs could be enacted during the subsequent refueling outage. It is noted that the subject piping is inaccessible and would require destrnctive removal of portions of the permanent reactor cavity seal plate, removal of the concrete shield blocks around the RV, and removal of RV insulation in order to access the subject piping for repair/replacement.

System Engineering has previously initiated SIPD Record 4955 ( I 1/ 18/2009) and SlPD Record 5678(5/5/20 11) to request funding/resources for the development of the pressure test and required contingency measures. During the 1/ 15/20 l 5 ECRG meeting, the pressure test was once again deferred to IR27 due to the lack of available Engineering resources to oversee development of the required contingency measures. Note that since the initiation of the 2003 Davis Besse OE, RVCH leak-off piping degradation has been discovered at other B&W, CE and Westinghouse PWRs.

This CR, like previous CR-AN0-1-2005 -01140 and CR-ANO- l-2008-02560, does not idcntify known degradation of a plant SSC, but rather the potential for degradation of a plant SSC based on industry OE.

Entergy I CORRECTIVE ACTION I CR-AN0-1-2008-02560

Response

CA REQUEST:

C R-AN0- 1-20 15-01950 was Administratively C losed to this CR. As Responsible Manager for this CR, ensure that the condition documented in that CR is appropriately addressed within the scope of this C R's Corrective Action Plan.

CA RESPONSE:

CR-AN0- 1-20 15-01950 identifies the same condition originally documented in this CR (CR-AN0-1-2008-02560). As documented in CA-25, the pressure test of the leak-off lines is currently not scheduled until 1R27. In summary, the current CA P lan outlined in CR-AN0-1-2008-02560 is:

CA Obtain a cost estimate from AREY A to design/qualify the rolled tube plug contingency to be used following pressure testing of the leak-off piping and obtain funding to move forward with contingency development; CA Determine if AREYA could support performing the rolled tube plug design/qualification in a timeframe suitable to support I R26; CA If AREVA can support 1R26, re-present SJPD Record 5678 to ECRG for I R26 consideration instead of I R27.

Based on the condition documented in C R-AN0- 1-20 15-02 179 (RV inner o-ring appears to be leaking) and the current CA Plan for CR-AN0-1-2008-02560, the fo llowing additional corrective action has been initiated to drive the replacement of the leak-off detection line and to permanently remove and cap the pressure test connection line to ultimately resolve the Davis Besse (and other plants) O E, regardless of whether or not the piping passes a future one-time pressure test. This action is ultimately required since the likelihood of having induced chloride-induced intergranular stress corrosion cracking (lGSCC) is considered high based on past ANO- I operating practices. T he performance of a one-time pressure test will not ensure that future wetting & pressurization of this piping does not result in further propagation and fa ilure ofan existing lGSCC flaw.

CA Initiate SIPD Record to replace the leak-off detection line and to permanently remove and cap the pressure test connection line to u ltimately resolve the Davis Besse (and other plants) OE, regard less of whether or not the piping passes a future one-time pressure test. This action is ultimately required since the like lihood of having induced chloride-induced intergranular stress corrosion cracking (IGSCC) is considered high based on pas t ANO- I operating practices (i.e. water left in pressure test connection piping for 30+ years). The perfonnance of a one-time pressure test will not ensure that future wetting & pressurization of this piping does not result in further propagation and fai lure of an existing IGSCC flaw . Initiate additional corrective actions as required to obtain approval the SIPD record scope.

The issuance of CA-29, a long with the balance ofopen corrective actions for CR-AN0-1-2008-02560, constitutes an acceptable CA plan for the condition documented in C R-AN0- 1-2015-0 1950. Additional corrective actions will be issued as required as the existing corrective actions are completed.

WDB 6/4/2015 Subresponse :

Closure Comments:

Concur with response, action is complete.

Entergy I CORRECTIVE ACTION ICR-AN0-1-2008-02560 CA Number: 29 G roup Name Assigned By: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Asstgned To: Eng Systems & Comps Mgmt ANO Woodson P.E.,Timothy R Subassig ned To : Eng Systems NSSS Staff ANO Barborck,W Douglas Originated By: Woodson P.E.,Timothy R 6/4/2015 14:27: 13 Performed By:

Subperformed By:

Approved By:

Closed By:

Current Due Date: 08/06/2015 Initial Due Date: 08/07/2015 CA Type: GENERAL ACTION CA Priority: 5 Plant Constraint: NONE CA

Description:

Initiate SJPD Record to replace the leak-off detection line and to permanently remove and cap the pressure test connection line to ultimately resolve the Davis Besse (and other plants) OE, regardless of whether or not the piping passes a future one-time pressure test. This action is ultimately required since the likelihood of having induced chloride-induced intergranular stress corrosion cracking (IGSCC) is considered high based on past ANO-I operating practices (i.e. water left in pressure test connection piping for 30+ years). The performance of a one-time pressure test will not ensure that future wetting & pressurization of this piping does not result in further propagation and fai lure of an existing lGSCC flaw. Initiate additional corrective actions as required to obtain approval the SIPD record scope.

Response

S ubresponse :

Closure Comments:

Entergy I CA DUE DATE EXTENSION ICR-AN0-1-2008-02560 Corrective Action : CR-AN0-1-2008-02560 CA-00029 Version: Approved: r,/

Requested Duedate: 08/07/2015 Previous Duedate: 07/03/20 15 Requested By: Barborek,W Douglas 07/02/2015 Approved By: Woodson P.E.,Timothy R 07/02/2015 Request

Description:

Due to current base workload and emergent plant issues, this CA could not be completed by the assigned due date and must be extended. Extension of the due date to 8/8/20 15 will not deleteriously impact any installed plant SSCs and is therefore acceptable. WDB 7/2/20.15 Approved

Description:

DDE approved.

From: Barrett, Andy To: Tindell Brian; Tobin Margaret; Choate 1ackson; BenneJJ Ma,:y

Subject:

CRs 11-13 Date: Monday, Novemb er 14, 2016 7:39:19 AM Several CRs (not included below) from the effectiveness reviews offloading penetrations corrective actions Unit 1 Rx Vessel leakage not from 0-ring, or the expected degraded point.

Unit 1 A EOG Inoperable due to problem unloading - 3 CRs Missing concrete block in fire wall in fuel building Drain valve misaligned allowing BWSTwater to get into Decay Heat Vault Trend CR on mlspositionlng events CR-AN0-1-2016-04627 11/12/2016 5:57:09 AM 11/12/2016 6:38:56 AM Barborek, W Douglas I Eng Systems NSSS Staff ANO The 3.5 inch segment of Reactor Vessel inner o-ring wh ich was sent to LPI (Contract 10494381) for forensic analysis was determined by LPI to not have a t hrough wall leak. Additionally, stereomicroscopic images of the o-ring segment revealed no appreciable loss of t he silver plating layer at the o-ring seating surfaces, thu s indicating that the location of the Cycle 26 inner o-ring leak w as not associated w ith this segment of o-ring. This segment of inner o-ring was taken from the leak-off port location between studs 52 and 53. LPI report is attached.

Cause determination for the Cycle 26 inner o-ring leakage is being performed under CR-AN0-1-2015-02967.

Discovered: Originated: Y Originator:

Operability:

Immediate Action:

Operability Version: Status: Approved By OP Code Performed By Date Performed Operability

==

Description:==

1 APPROVED M artin.Michael R ADMIN - NA Walls,Donald E 11/12/2016 3 :16:10 PM This condition report documents the condition of the Reactor Vessel inner o -ring t hat was sent off site for testing and is no longer an installed SSC.

This condition is ADMIN-NA IAW EN-OP-104.

REAP has been reviewed for immediate NRC reportab ility w ith respect to this condition IAW EN-Ll-108 and OP-1015.047, and no immediate reportability crit eria have been met.

Suggested Action: Inspect inner (and outer) o-ring grooves of the RV and RVCH for flaws under W0-00420217 Task 05 wh ich could result in o-ring leakage.

Non Responsive Record

Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record From: Barrett, Andy To: Tindell Brian; Tobin Margaret

Subject:

CRs Date: Tuesday, September 01, 2015 3:36:41 PM CR-AN0-1-2015-03240 Originator: Crosby,Patrick Group: Eng Code Programs Staff A Phone: 4903 Discovered : 8/31 /2015 3:50:48 PM Supv: Greeson.William C Initiated : 8/31/2015 4:01 :55 PM CR-AN0-1-2015-02179 CA-15 requires the Boric Acid Corrosion Control Program (BACCP) to perform a Boric Acid Evaluation for the U1 RVCH inner 0 -ring leak documented in CR-AN0-1-2015-02967. The evaluation is required to be documented in an EC (EVAL), sub-type (BOR) IAW EN-DC-1 15 to evaluate acceptability of the leak until the next refueling outage (1 R26 - Fall 2016).

During the evaluation process it was determined that unpredictable steam cutting and impingement at the Oring and reactor vessel flange area would neither allow for an acceptable evaluation nor provide justi1fication for extended operation without mitigating actions. Per EN-DC-319 (BACCP), periodic monitoring of the leak is required to mitigate and determine the condition of the stainless steel cladding of the reactor vessel and closure head.

The carbon steel reactor vessel and closure head are cladded internally with stainless steel. Stainless Steel is resistant to boric acid corrosion but is not immune to steam cutting or impingement. Per industry tests documented in EPRI Report No. 1000975, steam cutting and impingement could degrade the cladding and expose carbon steel.

Based on the last visual inspection performed on 8/4/2015 and the lack of discoloration of boric acid

crystals, the steel cladding is intact and performing its intended function.

The ODMI and Critical Decision document for CR-AN0-1-2015-02179 reference SER 3-09, which is industry OE regarding 0 -ring leakage at Browns Ferry Unit 1 in 2008. This OE documents that Browns Ferry Unit 1

experienced 0-ring leakage for 18 months across both the inne,r and outer 0-rings, which resulted in an increase in unidentified Drywell leakage from Oto 1.2 gpm over the course of the 18 month operating cycle.

Following removal of the RV head, damage to both the RV and RVCH seating surfaces from steam cutting was observed . Regarding SER 3-09 applicability to AN0-1, it is noted that Browns Ferry Unit 1 is a BWR, which utilized non-borated water in the RCS and operates at a nominal RCS pressure of 1000 psig. As

such, the observed damage at Browns Ferry Unit 1 resulted from a 1000 psig delta-P across two 0-rings, whereas AN0-1 is currently experiencing a 2155 psig delta-P across one 0-ring. As such, degradation of the AN0-1 RV and RVCH flange 0 -ring seating surfaces could accelerate at a much faster rate than the Browns Ferry Unit 1 scenario.

None Perform a containment entry every 30 days to verify the stainless steel cladding is intact by observing the drain piping for discoloration, such as, red or brown boric acid crystals indicative of carbon steel corrosion. The last visual observation was performed on 8/4/2015. Perform next inspection during the same power entry when adding oil to RCPs scheduled 9/15/15.

Consider updating current ODMI to take action upon discovery of discoloration in boric acid crystals.

Consider updating the current operability to OP-DNC or OP-Comp Meas. with monitoring as the compensatory measure.

Suggested Action

Description:

Immediate Action

Description:

Condition

Description:

Operability Required: Y Reportability Required: Y OPERABILITY VERSION:

1 PERFORMED BY:

Ward,Daniel E PERFORMED DATETIME:

8/31 /2015 4:20:48 PM INITITIAL REPORTABILITY:

NOT REPORTABLE OPERABILITY CODE:

ADMIN - NA OPERABILITY DESCRIPTION This condition report identifies a Boric Acid Evaluation could no1 be completed for CR-AN0-1-2015-CR-AN0-1-2015-03243 Non Responsive Record

Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record Non Responsive Record From: Haagensen, Brian To:

Cc: Gray Harold: Schroeder Daniel

Subject:

FW: FW: Oversized 0-Ring Mod Note: This 14-page attachment is withheld in its Date: Wednesday, December 13, 2017 11 :45:00 AM entirety under FOIA exemption 4.

Attachments: oversized o-ring Mod EC 72951,doc Attached is the EC justification for installing the oversized 0-rings at Unit 2. I can provide any of the references upon request.

From: Haagensen, Brian [ma i1to:bhaag90@entergy.com]

Sent: W ednesday, December 13, 2017 11:31 AM To: Haagensen, Brian <Brian. Haagensen@nrc.gov>

Subject:

[Externa l_Sender) FW: Oversized 0-Ring M od Brian C. Haagensen Senior Resident Inspector Indian Point Energy Center 914-739-9360 (Office)

!(b)(6)  !(cell)

In p lant x5347 From: LoPiccolo, Angela Marie.

Sent: Wednesday, December 13, 2017 11:15 AM To: Haagensen, Brian Cc: Wittich, Walter

Subject:

Oversized 0-Ring Mod Hi Brian, I have attached the Topic Notes from the Oversized 0 -ri ng mod EC 72951 Rev 0. There is a P2E folder associat ed with t his mod which conta ins more documents. Do you have access t o P2E? If not, let me know if there are any addit ional documents you requ ire and I'll send them over!

Angela LoPiccolo Engineeri ng NSSS X5657

From: Taylor, Nick Sent: Tuesday, July 21, 2015 11:30 AM To: Burritt, Arthur Cc: O'Keefe, Neil; Tindell, Brian; Correll, Brian; Egli, Richard; Tobin, Margaret; Dixon, John

Subject:

FW: Unit 1 RCS Leakage ODMJ Attachments: RV Flange 0 -ring Leakage.pdf Note: This 8-page attachment, EN-OP-111 Rev .11 , is withheld in its entirety under FO IA exemption 4.

Art, You were looking for OE last week on operations with both inner and outer O rings failed. I hadn't discovered any previous occurrences. But we got ANO's ODMI today and the licensee lists a number of previous occurre nces both at ANO, Browns Ferry 1 and Wolf Creek.

Hope this helps, Nick Taylor Senior Project Engineer Division of Reactor Projects USNRC Region IV 0: 817 200-1520 C: (b)(6)

E: nick.taylor@nrc.gov From: Correll, Brian Sent: Tuesday, July 21, 2015 9:11 AM To: OKeef e, Neil; Taylor, Nick; Farina, Thomas; Tindell, Brian; Tobin, Margaret Cc: Egli, Richard

Subject:

Unit 1 RCS Leakage ODMI Unit 1 RCS Leakage ODMI is attached. Pages 5 and 6 are the Thresholds and Actions to be taken if the thresholds are exceeded.

Brian

F,om: rlwWLI!lio To: ~

Cc: ~ ~ T o b i n Margacer ~ Choa1e 1acksoo: George Andrea.

Subject:

FW: ANO~I RCS Leak Rate P,ojectlon Date: Thursday, August 20, 2015 2:10:55 PM Attachments: illw!<006.on,;

lmag.007,ong FYI - see graph below.

From: BARBOREK, W DOUGLAS (mailtoWBARBOl @ent ergy.com]

Sent: Thursday, August 20, 2015 10:42 AM To: Tindell, Brian Cc: WOODSON, TIMOTHY R

Subject:

[Ext ernal_Sender) FW: AN0-1 RCS Leak Rate Projection Brian, FYI. See latest projection below regarding RV Flange o-ring leakage.

Thanks, Doug Barborek Enterav O~mloni, tnc I Atk1nwi Nucle,r Ooe System Engineer - AN0-1 & AN0-2 Reactor Coola nt System s and AN0-1 S~nt Fuel Coo 1ng & Purifw;atlon Syst em Sv,.tem Engineering Bu ld1ng/ N *SY(.4 wtarhol@enlerrv rorn A79 8~8 4337 (office)

~ (p.age-11

~ RBOREK, W DOUG LAS Sent: Thursday, August 20, 2015 10:41 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD, JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY 8; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL,JERRY W; HILL, STEVEN D; Parker, Bobby J oe; CRANE, NELSONS; Pace, Robert D. (INPO) (paceRD@INPO oq:)

Subj ect : RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 94 days from now, or November 22nd, 2015. This proJection is about two weeks beyond the last three weeks of projections due to a slight leveling off of the Total RCS leak rate over the last week.

Thanks, Doug Barborek lntetQ'f O~r.HIOn~, inc / Arkans.as Nut le,1, One Svstem Eng1n,eer - AN0-1 & AN0-2 Reactor Coolant Systems and AN0* 1 Spem r ut l Coolin& & Pur 1ficat1o n Sysl f'm Svstem Eng1n.eenng Bu ld1ng/ N-S'fE-11 wbachol@enrer@v oom 479-SSS-433' (oflk:*)

!(b)(6) !1 1

RC s Leakrate Cycle 26

- - OTFlil~tt(Oon,)

Unld4nDlto ltak Flalt ( IJIN'l1)

--Tolal RCS LC,..!(OPm)

- -wean*2 S4oma

- - l,ltan-2 Sigma

- 1.1,111*)$Ao,n1 0 200 - -1,1,a,,.JSlgma

- - --* liltM

--Poly (T<>lalRCS Los,..(o,,lll))

As of 8 20/2015, the 2nd order 0 150 potynoml1I projection to reach 0.250 Total RCS Leakce Is 94 days from now, or 11/22/2015 0 100 O050 0 000

"'0 "' "' "' "'0 "' "' ...0 "' ...ij ... ...0 "' "' "'0 ...

<:I § § s 5 5 I <:I i § <:I Si Iii ~ s; la 5i $

i § ~

~

~ § :s<:I §

~ s ii From: BARBOREK, W DOUG LAS Sent : Thursday, August 13, 2015 3:1 5 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD, JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY 8; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS.JOHN M; HOWELL,JERRY w; HlLL, STEVEN D; Parker, Bobby J oe; CRANE, NELSONS; Pace, Robert D. (INPO) (PaceRD@INPO QC&)

Subject:

RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Total RCS Losses is 85 days from now, or November 6th, 2015. This is consistent with the last two weeks of pro1ect1ons.

Thanks, Doug Barborek Entergv Oper.itions, Inc./ Alkansas Nuclear One S~tem Enclneer AN0-1 Re.ictor Coolilnl System and ANO 1 Spent F-uel Coo 111a & Purlf1ut1011 Sy~tcm SV1lCrn Eogin('~ring Bu Id ins/ N SYE 4 wtmchol Pcnsecav mm 479-858-.4337 (office)

~ (p.]ffit>rl RC s Leakrate Cycle 25 0 300

--OTFII Ralt (OPIIII Unlltnllnt4 LtakRalt (ll)m)

- - Toc.i FlCS Lo*H IOPIIII

- - uean*2S.om.1

- - uun-2Sio..

- uean*l s,grna 0200

- uun,3Stgma

... --*~Mean

..~

C From: BARBOREK, W DOUGLAS Sent : Thursday, August 06, 201 5 11 :48 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY 8; Davis, Barry; PUTNAM, REX G Cc: MEYERS,JOHN M; HOWELL,J ERRY W; HlLL, STEVEN D

Subject:

RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0 .250 gpm Tota l RCS Losses is 91 days from now, or November 5th, 2015 (ref. attached ODMI, Rev. 01). This Is based on leak rate data from 4/1/2015 to present, which corresponds to the approximate date that t he total RCS leakrate began increasing. See projection graph below.

Note that the nominal inner o-ring leak rate is determined by adding the increase in t he RB Sump fill rate since -4/1/2015 to the increase in the T-111/T-42 fill rat e since -6/22/2015. The increase in the Unident ified leak rate is NOT indicative of the total o-ring leakage rate since some of the o-rlng leakage 1s apparent ly being condensed in the RCP Seal Collection system loop seal and is bemg returned to the Quench Tank as Identified leakage.

Thanks, Doug Barborek Ent*rav O~ratlons, Inc: / Atkanus NuclHf Ona System Engineer - AN0-1 & AN0-2 Reactor Coola nt Systems and AN0-1 S~nt Futol Coo 1n1 & Puril1Cation Sy11trm SV11t!m Er,glni!:t!rin.g Bu ld1ng / N-SYE-4 wtzachqJ@rntrrev mm I

479 8S3-4337 (office)

!(b)(6) ,,.g.,J

RCS Leakrate Cycle 25

____J 0.300 ,- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . .

- - OT Fil Rate (gpm)

- - Unldentlfled Lea~ Rate (QPm)

- - Total RCS Losses(gpm)

- - Mean+2 Sigma

- - Mean-2 Sigma

- Mean+3Slgma 0 200 - Mean-3 Sigma

Mean

--Poly. (Total RCS Losses (gpm)) As of 8/ 6/ 2015, the 2nd order 0 150 +-~~~~~~~~~~~~~~~~~~~~-tr-c:,...,,::...-..IL..~~~~~~-..:.p_o_ly , n_o~m

-==ia~l~p-r_o ~

j*~c~ti~o~nr.t~o==-~~

rue 0.250 gpm Tota RCS Leakag* Is 91 days from now, or 11/5/2015 0.0 00 0 "' "' "' "'

"' "' "' 0"' "' "' "' "' "' "' "'

c::::! ~ ~ OJ~ 0c::::! <D~

0

~ ~ ~ ~ 0~ 0~ ~ ,.._~ 0~

~ "' C:! ~ "'

C:! ~ ;a <D C:! C? ,.._ C:! ~

~ ~ ~ ~ .....

"' "' <D From: BARBOREK, W DOUGLAS Sent: Friday.July 31, 2015 10:24 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B Cc: MEYERS, JOHN M; HOWELL, JERRY W; HILL, STEVEN D Subj ect : AN0-1 RCS Leak Rate Projection All, Obviously, t his is not an exact science, but the latest 2nd order polynomial projection for reaching 0.250 gpm Tot al RCS Losses is 99 days from now, or November 7, 2015 (ref. attached OOM I, Rev. 01). This is based on leak rate data from 4/1/2015 to present, which corresponds to t he ap proximate date that the total RCS leakrate began Increasing. See projection graph below.

FYI, t he same projection two weeks ago calculated a date of 12/19/2015 to reach 0.250 gpm. It is anticipated t hat this va lue will cont inue to move towards the present.

I will be participating in a 2nd Power Entry o n 8/ 4/ 2015 to quantify the leakage at t he drain header near the west end of the RB Sump.

I will attempt to update th is projection on a weekly basis from this point forward .

Thanks.

Doug Barborek Enterg'( Oper;itions, Inc / Atkanws Nuclear One System En&lneer- AN0-1 & AN0*2 R.o ctor Coolant Systems

,;ind AN0-1 $pent Fue l Coolmg & Pur ificatio n System SV$11m Enaineerin.a: Bu1ld1n1/ N,$YE*4 wtachoJ@rnsecev mm 4 79-8 58-4337 (o.ffice)

(b)(6) 1 r,.i.. l

RCS Leakrate Cycle 25 0.300 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

- - QT Fil Rate (gpm)

- - Unidentified Leak Rate ( gpm)

- - Total RCS Losses (gpm)

- - Mean+2 Sigma

- - Mean-2 Sigma

- Mean+3Sigma 0-200 - Mean-3 Sigma

  • Mean /u of 7/30/2015, the 2nd

- - Poly. (Total RCS Losses (gpm)) order polynomial 0_150 - - - - - - - - - - - - - - - - - - -- ~ " " " ' - - - - - - - -....r::.':..:o:..*e=:c;-:t.:io . :::..:,n;.-:t;.:o:,,:rce::.:a.:.

ch:,:..:.0.:.2:.:

. :: S.:.

o_ __,

gpm Tota l RCS Leakage is 99 days from now, or 11/7/2015.

0.000

"'0 "'0 "' "' "' "'0 "' "' "' "'0 "' "' "' "'0 "'0 "' "' "'

~ ~ ~ ~ ~ ~ ~ ~ ,.._ ~ ~ 0~ .....~ ...~ ~ ~ ~ ~N a,~

,:: lo? "' "' a, 112 (") 0 1::1  ;:: .....lo? "';:::: .....C!. .....C!.

.... .... ~ C!.... ...C!. "' in C!. C!.

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(D

From: IiilllfJWaJl To: ~

Cc: llixwl..Jl!ll!l umll..J!iwJ; E*ao* Jbomas* Iobio Mac*acer George Aoacea, ~ Note: The same 8-page attachment as prior string

Subject:

FW: AN0-1 RCS Leak Rate Projection Wednesday, August OS, 201 S 11 :04:31 AM

( EN- OP- 111 ReV 11 ) and anOth er 6-page att aChment Date:

Attachments: OOMLl!<v.OJ. RV_Flil!ll<tOc11Qtlea_lL- aomll>vOoill2Ll>d( (EN-FAP-OM-021 Rev 1)- have been withheld in their lmageoo, .png

- - - - - - - - - - - - - - - - - - - - - - - 1 entirety under FO IA exemption 4.

Neil, I just got done talking with the Unit 1 RCS system engineer. He showed me the below graph, which is a great illustration for the ANO Unit 1 inner vessel 0 -ring leakage trend.

He guesses, which seems right assuming nothing changes, that ANO w ill hit their arbitrary .25 gpm total leakage threshold during the Unit 2 outage. Because they won't want to have both units offline at the same time, they will likely keep Unit 1 online at least until after the Unit 2 outage.

More to come, Brian From: BARBORE K, W DOUGLAS [mai1to:WBARB0l@entergy.com]

Sent: Wednesday, August 05, 2015 11:00 AM To: Tindell, Brian

Subject:

[External_Sender] FW: AN0-1 RCS Leak Rate Projection Brian, I will include you on future updates, which I hope to put together weekly from this point forward.

Please let me know 1fyou have any questions regarding this information.

Thanks, Doug Barborek Entergy Operat '°ns, Inc./ Arkansas Nudear One Sys1em En1ineer AN0-1 & ANO* 2 Reactor Coolant Systems and ANO l Sptnt Fuel Coohr'II & Pur1f1catton Sy!ot*m System Engineering Building/ N-SYE-4 wtmbe1@COICCRY ,em 479-85S-4337 (office)

~ (pager!

~ RBOREK, W DOUGLAS Sent: Friday. July 31, 2015 10:24 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EOMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B Cc; MEYERS, JOHN M; HOWELL, JERRY W; HILL, STEVEN D

Subject:

AN0-1 RCS Leak Rate Projection All, Obviously, this is not an exact science, but the latest 2nd order polynomial projection for reaching 0.250 gpm Total RCS Losses is 99 days from now, or November 7, 2015 (ref. attached ODMI, Rev. 01). This is based on leak rate data from 4/1/2015 t o present, which corresponds t o the approximate date that the t otal RCS leakrate began increasing. See projection graph below.

FYI, the sa me projection two weeks ago calculated a date of 12/19/2015 to reach 0.250 gpm. It is anticipated that t his value will continue to move towards the present.

I will be participating in a 2nd Power Entry on 8/ 4/ 2015 to quantify the leakage at the drain header near the west end of the RB Sump.

I will attempt to update this projection on a weekly basis from this point forwa rd.

Thanks, Doug Barborek

[r'\lCf8V 0J)er'tll0r'IS, Int./ A, kansas Nutlta( Ont System Engineer - AN0-1 & AN0*2 Reactor Coolant Systems and ANO*l Spent Fuel Cooling & Punflcat t0n Sv~tem System Englt1ee,1ne Bulldln.f; I N-SYE-A wt,;irho)@PnlPrgy.m m

RCS Leakrate Cycle 25 0-300 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

- - OT Fil Rate (gpm)

- - Unk!enbfied Leak Rate ( gpm)

- - Total RCS Losses (gpm)

- - Mean*2Sigma

- - Mean-2 Sigma

- Mean+3Sigma 0200 - Mean-3 Sigma

  • Mean A., of 7/30/2015, the 2nd

- - Poly. (Total RCS Losses (gpm)) order po lynomial o_150 + - - - - - - - - - - - - - - - - - - - - --tr-=-' - - - - - - - - - - - ---'p'--r_o.,_je-=c=-t_io_n:-t::-0::-:-re,...a_c-=h_0_._2...,so_ ___,

gpm Total RCS Leakage is 99 days from now, or 11/7/2015.

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F,om: Iillll.ilL..adi To: '1:.lllilLlllll Cc: 01mn....J2hn: ~ Choate 1acksoo* ~  ; Tottln Ma,garet

Subject:

FW: ANO-I RCS Leak Raie Ptojectlon Date: Thursday, S*ptember 03, 2015 1:23:02 PM Attachments: ima£e005.one.

lmagt006.png Neil, FYI below.

Thanks, Brian From: BARBOREK, W DOUGLAS (mailto:WBARBOl@entergy.com)

Sent: Thursday, September 03, 2015 1:08 PM To: Tindell, Brian

Subject:

[Ext ernal_Sender) FW: AN0-1 RCS Leak Rate Projection Brian, The latest update.

Thanks, Doug Barborek fnte1sv Optm1on,, tnc / Ark1n\a\ NuclNr On£>

System Engineer ANQ.1 & AN0 -2 Re~tor Coolant Systems and AN0-1 Spent Fuel Cooling & Purification Syst em Sys11m E~1nci1r1n1 Bu ld1n1 / N SYE-4 wharh21@rowex rnm

'}'L"\!fU, (office)

!(hl(g.} ~ (p.gc,J From: BARBOREK, w DOUGLAS Sent: Thursday, September 03, 2015 1:04 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD. JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL.JERRY W; HILL, STEVEN D; Parker, Bobby J oe; CRANE, NELSONS; PACE, ROBERT D; OLIVER.JASON R; MEATHEANY, DANIEL J; SCHLUTER MAN, PAMELA S; Crosby, Patrick; Beldin, Joshua; GREESON, WILLIAM C Subj ect: RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 114 days from now, or December 26, 2015.

Thanks, Doug Barborek fntt"f8V OJ)f'n111on1, Inc / Ark&n$.l1 Nut lt>ar Ont' System Engineer ANQ...1 & AN0 -2 Re~tor Coolant System~

and AN0-1 Spent Fuel Cooling & Purification Syst em Sy,11m Ert1Mlrlnl Bu Id 1n1 / N -SYE -4 whach21@rnrecrx row 479 85S 4337 (office)

~ IP.B*fi RCS Leakrate Cycle 26 0 300

- - OTFII Rall (9pm)

Unldenllfltd Leak Raie (,gpm)

- - Tola /RCS L0$SfS(OPIIIJ Mu n*2Sigma

- -Mu n-2S.gma

- tJun*J~oma 0 200 - -Mean-3 Sigma

-M oan

--PolJ' (T~ RCSL..sa (!IJ)m)) order polynomial 0 150 +-- - - - - - - - - - - - - . , , . . , .~ .dC=---------..c.:.r::. oi::e:.:::

ct.:.:l~on to ::..:.:

re:.:*:.:t::.:

h..:0::.:.2:.:S:.::O

__-I cpm Total RCS Leakage Is 114 days from now, or 12/26/201S From: BARBOREK, W DOUGLAS Sent: Thursday, August 27, 2015 5:46 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PE RKINS, DARRELL L; GORDON, ROBERT A; FORD.J EFFREY; Evans,

Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL, JERRY W; HILL, STEVEN D; Parker, Bobby Joe; CRANE, NELSON S; PACE, ROBERT D; OLIVER. JASON R Subject : RE: AN0-1 RCS leak Rate Proj ection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS l osses is 104 days from now, or December 9, 2015. This projection is beyond the last four weeks of proiections due to a slight leveling off of the Total RCS leak rate over the last week to 10 days.

Thanks.

Doug Barborek Ente,gy Operations, Inc: / Alkansas Nuclea, One S~tem lnglrlter - ANO J & ANO 2 Atactot Cool1nt SY')\ems Jnd ANO* l Spent fuel coo11ng & Purification System Svstem (nijll')t'erlng Bu Id 1ne / N*SYf*4 wbocbotd!eosec,v oom

~79-858-4337 (office)

!(b)(6) l'PJ..rf RCS Leakrate Cycle 25 G~~~~~~~~~~~~~~~~~~~~~~

OTF- Ratt ~pm)

Unidenbfte~ Leal< Rate (Q?tl1)

- - Total RCS lo .... (gpm)

- - Mtan*2 Sigma Mtan,2 Sigma

- Mean*3Ssgma 0200

- -Mtan-3S.gma

*tJean As of 8/27/2015, the 2nd

--Poly (TOUI RCS LOI- ~pm))

0 150 t-----------------r,_-.:z,;.-=-'----------:'o"-rd .::..e:..:r..;P 0;0.:..l-::

y':'

n':::

om 7 "1:*.:l-<pccro1 2..:

  • ec: 71::o'=n=---l ct to reach 0.2501pm Total RCS Leak11e is 104 devs from now, or 12/9/2014.

0100 ooso 0000 8"' "'§ "'~ 8"' "'~

i ;i s I I ~ 5i From
BARBOREK, w DOUGLAS Sent: Thursday, August 20, 2015 10:41 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL, J ERRY W; HILL, STEVEN D; Parker, Bobby Joe; CRAN E, NELSON S; Pace, Robert D. (INPO) (paceRD@JNPO or~)

Subject : RE: AN0-1 RCS Leak Rate Proj ection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 94 days from now, or November 22nd, 2015. This projection is about two weeks beyond the last three weeks of project ions due to a slight leveling off of the Total RCS leak rat e over the last week.

Thanks, Doug Barborek Ente<sv Optra11ons, 11"1(: / A<kansas Nuclea< one SV)tem l~l~tr-ANO. l & AN0*2 R.t~tot Coolant !>~tems and ANO* l Spent Fuel Coolins & Purification System Svs1em (n11lneeriF'g 8u lcl 1ne / N*SY£ *4 wbarbot@entersv.oom 479-858-4.UJ (offl~)

!{b)(6) 11,,a.,,

RCS Leakrate Cycle 25

- - OTF-Ralt(Op,11) un~n11t1olukRa11 t - J

- - TotalRCSlouu(gpm)

- - wean*2'Sio,na lolt.,..2$io,na

- Ytan*l~gma 0 200 - 1o11.,..3$i0,na

--- --- I.lean

--Poly (TOlalRCSlOl,. (Qll"1))

0150 - l - - - - - - - - - - - - - - - - - - ; . - : :,....=--'-----'-' A"'c s c"o'-

f 8= 2cc0cc/.;c 20 .c.1;;.,Scc'-'t-'-

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polynomlal projection to reach 0.250 Total RCS Leakce Is 94 days from now, or ll/22/ 201S From: BARBOREK, w DOUG LAS Sent: Thursday, August 13, 2015 3:15 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUN D; PE RKINS, DARRELL L; GORDON, ROBERT A; FORD.J EFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS. J OHN M; HOWELL.JERRY W; HILL, STEVEN D; Parker, Bobby J oe; CRANE, NELSONS; Pace, Robert D. (JNPO) (PareRD@)NPO org)

Subj ect: RE: ANO-1 RCS Leak Rate Projection

All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 85 days from now, or November 6th, 20 15. This is consistent with the last two weeks of projections.
Thanks, Doug Barborek

[r\lCtQ'f 0J)trj llOr'15, ln,c / Afk6r'l1.;l5 Nutlt,H 01\C S~tem £nglf'}Ccr - ANO 1 Re.actor C<lolant 5y5tcm and A'II0 -1 S1>tnt Fuel Cooltn8 & P1.mhcatlon 5vstem Svsttm (nglnttdng 8u ld1n9 / N*SY(-4 wti~rbol@enrergy com RCS Leakrete Cycle 26 0 JOO

--OTFII Rate (gpn,)

Undtn*neo Ltak Race (Ql>"11

- - Totot FICS losstS(Ol)n,)

1o1un*2Sooma

- -uean-2 Stoma 0200

- uun*J s.oma

- uun-3Stoma

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i§ ct Q!

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From: BARBOREK, W DOUG LAS

Sent: Thursday, August 06, 201 5 11 :48 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD, JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G Cc: MEYERS. J OHN M; HOWELL.JERRY W; HILL. STEVEN D Subj ect: RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 91 days from now, or November 5th, 2015 (ref. attached ODMI, Rev. 01). This is based on leak rate data from 4/ 1/ 2015 t o present, which correspo nds to t he approximate date that the total RCS leakrate began increasing. See projection graph below.

Note that the nominal inner o-ring leak rate is determined by adding the increase in t he RB Su mp f ill rate since -4/1/2015 to the increa se in the T-111/T*42 fill rat e since ~6/22/2015. The Increase in the Unidentified leak rate is NOT indicat ive of t he total o ring leakage rate since some of the o ring leakage is apparently being condensed in the RCP Seal CollectiOn system loop seal and is being returned to the Quench Tank as Identified leakage.

Thanks, Doug Barborek Ente<gy Optr.ttions, lne / A/kansas Nuclea, Ooe S~t~m Eoglnte,r-ANO l & ANO 2 RektOt Coolaol SV'\tC'mS and ANO 1 Spent Fuel (001mg & Pu<1f1C,1tlon System S~1em Et'IBl~ering euId 1nA / N-SYf-4 Yd:iatboJ@emer,v com 479-858-433? (office)

~ (pagH\

RCS Leakrate Cycle 25

_J 0.300 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

- - OT Fil Rate (gpm)

Unldentifled Leak Rate (gpm)

- -Total RCS Losses(gpm)

- - Mean*2 Sigma

- - Mean-2 Sigma

- Mean*3Slgma 0.200 - Mean-3 Sigma

Mean

--Poly. (Total RCS Losses (gpm)) As of 8/ 6/ 2015, t he 2nd order polynomial projection to 0 150 +--- - - - - - - - - - - - - - - - - t , - : : ; f ' ~ - - - " - - - - - - - - - ' :r"'e* .:".::-:c,:-;;

0-;;

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From: BARBOREK. W DOUGLAS Sent: Friday, July 31, 2015 10:24 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUN D; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERN ER, GEORGE W; HOLLOWOA, TROY B Cc: MEYERS. J OHN M; HOWELL.JERRY W; HILL, STEVEN D Subj ect: AN0-1 RCS Leak Rate Projection All, Obviously, this is not an exact science. but the latest 2nd order polynomial projection for reaching 0.250 gpm Tot al RCS Losses is 99 days from now, or November 7, 2015 (ref. attached OOM I, Rev. 01). This is based on leak rate data from 4/1/2015 to present, which corresponds to t he ap proximate date that the total RCS leakrate began increasing. See projection graph below.

FYI, the same projection two weeks ago calculated a date of 12/19/2015 to reach 0.250 gpm. It is ant icipated that this va lue will cont inue to move towards the present.

I will be participating in a 2nd Power Entry on 8/4/2015 to quantify the leakage at the drain header near the west end of the RB Sump.

I will attempt to update th is projection on a weekly basis from this point forward.

Thanks.

Doug Barborek Entergy 0~1ations, loc / Alkans.as Nucl!ar Ont Svs1em Engll'\ter- ANO, t 8, AN0*2 Reactor Coolant Svs1ems

and AN0-1 Spent Fuel Cooling & Purific ation S~tem S~tem Engineering Building / N-SYE-4 wNrhoJ@tnrrrev mm 479*853~337 (offia!)

I(bV6' 11"'8" 1 RCS Leakrate Cycle 25 0.300 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

- - OT Fil Rate (gpm)

- - Unidentified Leak Rate ( gpm)

- - Total RCS Losses (gpm)

- - Mean+2 Sigma

- - Mean-2 Sigma

- Mean+3Sigma 0.200 - Mean-3 Sigma

l~ean As of 7/30/2015, the 2nd

- - Pol)'. (Total RCS Losses (gpm)) order po lynomial 0 _1 50 +-- - - - - - - - - - - - - - - - - - - --tr- >-./<"=---- - - - - - - -..!p'--'r-=- o'-'

je:.:c.::t:i.:oc.cn:...t:.:o:...r:...:e:.:a:.:ccch_:0:...2

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gpm Tota l RCS leakage Is 99 days from now, or 11/7/2015.

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F,om: ~

To: llJlwU,tll Cc: Tobin Margaret* ~ OJx.c.n...J..Q.

Subject:

FW: ANO~1 RCS Leak Rate Ptojectlon Date: Thursday, AuguSI 13, 2015 3:27:22 PM Attachments: irM<e002.olUI; lmag.003.png FYI - Unit 1 RCS leak rate update.

From: BARBOREK, W DOUGLAS (mailto:WBARB01@entergy.com]

Sent: Thursday, August 13, 2015 3:26 PM To: Tindell, Brian

Subject:

[Ext ernal_Sender) FW: ANO* l RCS Leak Rate Projection Brian, FYI. I believe I told you I would forward these updates to you.

Thanks, Doug Barborek Entergy Oper.ations, Inc./ Alkansas Nuclear One Svstem En&lnter-ANO l & ANO 2 Rt.ctor Cool1nt SV$1Cm~

and ANO 1 Sp('nt Fue-1Cooling & Purlf1c.1t!on Systrm wtwholf'bemersv rom 419-858-A337 (offlce)

ITiiiZill]lpogerl From: BARBOREK, W DOUGLAS Sent: Thursday, August 13, 2015 3:15 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL, JERRY W; HILL, STEVEN D; Parker, Bobby Joe; CRANE, NELSON S; Pace, Robert D. (INPO) (PaceRD@INPO or~)

Subject:

RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 85 days from now, or November 6th, 2015. Th is is consistent with the last two weeks of proiections.

Thanks, Doug Barborek Entergy Oper.1tions, Inc. / Arkansas Nucle,1r One Sys11m Enc1nHr AN0.1 R*.;ictor Cool1nl Systf'm and AN0-1 Spent Fuel Cooling & Pur1t11:.1tion System Sntem Ef'llin-eerin.s Bu Id 1ng / N -SYE -4

'41WheJ@cmecev mm 479-8S84H7 (office)

~ IP.Serl RC s Leakrate Cycle 26 0 300

- - OTFII Ra1t(OP"II unad1n11n,o LnkR~* ( gpff1l

- - roca1 ~ LONU(OPml Unn*2S.gnw

- - Mtan*2StOMl 0200

- - uean*l s.oma

- - Mnn-3 Slgna

          • Utan From: BARBOREK, W DOUG LAS Sent: Thursday, August 06, 2015 11 :48 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD,JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G Cc: MEYERS. J OHN M; HOWELL.JERRY W; H.ILL, STEVEN D

Subject:

RE: AN0-1 RCS Leak Rate Proj ection

All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 91 days from now, or November 5th, 2015 (ref. attached ODMI, Rev. 01). This is based on leak rate data from 4/1/2015 to present, which corresponds to the approximate date that t he total RCS leakrate began increasing. See projection graph below.

Note that the nominal inner o-ring leak rate is determined by adding the increase in the RB Sump fill rate since -4/1/2015 to the increase in the T-111/T-42 fil l rate since -6/22/2015. The increase in the Unidentified leak rate is NOT indicative of t he total o-ring leakage rate since some of the o-ring leakage is apparent ly being condensed in the RCP Seal Collection system loop seal and is being returned to the Quench Tank as Identified leakage.

Thanks, Doug Barborek En ter gy Operations, Inc / Arkansas Nuclear One Systtm E"III\Hr AN().l & AN0-2 Rttctot Coolin~ System, and AN0-1 Spent Fuel Cooling & Pur iftcation System System e,,.in-eerin1 Bu ld1ng / N SYE-4 whachol @eo1ecev com 479 853 4337 (offitt)

!(b)(6) 11"'8"'

RCS Leakrate Cycle 25

__J 0.300

- - OT Fil Rate (gpm)

Unklentifted Lea~ Rate (gpm)

- - Total RCS Losses(gpm)

- - Mean*2 Sigma

- - Mean-2 Sigma

- Mean*3Slgma 0 200 - Mean-3 Sigma

Mean

- - Poly. (Total RCS Losses (gpm)) As of 8/ 6/ 2015, the 2nd o rder polynomial pro jection to 0 150 roac 0.250 1pm Tota RCS Leakage is 91 days from now, or 11/5/ 2015 0 100 0 050 0 000 or, or, or, or, or, or, or, or, or, or, or, or, or, or, or,

~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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From: BARBOREK, W DOUGLAS Sent: Friday.July 31, 2015 10:24 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD, JEFFREY; Evans, Terry Alan; ANO OPS1 $M's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B Cc: MEYERS.JOHN M; HOWELL.JERRY W; HILL. STEVEN D

Subject:

AN0 -1 RCS Leak Rate Projection All, Obviously, this is not an exact science, but the latest 2nd order polynomial projection for reaching 0.250 gpm Total RCS Losses is 99 days from now, or November 7, 2015 (ref. attached ODMI, Rev. 01). This is based on leak rate data from 4/1/2015 to present, which corresponds to t he approximate date that the total RCS leakrate began increasing. See projection graph below.

FYI, the same projection t wo weeks ago calculated a date of 12/19/2015 to reach 0.250 gpm. It is ant ici pated t hat this va lue will continue to move towards the present.

I will be participating in a 2nd Power Entry o n 8/4/2015 to quantify the leakage at the drain header near the west end of the RB Sump.

I will attempt to update this projection on a weekly basis from this point forward.

Thanks, Doug Barborek En tetg'( Operc1tions, Inc./ ArkanS.Js Nuclear One Sv,tem Entlnecir - AN0.1 & AN0*2 RtKtor Cooli n! Sv,tcmi and AN0-1 Spent Fu@ICool,ng & Puriflu tion Systt m System Engineering Bu1ld1ne / r*ViYE*4 wborhQIl!!cmrcev com 479 858--4337 (office)

!(b)(6) 11,.serl

RCS Leakrate Cycle 25 0.300 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

- - QT Fil Rate (gpm)

- - Unidentified Leak Rate ( gpm)

- - Total RCS Losses (gpm)

- - Mean+2 Sigma

- - Mean-2 Sigma

- Mean+3Sigma 0-200 - Mean-3 Sigma

  • Mean /u of 7/30/2015, the 2nd

- - Poly. (Total RCS Losses (gpm)) order polynomial 0_150 - - - - - - - - - - - - - - - - - - -- ~ " " " ' - - - - - - - -....r::.':..:o:..*e=:c;-:t.:io . :::..:,n;.-:t;.:o:,,:rce::.:a.:.

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gpm Tota l RCS Leakage is 99 days from now, or 11/7/2015.

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From: Tobin Margartt To: Choate 1ackso0* ~ ~ ~

Subject:

FW: AN0-1 Reactor Vessel Flange I nner 0 -rlng Leakage - Video rrom 7122/2016 Power Entry Date: Friday. July 22. 2016 11:18:00 AM Attachments: jma.ee001.on.e.

~oflloarisorl 6 entries.odf FYI From: BARBOREK, W DOUGLAS [mailt o:WBARB01@entergy.com)

Sent: Friday, July 22, 2016 11:16 AM To: Tindell, Brian Cc: Tobin, Margaret; Barrett, Andy ; PYLE, STEPHENIE L ; WOODSON, TIMOTHY R; SKARTVEDT, MARK EDMUND

Subject:

[External_Sender] AN0 -1 Reactor Vessel Flange Inner 0 -ring Leakage - Video from 7/22/ 2016 Power Entry Brian, For your information.

The (RV inner o-ring) leakage at t he HSD-lS-2" header was observed to be approximately 48 drops/minute during this morning's power entry. Boric acid deposits remain white with no discoloration. Initiated CR-ANO*l -2016-02183 to document observations.

A still shot from the videos from the past 6 power entries is attached. Cycle 26 RCS/ Sump leak rate info (as of today) is provided below .

Please let me know if you have any questions regarding today's entry observations. If you would like to see the video from today's entry, please feel free to stop by my cube.

Thanks, Doug Barborek Entergy Operations. 1nc. / Arkansas Nuclea, One S'(Stll!m Engineer -ANO*! ReactOf' Coolan1 Sv~tem and ANO 1 Spent Fuel Coollna & Purlflcahon Sv~tem System Engineering Buildmg / N-SYE-4 whachoJ 6Pc01c11v com 479-8S8*4337 (offtet!

!(b)(6) !10 1 RCS Leakrate Cycle 25 - - OT Fil Rate (gpm)

Unidentified Leak Rate (gpm) 0.200

- - Total RCS Losses (gpm}

0.180 - - Mean*2 Sigma

- - IAean-2 Sigma 0 160 - l.!ean+3 Sigma

- l.!ean-3 Sigma 0.140 0.120 0.100 0 080 0.060 0.040 0.020 0.000

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From: Tobin Macucet To: Barrett Andy: Tindell Brian

Subject:

FW: AN0-1 RV o-ring leakage

  • 101231201S Power Entry Date: Wednesday. October 28. 201S 7:36:00 AM FYI From: BARBOREK, W DOUG LAS [3]

Sent: Tuesday, October 27, 2015 5:17 PM To: Tobin, Margaret Cc: SHORT, BRADLEY W ; WOODSON, TIMOTHY R; Conyers, Daniel

Subject:

[Externa l_Sender] RE: ANO* l RV o-ring leakage - 10/23/2015 Power Entry Maggie, We have not written a CR to document the fact that the leak rate essentially plateaued . We are, however, monitoring the leak rate on a daily basis. My intent was to initiate a new CR if the leak rate takes a step change in the increasing direction.

Thanks, Doug Barborek Entergy Operations, Inc./ Arkansas Nuclear One System Engmeer ANO*l & AN0-2 Reactor Coolant Systems and AN0-1 Spent Fuel Cooling & Puriflcatoon System System Eng,ncerong Building/ N SYE 4 wbarbol@eoternv com 479-858-4337 (off,ce)

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From: Tobin, Margaret [mailWMari:;aret Tobjn@nrc i:;ovJ Sent: Tuesday, October 27, 2015 3 :58 PM To: BARBOREK, W DOUGLAS Subj ect: RE: AN0 -1 RV a-ring leakage - 10/23/2015 Power Entry Thanks for the explanation Doug. Was there a CR written to document the unexpected behavior?

From: BAR BOREK, w DOUG LAS [maj!toWBARBOl@entergy com]

Sent: Tuesday, October 27, 2015 2:41 PM To: Tobin, Margaret <M argaret.Tobio@orc.gov>

Cc: Tindell, Brian <Brian Iindell@orc gov>; Barrett, Andy <Andy Barrett@nrc gov>; SHORT, BRADLEY w

<bshort@entergy com>; WOODSON, TIMOTHY R <TWOODSO@entergy com>; Conyers, Daniel

<dcoover@eot ergy com>

Subject:

[Externa l_Sender] RE: AN0-1 RV o-ring leakage - 10/23/2015 Power Entry Maggie, We have discussed this internally and briefly w ith AREVA and have no definitive explanation fo r t he leakage to plateau as it has since late July. Our issue has certainly exhibited different behavior than the Indian Point-3 leak and other previous o-ring leaks in the industry. It seems very likely that our o-ring leakage has a different cause than the IP3 issue.

Theo-ring leakage temperature alarm is still locked in on t he Control Room annunciator, and t he leakage is still emanating from the header near t he sump, so we still believe the leakage is from the RV a-ring. The only confirming piece of inform ation we don't have is thermography of t h e leak-off drain piping w hich we cannot obtain at power due to the location of the piping. That w ill be a piece of informat ion we plan t o obt ain if we have a planned or unplanned RX trip which allows general access to the RB basement.

During our next power entry in December, I have informed RP t hat my intent is to take another walk around elevation 354' to look at the basement floor (elevation 335') t hrough the grating and check for signs of any other leakage, as well as checking t he temperature of the RCP int ergasket leak-off lines aga in to ensure they are at ambient temperature and not leaking by. The last entry in which we performed this effort wa s the July entry.

Please let me know if t his d id not answer your q uestions or if you have any additional questions.

Thanks.

Doug Barborek Entergy Operations, Inc./ Arkansas Nuclear One System Engineer- ANO I & ANO 2 Reactor Coolant Systems

and AN0-1 Spent Fuel Cooling & Purification System System Eng,neerong Bulld,ng / N-SYE 4 wbarbol @eotergy com 479-858-4337 (off,ce)

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From: Tobin, Margaret gov

Sent: Tuesday, October 27, 2015 12:48 PM To: BARBOREK, W DOUGLAS Cc: 6ciao,liodell@orc eov; Andy Barrett@orc.eov

Subject:

RE: AN0 -1 RV o-ring leakage - 10/23/2015 Power Entry Doug, I have a follow-on question. I'm curious if you have a theory on why the leak rate seems to have more or less settled out? Everything I heard about this issue from the start was that Op-E from other sites suggested this leak should slowly get worse until you hit a trigger point and needed to shut down.

I'm concerned that somehow we may be missing some important piece of information somewhere because it isn't following the expected physical phenomenon.

Thanks, Maggie From: BAR BOREK, w DOUGLAS [ma j!toWBARBOl@entergv com)

Sent: Tuesday, October 27, 2015 10:33 AM To: Barrett, Andy <Andy Barrett@nrc.gov>

Cc: Tobin, Margaret <Margaret Tobio@orc gov>; Tindell, Brian <Brian I iodell@orc,goy>; SHORT, BRADLEY w

<bshort@entergy.com>; WOODSON, TIMOTHY R <TWOODSO@entergy,com>; Conyers, Daniel

<dconver@entergy com>; EDGELL, DOUGLAS w <PEPGELL@entergy com>

Subject:

[Externa l_Sender) AN0-1 RV o-ring leakage - 10/23/2015 Power Entry Andy, Regarding your questions on t he Unit 1 power entry last Friday.

The boric acid at the ta il pipe is still pristine white, indicating no wa stage of carbon st eel. The leak rate wa s visually typical of previous entries. The rate of increase of the leak slowed down in late July, and is now on a very slow increasing t rend. Below is a snapshot fro m t he Shift Engineer's spreadsheet for RCS leakage which shows the Cycle 26 data. Also attached are some still shots from the video I took. You are more t han welcome to stop by my cube and w at ch t he entire video.

Please let me know if you have any addition quest ions concerning this issue.

Thanks, Doug Barborek Entergy Operations, Inc. / Arkansas Nuclear One System Engineer - ANO-I & AN0 -2 Reactor Coolant Systems and AN0-1 Spent Fuel Cooling & Purification System System Engineering Building / N-SYE-4 wbarbol@eotergy com 479-858-4337 (office)

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From: Tobin Marwei To: OKeefe Neil* Correll Brian; Dixon lobo; Choate 1ackson

Subject:

FW: AN0-1 RV o-ring leakage

  • 1012312015 Power Entry Date: Monday, November 02. 2015 11:02:00 AM Attachments: still shot 1.pd(

FYI, an update for RCS leakage From: BAR BOREK, W DOUGLAS [mai1to:WBARB0l@entergy.com)

Sent: Tuesday, October 27, 2015 10:33 AM To: Barrett, Andy Cc: Tobin, Margaret; Tindell, Bria n ; SHORT, BRADLEY W; WOODSON, TIMOTHY R; Conyers, Daniel ; EDGELL, DOUGLAS W

Subject:

[Externa l_Sender) AN0-1 RV o-ring leakage* 10/23/2015 Power Entry Andy, Regarding your questions on t he Unit 1 power entry last Friday.

The boric acid at the ta il pipe is still pristine white, indicating no wastage of carbon steel. The leak rate was visually typical of previous entries. The rate of increase of the leak slowed down in late July, and is now on a very slow increasing trend. Below is a snapshot from t he Shift Engineer's spreadsheet fo r RCS leakage which shows the Cycle 26 data. Also att ached are some still shot s from the video I t ook. You are more t han welcome to stop by my cube and watch the entire video.

Please let me know if you have any addition questions concerning this issue.

Thanks, Doug Barborek Entergy Operat io ns, Inc. / Arkansas Nuclear One System Engineer - ANO*! & AN0-2 Reactor Coolant Systems and ANO* l Spent Fuel Cooling & Purification System System Engineering Building/ N*SYE* 4 wbarbol@entergy com 479-858-4337 (office)

!(b)(6) l (pager)

RCSLeakrate Cycle 26 0 200

- - CT FIi Rale (OPml Unleltnb~ed Uak!Ute (gpm) 0 180 - - Tot,1IRCS lOSMl (OPIT1) 1--- - - - - - - - - - - - - - - - - - - - - - - --<

- - uean*2S1Qma 0 160 LI ean,2 sioma

- u tan+:l &oma

- u,an-3Sigma 0 140

          • Mean 0 120 0100 0 080 0 060 0 040 0 020 0 000

~

5

10/23/2015 Power Entry - U 1 RV Inner 0 -ring Leakage at HSD-15-2" Drain Header just West of RB Sump.

Fro m: Tobin Margaret To: eamrn Andy

Subject:

FW: FW: AN0-1 RCS l@i:tk R,Ue Projection Note: The attached images appear on Da te: Tuesday, Octobe, 27, 20 1S 8:27:00AM Attachments: ~

subsequent pages within the body of the email jmue007.. D1li.

string.

From : Tindell, Brian Sent: Mond ay, September 21, 2015 8:40 AM To : OKeefe, Neil Cc : Dixon, John ; Correl l, Brian ; Choate, Jackson; Barrett, Andy; Tobin, Ma rgaret Subject : FW: FW: AN0-1 RCS Leak Rate Projection FYI.

From : BARBOREK, w DOUGLAS [maijtpWBARBOj@entergy cam]

Sent: Thurs day, September 17, 2015 5:26 PM To: Tindell, Brian <Brian .Tjndell@nrc gov>

Subject:

[External_Sender) FW: AN0-1 RCS Leak Ra te Project ion Brian, FYI. I was out last week ... this is the first update in two weeks.

Thanks, Doug Barborek fnte<gv Oper.;Hlons, inc / Atkan~s Nuc.lea, One System Engineer - AN0-1 & AN0-2 R.eiM:to, Coolant Systems and AN0* 1 Spent Fuel Coo mg & :Purification S~tem System Engineering Bu ld1ng/ N-SYE-4 itfbarbo]@entergy com 479-858-4337 (office) 1/h\ffi\1 (p.ge,I

~ RBOREK, W DOUG LAS Sent: Thu rsday, September 17, 2015 5: 14 PM To : WOODSON, TIMOTHY R; EDGE LL, DOUGLAS W; SKARTVEDT, MARK EDMUN D; PERKI NS, DAR RELL L; GORDON, ROBERT A; FORD. JEFFREY; Evans, Terry Alan; ANO OPS1 SM 's; Palmer, Charles; BUTLER, PAU L WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, JOHN M; HOWE LL, JERRY W; HILL, STEVEN D; Parker, Bobby Joe; CRAN E, NE LSONS; PACE, ROBERT D; OLIVER, JASON R; MEATH EANY, DANIELJ ; SCH LUTERMAN, PAMELA S; Crosby, Patrick; Beldin. Joshua; GREESON, WILLIAM C

Subject:

RE : AN0 -1 RCS Leak Rate Projection Resend with corrected date.

All, The latest 2nd order po lynomial projection for reaching 0.250 gpm Total RCS Losses is 230 days from now, or May 4th, 201 6. The recent stabilization of the leakage ra te is moving the projected ODM I trigger pomt further into Cycle 26.

Thanks, Doug Barborek Entefgy Oper.-itions., Inc./ Arkanws. Nuclear One System Engineer - AN0-1 & AN0-2 Reactor Coolant Systems and A.N0-1 Spent Fuel Cooling & Purification S~tem svstem El'l8iritNing Bu ld ins/ N SYE 4 wharh2J..tll"'Oferey mm 479 85S--4H7 {office)

!(b )(6) l (p.ge,J

0 300

- -----------~RCS Leakrate Cycle 25

- -OTFII Rate(9prn)

Unldtnl fttd lea~ Ra11 (111>m)

- -TolalRCSLos,os (9p!TI) t.tun*2 Sloma 0 250 - -t.tur>2S19ma

- t.ttan*3 Sloma

- Mean-Js.oma 0 200 *****Mun As of 9/17/2015, the 2nd order polynomlal 0 150 -+-----------i=f...ll~--:...._---------:. Pro h__0__.2::.s;..o~ ---1

__;.:J.;,e=-tt__io.._,n:-:::-:to=:-:re,..a__, .,..

1pm Total RCS leekace Is 230 days from now, or 5/4/2014.

From: BARBOREK, W DOUGLAS Sent: Thursday, September 03, 2015 1:04 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL, JERRY W; HILL, STEVEN D; Parker, Bobby Joe; CRANE, NELSON S; PACE, ROBERT D; OLIVER. JASON R; MEATHEANY, DANIEL) ; SCHLUTERMAN, PAMELA S; Crosby, Patrick; Beldin, Joshua; GREESON, WILLIAM C Subj ect : RE: AN0* 1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 114 days from now, or December 26, 2015.

Thanks, Doug Barborek E.nterg\f Operations, Inc I Atkansas Nucle.a, Ooe Svs1em fn,g1neer-ANO-l & AN0-2 Ae3Ctot Coolant Systems and ANO*l Sptnt Fuel Coo 1na & Pur1fic.1t1on Systtm S.Y$1em Engineering Su ld1ng/ N-S.YE-4 wbMbot@cosecsv oom r,....

479-SSS-.4317 (Gl!it@)

(b)(6) J 1 RCS Leakrate Cycle 25 0 300

- - OTFII Ratt (9pm)

Unldtnllfltd Lut Ralt (Q)m)

--TOla lRCSLosses(gpmJ

- -Llun*2Sl9ma

- - Mtan*2 Slgmt

- uun--3SJoma 0 200 - -Mean,J&grru

*MUii

- - POIY (Total RCS LOI- (gpm))

0150 ,--------------1<-:-r;l,,..dl:::.________.L.:,:::,,:c=,:.:=.-==:.=::-=.::=:._---1 cpm Total RCS leaka&e Is 114 days from now, or 12/26/2015

From: BARBOREK, W DOUGLAS Sent: Thursday, August 27, 2015 5:46 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JE FFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, JOHN M; HOWELL, JERRY W; HILL, STEVEN D; Parker, Bobby Joe; CRAN E, NELSON S; PACE, ROB ERT D; OLIVER. JASON R Subj ect : RE: AN0-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 104 days from now, or December 9, 2015. This proj ection is beyond t he last four weeks of projections due to a slight leveling off of the Total RCS leak rate over the last week to 10 days.

Thanks, Doug Barborek Entt'iV OJ)t1o1t1on~. lr'IC / AJka,n~> Nut lt.lt Ont SV'!,\cm Engineer - ANO 1 & ANO 2 Rc.Ktot Coolant Sl/l,tcm~

and AN0-1 Spent r uel Cooling & P1.ir1hcatlon System S~tcm(ro~lr.t~dng Bu ld1n9/ N SYl-4 wbMbol@rnrew oom 479-858'-4H7 (offlct)

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-rnr~ - - - - - - - - - - - - - ~

- - OTFI Ralo(llpm)

Unldenbnod Leak Ralt ( gi,m)

- - Total RCS Losses (gpm)

- - Moan*2Slgma

- - Mun-2!Mgma

- - Moan*3Slgma 0200

- Mean-3SIOJ"l"la

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  • ec.:ct7i'o-::
'n::--i t o reach 0.2501pm Total RCS Leaka1e is 104 days from now, or 12/9/2014.

0 000 .,...................................................................._ .............. _........................................................_ ........................................

From: BARBOREK, W DOUGLAS Sent: Thursday, August 20, 2015 10:41 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD,JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS, J OHN M; HOWELL. JERRY W; HILL, STEVEN D; Parker, Bobby J oe; CRANE, NELSONS; Pace, Robert D. (INPO) (paceRP@INPO or~)

Subj ect: RE: AN0-1 RCS Leak Rate Proj ection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 94 days from now, or November 22nd, 2015. This project ion is about two weeks beyond the last three weeks of projections due to a slight leveling off of the Total RCS leak rat e over t he last week.

Thanks, Doug Barborek Entc,sv Optr4t10n), inc I Ark111nw~ Nut lf ,H o~

System En,g1neer - AN0-1 & AN0-2 Reac.tor Coola nt Systems and AN0-1 Spent r uel Coo ,ng & Pu,1f1Catlon System System f.ngln-eering Bu Id 1ng / N-SYE. -4 wharbol@POIPCID'-OOID 479*853 4H? (ofllc*)

!(b)(6) l (p.ge<)

RCS Leakrate Cycle 25

- - OTF-Ralt(Op,11) un~n11t1olukRa11 t - J

- - TotalRCSlouu(gpm)

- - wean*2'Sio,na lolt.,..2$io,na

- Ytan*l~gma 0 200 - 1o11.,..3$i0,na

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polynomlal projection to reach 0.250 Total RCS Leakce Is 94 days from now, or ll/22/ 201S From: BARBOREK, w DOUG LAS Sent: Thursday, August 13, 2015 3:15 PM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUN D; PE RKINS, DARRELL L; GORDON, ROBERT A; FORD.J EFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G; GARBE, CHARLES R Cc: MEYERS. J OHN M; HOWELL.JERRY W; HILL, STEVEN D; Parker, Bobby J oe; CRANE, NELSONS; Pace, Robert D. (JNPO) (PareRD@)NPO org)

Subj ect: RE: ANO-1 RCS Leak Rate Projection All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 85 days from now, or November 6th, 20 15. This is consistent with the last two weeks of projections.

Thanks, Doug Barborek

[r\lCtQ'f 0J)trj llOr'15, ln,c / Afk6r'l1.;l5 Nutlt,H 01\C S~tem £nglf'}Ccr - ANO 1 Re.actor C<lolant 5y5tcm and A'II0 -1 S1>tnt Fuel Cooltn8 & P1.mhcatlon 5vstem Svsttm (nglnttdng 8u ld1n9 / N*SY(-4 wti~rbol@enrergy com 479*8S8 4U7 (offl<<")

l<b)(6) r**.. 1 RCS Leakrete Cycle 26 0 JOO

--OTFII Rate (gpn,)

Undtn*neo Ltak Race (Ql>"11

- - Totot FICS losstS(Ol)n,)

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From: BARBOREK, W DOUG LAS

Sent: Thursday, August 06, 201 5 11 :48 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUND; PERKINS, DARRELL L; GORDON, ROBERT A; FORD, JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERNER, GEORGE W; HOLLOWOA, TROY B; Davis, Barry; PUTNAM, REX G Cc: MEYERS. J OHN M; HOWELL.JERRY W; HILL. STEVEN D Subj ect: RE: AN0-1 RCS Leak Rate Projection

All, The latest 2nd order polynomial projection for reaching 0.250 gpm Tota l RCS Losses is 91 days from now, or November 5th, 2015 (ref. attached ODMI, Rev. 01). This is based on leak rate data from 4/ 1/ 2015 t o present, which correspo nds to t he approximate date that the total RCS leakrate began increasing. See projection graph below.

Note that the nominal inner o-ring leak rate is determined by adding the increase in t he RB Su mp f ill rate since -4/1/2015 to the increa se in the T-111/T*42 fill rat e since ~6/22/2015. The Increase in the Unidentified leak rate is NOT indicat ive of t he total o ring leakage rate since some of the o ring leakage is apparently being condensed in the RCP Seal CollectiOn system loop seal and is being returned to the Quench Tank as Identified leakage.

Thanks, Doug Barborek Ente<gy Optr.ttions, lne / A/kansas Nuclea, Ooe S~t~m Eoglnte,r-ANO l & ANO 2 RektOt Coolaol SV'\tC'mS and ANO 1 Spent Fuel (001mg & Pu<1f1C,1tlon System S~1em Et'IBl~ering euId 1nA / N-SYf-4 Yd:iatboJ@emer,v com 479-858-433? (office) lifilfil](p.1gH\

RCS Leakrate Cycle 25

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- - OT Fil Rate (gpm)

Unldentifled Leak Rate (gpm)

- -Total RCS Losses(gpm)

- - Mean*2 Sigma

- - Mean-2 Sigma

- Mean*3Slgma 0.200 - Mean-3 Sigma

Mean

--Poly. (Total RCS Losses (gpm)) As of 8/ 6/ 2015, t he 2nd order polynomial projection to 0 150 +--- - - - - - - - - - - - - - - - - t , - : : ; f ' ~ - - - " - - - - - - - - - ' :r"'e* .:".::-:c,:-;;

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From: BARBOREK. W DOUGLAS Sent: Friday, July 31, 2015 10:24 AM To: WOODSON, TIMOTHY R; EDGELL, DOUGLAS W; SKARTVEDT, MARK EDMUN D; PERKINS, DARRELL L; GORDON, ROBERT A; FORD.JEFFREY; Evans, Terry Alan; ANO OPS1 SM's; Palmer, Charles; BUTLER, PAUL WAYNE; WOERN ER, GEORGE W; HOLLOWOA, TROY B Cc: MEYERS. J OHN M; HOWELL.JERRY W; HILL, STEVEN D Subj ect: AN0-1 RCS Leak Rate Projection

All, Obviously, this is not an exact science. but the latest 2nd order polynomial projection for reaching 0.250 gpm Tot al RCS Losses is 99 days from now, or November 7, 2015 (ref. attached OOM I, Rev. 01). This is based on leak rate data from 4/1/2015 to present, which corresponds to t he ap proximate date that the total RCS leakrate began increasing. See projection graph below.

FYI, the same projection two weeks ago calculated a date of 12/19/2015 to reach 0.250 gpm. It is ant icipated that this va lue will cont inue to move towards the present.

I will be participating in a 2nd Power Entry on 8/4/2015 to quantify the leakage at the drain header near the west end of the RB Sump.

I will attempt to update th is projection on a weekly basis from this point forward.

Thanks.

Doug Barborek Entergy 0~1ations, loc / Alkans.as Nucl!ar Ont Svs1em Engll'\ter- ANO, t 8, AN0*2 Reactor Coolant Svs1ems

and AN0-1 Spent Fuel Cooling & Purific ation S~tem S~tem Engineering Building / N-SYE-4 wNrhoJ@tnrrrev mm l<b )(6) I 479*853~337 (offia!)

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RCS Leakrate Cycle 25 0.300 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~

- - OT Fil Rate (gpm)

- - Unidentified Leak Rate ( gpm)

- - Total RCS Losses (gpm)

- - Mean+2 Sigma

- - Mean-2 Sigma

- Mean+3Sigma 0.200 - Mean-3 Sigma

l~ean As of 7/30/2015, the 2nd

- - Pol)'. (Total RCS Losses (gpm)) order po lynomial 0 _1 50 +-- - - - - - - - - - - - - - - - - - - --tr- >-./<"=---- - - - - - - -..!p'--'r-=- o'-'

je:.:c.::t:i.:oc.cn:...t:.:o:...r:...:e:.:a:.:ccch_:0:...2

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gpm Tota l RCS leakage Is 99 days from now, or 11/7/2015.

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From: Tobin Margaret To: Tindell Brian Note: The same two procedures, the first with a cover

Subject:

FW: fW: OOM I Update

  • Reactor Vessel Head Inner Gasket Leak sheet, IEN-OP-111 Rev.11 and EN-FAP-OM-021 Rev.1, Date: Thursday, July 16, 2015 9:38:00 AM Attachments: RY Flange 0-riog l eakave odf are withheld in their entirety under FOIA exemption 4.

imageOOJ nag From: SHORT, BRADLEY W [4]

Sent: Thursday, July 16, 2015 8:55 AM To: Tobin, Margaret

Subject:

[External_Sender] RE: FW: O DMI Update - Reactor Vessel Head Inner Gasket Leak It has been signed off. See attached BRAD SHORT LICENSING REGULATORYSUPPORT PHONE 479-$5$*32r CELLl(b)(6) _

From: Tobin, Margaret [majlro-Mariaret Jobjn@nrc KPYJ Sent: Thursday.July 16, 2015 7:49 AM To: SHORT, BRADLEY W

Subject:

RE: FW: ODMI Update - Reactor Vessel Head Inner Gasket Leak

Brad, Did this ODMI ever get signed off?
Thanks, Maggie From: PYLE, STEPHENIE L [majlto*SPYLE@entergy com]

Sent: Wednesday, July 08, 201510:04 AM To: Tobin, Margaret Cc: SHORT, BRADLEY W

Subject:

[External_Sender] RE: FW: 0 DMI Update - Reactor Vessel Head Inner Gasket Leak The ODM I should be signed off by COB Thursday and we can get you a copy then.

Thanks, Stephenie From: Tobin, Margaret rmailto*Mamret Jobjn@nrc KPYl Sent: Tuesday, July 07, 2015 3:56 PM To: PYLE, STEPHENIE L

Subject:

RE: FW: ODMI Update - Reactor Vessel Head Inner Gasket Leak Hi Stephanie, Thanks for the update on this. Do you know if the full ODMI is available?

Thanks, Maggie From: PYLE, STEPHENIE L [majlto*SPYLE@entergy com]

Sent: Tu esday, July 07, 2015 3:48 PM To: OKeefe, Neil; Tindell, Brian; Taylor, Nick; Tobin, Margaret

Subject:

[Externa l_Sender] FW: ODMI Update - Reactor Vessel Head Inner Gasket Leak FYI From: ANO MAIL Sent: Tuesday, July 07, 2015 3:38 PM To: Server ANO-EXS01; Server ANO-EXS02

Subject:

ODMI Update - Reactor Vessel Head I nner Gasket Leak

From: Taylor Nick To: Burritt Arthur Cc: O"Keefe Neil: Tindell Brian: Correll Brian: Egli Richard: Tobin Margaret* Dixon John

Subject:

FW: Unit 1 RCS Lea kage ODMI Date: Tuesday, July 21, 20 15 11:30:11 AM Attachments: RV Flange 0-ring Lea kage.pdf Note: The same two procedures ( EN-OP-imageoo1 png 111 Rev.11 and EN-FAP-OM-021 Rev.1 )

- - - - - - - - - - - - - - - - - - . are withheld in their entirerty under FO IA ex

Art, You were looking for OE last week on operations with both inner and outer O rings failed. I hadn 't discovered any previous occurrences. But we got ANO's ODMI today and the licensee lists a number of previous occurrences both at ANO, Browns Ferry 1 and Wolf Creek.

Hope this helps, Nick Taylor Senior Project Engineer Division of Reactor Projects USNRC Region IV 0: 817 200-1520 C: (b)(6)

E: oick.taylor@nrc.gov From: Correll, Brian Sent: Tuesday, July 21, 20 15 9 :11 AM To: OKeefe, Neil; Taylor, Nick; Farina, Thomas; Tinde ll , Brian ; Tobin, Margaret Cc: Egli, Richard

Subject:

Unit 1 RCS Leakage ODMI Unit 1 RCS Leakage ODMI is attached. Pages 5 and 6 are the Thresholds and Actions to be taken if the thresholds are exceeded.

Brian

From: Egli, Richard Sent: Tuesday, July 21, 2015 12:42 PM To: Taylor, Nick; Burritt, Arthur Cc: O'Keefe, Neil; Tindell, Brian; Correll, Brian; Tobin, Margaret; Dixon, John

Subject:

RE: Unit 1 RCS Leakage ODMI Sizewell B Operation with both 0-rings leaking:

http://nrr 10.n re.gov/rorp/airs/00003428.html From: Taylor, Nick Sent: Tuesday, July 21, 2015 11:30 AM To: Burritt, Arthur Cc: OKeefe, Neil; Tindell, Brian; Correll, Brian; Egli, Richard; Tobin, Margaret; Dixon, John

Subject:

FW: Unit 1 RCS Leakage ODMI

Art, You were looking for OE last week on operations with both inner and outer O rings failed. I hadn't discovered any previous occurrences. But we got ANO's ODMI today and the licensee lists a number of previous occurrences both at ANO, Browns Ferry 1 and Wolf Creek.

Hope this helps, Nick Taylor Senior Project Engineer Division of Reactor Projects USNRC Region IV 0: 817 200-1520 C: (b)(6)

E: nick.taylor@nrc.gov From: Correll, Brian Sent: Tuesday, July 21, 2015 9:11 AM To: OKeefe, Neil; Taylor, Nick; Farina, Thomas; Tindell, Brian; Tobin, Margaret Cc: Egli, Richard

Subject:

Unit 1 RCS Leakage ODMI Unit 1 RCS Leakage ODMI is attached. Pages 5 and 6 are the Thresholds and Actions to be taken if the thresholds are exceeded.

Brian

N RC: International Incident Reporting System Page 1 of 3 FOR NRC INTER..l\IAL USE ONLY Not Ror Pub Iic Di~lribuiion

  • international. Incide nt Reporting System (IRS)

- IRS Number- 0 00764 3

- Rep0rt Type- Mai n

-IAEA NUmbar-

- NEA Number-

-Date of Receipt- 20040916

  • Title* - CONTINUED REACTOR OPERATION WITH BOTH REACTOR PRESSURE VESSEL HEAD ' 0 ' RING SEALS LEAKING DUE TO LEAK-DETECTION SYSTEM DESIGN DEFICIENCI

-Country- Uni t ed Ki ngd om

-Date of Incident- 20010511

-Planl_N8ma- Sl ZEWELL*B

- Power- 1196

  • Plant Code- GB-24
  • PRIS_Pl*nt_Code- GB-24

-De s i g ne r - PPC

-Reactor Type- PWR

-SLart_of_operation- 19950922

-Abstract-Whilst starting up after the refuell ing outage the then11ocouple in the Reactor Pressure Vessel (RPV) head seal leak detection system indicated leakage from the inner head o-ring seal. The inner seal lenknge detection path was isolated as per operating instructions. Later in the fuel cycle airborne activi ty levels, humidity and sump levels provided evidence of leakage from the Reactor Coolant System (RCS), but this always remained well within the Technical Specification limit for unidentified leakage.

Several containment entries were made at powel', followed by an entry during a forced outage. to search for the source of the leak but the leak site was not detected. Based upon the lack of evidence from the leak searches in the other areas, and the fact that the thermocouple on the outer seal leak detection system showed no sign of leakage, an eva luation of possible leak sites determined the most likely to be a 'conoseal' thermocouple connection. The reactor was returned to power afler the forced outage with two areas uninspected; the reactor head package and the instrument tunnel beneath the reactor.

Preparations were made for a planned shutdown to fut1her progress identification a nd resolution of the problem) but indication ofan increasing instrument tunnel sump level in the reactor building gave cause for concern and plans for the shutdown were brought forward. The reactor was shutdown on 11 May 200 I to enable personnel to conduct a deta iled leak search in the reactor building.

The leak source was identified as a leak from the RJ'V Head outer o-ring seal. The subsequent investigation revealed that the outer o-ring leakage detection system would not reliably detect outer o-ring leakage w hen the reactor is at operating cond itions.

-Guide Words-C ODED WATCH LIST O F GUIDE WORDS A_ I REPORTING CATEGORIES A_ 1_3 Deficiencies in design, construction, operntion (including maintenance and surveillance), quality ossumncc or sofcty evaluation A_2 PLANT STATUS PRIOR TO Tl-IE EVENT A_2_ 1 On powe*r A_3 FAILED/AffECTED SYSTEMS i\_3_/\C Reactor vessel (with core internuls. PHWR or LWGR pressure tubes,...)

A 4 FAILED/AFFECTED COMPONENTS A_4_2_0 Other A_5 CAUSE OF THE EVENT i\_5_1_1_6 Leak A 5 I I 8 Blockage, restriction, obstruction, bi nding, foreign material A 6 EFFECTS ON OPERATION A_ 6_ 2 Controlled shutdown A_7 CHARACTERISTICS OF THE INCIDENT A_7_ 2 Degraded reactor coolant boundary A 8 NATURE Of FA ILURE OR ERROR A_8_2 Multiple failure or multiple error A_9 NATURE OF RECOVERY ACTIONS A_9_ I Recovery by human action

-Full_Rcport-

2. NARRATIV E DESCRIPTION Introduction http://nrrlO.nrc.gov/rorp/airs/00003428.html 01/10/2018

NRC: International Incident Reporting System Page 2 of 3 The RPV flange is fitted with two leak off lines 10 detect inner and outer o-ring seal failure (Attachment I). The two lines from the vessel flange common together before passing through a remote controlled isolution valve and non-return valve to the Reuctor Coolant Druin Tank (RCDT) (Attuchment 2).

The arrangemcnl is such that if the inner seal fails, the inter-space between the inner and outer seal pressurises and hot water/steam is forced down the inner leak detection line causing the temperoture 10 rise at a temperature probe. This initiates an alarm in the Main Control Room.

With the inter-space between the o-ring seals pressurised. the full reactor pressure is taken by the outer seal. The expectation was that should the outer seal fail the outer seal leak detection line would pick up the leakage of hot water flowing past the temperature probe and re-initiate the alarm.

The Ileliconex o-ring seals consist of an inner lnconel spr*ing surrounded by a C-shapcd lnconel inner jacket 10 spread the spring load. This in tum is covered with a similar C-shaped outer jacket made of silver to provide the confonnity with the seal face roughness. The seals arc installed with the opening of the C-shaped jacket facing away from the pressurised water.

Event Description During preparations on 23 October 2000 for reactor return to power at the end of a refuelling outage, evidence for failure of the Reactor Pressure Vessel (RPV) inner o-ring seal was observed. The inner o-ring seal, together with the outer o-ring seal, constitute the barriers to reactor coolant escape to the containment building from the interface between the RJ>V itself and the RPV head. Both seal arrangements i11eludc leakage detection equipment nnd arc subject to station procedures for the action to be taken on receipt of n seal leakage detection alarm.

As station Technical Specifications J>Cnnit operation with the inner seal in a failed condition, the decision was taken 10 isolate the inner o-ring se.11 lcak detection system as per the relevant procedure and continue with reactor start-up. The outer o-ring seal leakage de1ection system remained in service.

In mid-December 2000, evidence for water leakage into the reactor building was observed, but the source of the leakage was unknown. Technical Specifications permit operat ion with both identified and unidentified leakage, but the allowable leakage rates are smaller for unidentified leakage. Although the plant remained within the Technical Specification limits for unidentified lei,kage. severnl containment entries were made (both at power :md during an unplanned shutdown) during December 2000 and early January 200 I in an unsuccessful effort to identify the source of the leak. The diagnostic analysis of water samples, increased containment bui lding humidity levels, airborne activ~ty levels and the presence of boric acid powder on some surfaces all indicated that 1he leakage was reactor coolant. but the source of the leakage remained unknown. The area beneath the RPV and the area around the RP\/ head were not inspected for radiological dose reasons, as the outer o-ring seal leakage detec1ion system continued 10 give no indication of a leak in this area.

Following u further, unsuccessful leak search at power. arrangemcnls were made 10 shutdown the reactor on 18 June 200 I to identify the source of the leak. However, over the weekend of 5 Mny 2001. indications of an increasing sump level in the instrument tu nnel at the bottom of the reactor building gave further cause for concern. The reactor coolant leakage rate was being detennined through the level indication in the sump; once the sump had filled, level indication would have been lost and it would have become impossible to quantify the leakage rate. In response, the decision was taken to carry out a controlled reactor shutdown on I I May 200 I.

The resulting dcrnilcd leak search revealed that the rcnctor coolarit lcak source wus from the RJ'V outer o-ring seal (shown later to be on the opposite side of the RPV head to the inner o-ring sc:il leak site). Both the inner and outer o-ring seal failures occurred at the interface between the seals and the upward facing flange, with the most likely initiating eve11t being debris present in the seal and RPV/hcad flange joint area from the recent refuell ing outage.

The leak site on the vessel nange is illustrated in Attachment 3. The upper internals are visible in the top left of the picture, whilst on the right can be seen part ofa head stud.

The leak site on the closure head is illustrated in Attachment 4. The head has been partially raised for initial inspection with the RPV to the lower left. The inner aod outer o-ring seals can seen in this view.

Theo-ring seals were visually inspected on removal and the secl io n of the inner and outer seals that contained the leak sites were retained. An additional section of seal that exhibited an unusual deformation was also retained (sec Attachments 5 & 6).

3. SAFETY ASSESSMENT Station Technical Specifications permit operation with the inner seal in a failed condition and also pem1it operation with both identified and unidentified leakage, although the allowable leakage rates are smaller for unidentified leakage.

The operating procedure for action on receipt of the inner seal leak detect ion alarm instructs the operator to confirm the leakage by mon itoring the RCDT or by using a c-0ntac1 thcrmorncter and then isolate the irrner seal leak off hoc aod establish monitori11g 011 the outer seal.

4. CAUSE ANALYSIS Design Intent Although the design intent was that should the outer seal fail, the outer seal leak detection line would pick up the leakage of hot water, investigations subsequent to the event indicated that this would not happen in practice when the reactor was at operating temperature. The RJ>V flange is designed wi th a I mm gap 011 the outside of the outer seal, which rneans there is a direct path from the outer seal to the containment environment. There is a single leak detection hole in the flange, which in this case was on the opposite side of the vessel 10 the leak site. When water (at 155 bar and 323oC) leaks 10 the containment atmosphere, approximately 40% will turn 10 steam straight away forming a two-phase steam/water jct. Any water that collects in the Imm wide gap will rapidly boil due to the heat from the hot (323oC) metal vessel. The boiling will cause boric acid to come out of solution forming crystals in the gap blocking the path around the circumference of the vessel.

Thus the leak wi.11 form a water/steam jct that takes the path of least resistance straight out of the gap in the vessel nange to the containment bui lding. It is therefore very unlikely that a significant quantity of liquid would travel around the flange via the inner/outer seal inter-space to the leakage detection hole un less the leak site is immediately adjacent.

This conclusion is supported by a physicul check undertaken on the seal leakugc detection system curried out whc11 the rcuctor was first shut down and before it was deprcssuriscd. By opening u manual drain valve it was confinned that only 5 drips per minute were collecting in the outer seal leak detection line. This could have been due to condensation of the steamy atmosphere that must have been prescm in the narrow gap in the outer flange. Given the length of the small bore pipework from the flange to the temperature probe. such a small amoun1 ofliquid would not cause any significant te111pe1111ure rise.

The Nuclear Steam Supply System al Sizewell B is of a Westinghouse design but the RPV and head were manufactured in France. A comparison of other vessels to the Sizewell B vessel shows some differences in the flange detail. Vessel heads e ither have a totally flat flange face with metal-to-metal contact on the outside of the outer seal (which forces the water/steam around the outer seal grove to the leak detection hole), or have a taper across the outer part of the nange with the gap being designed to close locally just outside the outer 0-ring groove during bolting of the head to the vessel. This gap then closes completely during the heat up transient due to rotation of the closure flanges, and re-opens as the temperature stabii ises and steady state is attained. However the gap on the outside oflhe Sizcwell B outer seal is greater and, therefore. docs not fully close at operating condit ions.

The conclusion is that the RPV head seal leak detection system installed at Sizewell B will provide an effective warning for an inner seal failure, but the presence of the gap results in a reduced reliabi lity in detecting the failure of the outer seal. Therefore, ifo:pernting on the outer seal, altc111atc indications such as containment activity, hum idity and drainage must also be used to detect a possible outer seal fai lure.

Leak Searches Three co11111in111cn1entries were made at power to try and identify the source of the leak but were unsuccessful. A ,*cactor trip then occurred and the opportunity was taken to carry out an inspection concentrating on the inner annulus (not accessible at power) but the source of the leak was not found. However, there was evidence of boric acid crystals and surfaces coated with boron powder within the reactor building. The reaclQI' was rcsta11cd the following day.

http://nrrlO.nrc.gov/rorp/airs/00003428.html 01/10/2018

N RC: International Incident Reporting System Page 3 of 3 In retrospect, no inspection of the RPV head package took place and no entry was made into 1he instrument tunnel to inspect the under side of the vessel (these areas only being accessible with the reactor shutdown). Even though the leak was small and may not have been readily seen, finding the source of the leak would huvc allowed more effective management of the issue, particulurly those associated with the corrosion of carbon steel due to the boric acid. Sizewell B did not have a boric acid leakage detection programme (as required in the US as a result of NRC Generic Letter 88-05) which should have prompted an inspection of the head pac kage and evaluat ion of the consequences of the leak before restarting the reactor and the conscqu.ences of the leak. The forced shutdown was a missed oppo11unity to identify the source of the leak.

Sizewcll B was built after 1988 and therefore, the site Operating Experience group did not review the NRC Generic Lener. There has been no requirement to review historical OE information, therefore the opportunity to learn from international operating experience was lost.

Foreign Material Inclusion (FME)

The most probable cuusc of seal fai lure was debris on the vessel n(lnge. Investigation has revealed that (lebris could have been deposited on to the nangc from two causes, both of which urc clearly identified in Westinghouse Technical Bulletin (NSD-TB-87-02). This document specifically explains the importance of protecting the opening between the reactor vessel head and reactor vessel nangc during hydrolasing and other refuell ing cavity cleaning operations. The control of water level in the RPV during setting and bolting the head 10 prevent washing small particles of debris 01110 the vessel nangc is also covered in the document.

When the closure head is initially placed onto the vessel after refuelling the seal faces do not make contact and there is a gap between the vessel nange and o-rings in the closure head. To protect the clean vessel head and prevent debris entering the flange area, a cover is placed over the head package extending to the cavity floor. This is important as the process of mechanically scrubbing and jct washing the walls and floors of the cavity can cause a lot mobile debris and paniculate matter. It has been identified that during the previous refuelling outage the protective cover was 1101 correctly fitted and had been left sat on the top of the closure head bolting ring. It therefore did not provide any protection against debris being washed into the flange gap.

Written work instn,ctions for the installation of the head cover did not provide comprehensive instructions for fitting of the cover, nor did they require a fina l independent inspection to confirm it was fitted correctly. The instructions also did no t require the use o f a secondary FME barrier (or debris dam providing additional protection to the flange joint), such as a soft deformable strip that can be pushed into the nangc gap before the plastic sheath is dropped on to the floor.

Westinghouse Technical Bulletin (NSD-TB-87-02) recommends 1ha1 plants with 'inverted top hat' support plate upper internals should maintain the water level approximately 361040 inches below the RPV nange during selling the head and until the head is tensioned. The reason for the lower level is that experience has shown that with higher water levels there is a risk of inadve11e11tly flooding of the vessel flange, particularly when the heud is placed on to the vessel. The problem is that when the h.ead is not tensioned it does not sit down on the flange joint, but rests 'proud' on the reactor internals. This means the o-ring seal are not in contact with the flange foce and there is risk that debris or boric acid crystals may be lef\ on the faces of the f'lange, which could damage the o~ring seals during tensioning.

Sizewell B had followed the water level advice since the first refuelling outage, maintaining the level at 965mm below the nange. However due to operat ional concerns during the last refuelling outage the station operating procedures were temporarily amended to allow the water level to be maintained between 200 mm and 965 mm below the flunge. Duling the refuelling outage the head was initially placed on the vessel with the water level nominally 965mm below flange. To maximise inventory the level was then raised and maintained between 24 5-360 mm below flange during the 40-hour period were the cavity was being de-contaminated and the RPV stud holes cleaned. The level was then lowered back to 965 mm for the fina l head tensioning. The head was lifted and replaced for the final nange inspection before tightening whilst the water was still at the higher level.

Another possible source of debris on the flange was identified during the cull'cnt forced outage to replace the o-rings. II was found that water was inte1111i1tently being splashed on to the flange the when RCS level was operating at a nominal 150mm below the flange. Due to the turbulent flow below the upper internals water was being forced up the upper head recirculation flow holes where it nowed across the flange face. This was managed by reducing the level to 350mm below the flange. Attachment 7 shows the water coming from the centre of the three holes, onto the upper intenrals and then onto the flange face.

5. LESSONS LEARNED AN"D CORRECTIVE ACTIONS Review o f NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Vessel Boundary Components in PWR Plants and robust implementation of its recommendations.

Confidence in the provision by reactor pressure vessel head seal detection systems of early warning of seal failure.

Operation with a failed i11ocr RPV o-ri11g. The following action statement has been included in the relevant Sizewcll B operating procedures:-

lf the RPV is operating on the outer 0-ring seal the conlui1m1cnt condit ions must be monitored carefully. If any RCS leakage is identified urgent act ion must be taken to identify the source of the leak. If the leak cannot be identified by physical containment inspections within 15 days the reactor should be shut down to allow a proper inspection of RPV flange, Conoseals and other inaccessible regions. The reactor should then only be operated once a formally agreed operating regime has been established.

The 15 day action period is based on the rules that EdF apply to their larger 4 loop PWRs.

Genera l Unidentified RCS Leakage: If there is physical evidence ofan RCS leak in conta inmenl urgent action must be taken at Sizewell 8 to determine the location and size of the leak. A comprehensive assessment should be made to evaluate the impact and risk of operating with the leak and clear gui dance must be provided as to what additional monitoring must mke place and DI what levels spec ific actions shall take place. If the leak source cannot be identified the reactor should be shut down in a timely manner to allow proper inspection of the RPV head and other inaccessible area in the inner annulus.

Implement a fonnal Boric Acid leak Detection Procedure: Sizewell Bare developing a fom,al inspection procedure for investigating and evaluating the effects of IRCS leakage in line with US NRC Generic Letter 88-05.

Improve the routine procedures for reactor re-assembly: A number of improvements were implemented on the work order cards and procedures for re-assembling the head during the forced outage.

These arc now being embedded in the routine refuell ing procedures.

lm1>rove RPV head seul leak detection system: Sizcwell B ,,re reviewing international procedures and methods in this flrea.

Controlling RCS water level w hen refitting the reactor vessel head: Sizewell B operating procedures have been amended to ensure the water level is maintained at the 965111111 below the flange at all stages from when the head is re-seated on to the vessel until it is finally tensioned.

Technical Specification for Reactor Coolant Leakage: The bases of the Sizcwell 13 Technical Specification for Reactor Coolant Leakage have been improved to explain how the assessment of pressure boundaiy leakage is made.

FME Precautions: The proc~dure for fining the FME protective sl1eath has been improved to give clear instructio,, as to how it should be fined and an inspection hold point inserted before c<rvity cleaning comme nces. It is recommended that an additional FME b arrier be insta lled in the RPV head to flange gap during stud hole cleaning and to provide secondary protection during cavity clean.

http://nrrlO.nrc.gov/rorp/airs/00003428.html 01/10/2018

From: Taylor, Nick Sent: Thursday, July 16, 2015 12:29 PM To: Burritt, Art hur Cc: O'Keefe, Neil; Tindell, Brian

Subject:

FW: Indian Point Head Leak OpE

Arthur, FYI, ANO unit 1 is running right now with a leaking inner seal, and the leakoff line between the seals cannot be isolated. Their leak is manifesting itself as a slow water leak into the sump (about 0.07 gpm and rising very slowly for now), and for some historical reasons they cannot close the isolation valve in the annulus drain line. Licensee is struggling with establishing with ODMI criteria. Not sure if this helps inform your situation at Indian Point or not. ..

We have had some RIV PWRs run the cycle with a failed inner seal, but never with both seals failed to my knowledge.

Hope this helps, Nick Taylor Senior Project Engineer Division of Reactor Projects USNRC Region IV 0: (817) ?QQ-1520 C: !(b)(6) I E: nick.taylor@nrc.gov From: Pa nnier, Stephen Se nt: Thursday, July 16, 2015 10:41 AM To: Taylor, Nick

Subject:

FW: Indian Point Head Leak OpE FYI From: Burritt, Arthur Se nt: Thursday, July 16, 2015 11:32 AM To: Pannier, Stephen

Subject:

RE: India n Point Head Leak OpE Thanks

From: Pannier, Stephen Sent: Thursday, July 16, 2015 11:28 AM To: Burritt, Arthur Cc: Pickett, Douglas

Subject:

RE: Indian Point Head Leak OpE There was also a question about a RIV unit currently operating with a failed inner seal. That is ANO Unit

1. But the outer seal is holding, so this is different than what is occurring at IP. Nevertheless, the POC in the Region for ANO for any questions is Nick Taylor.

Thanks Steve From: Pannier, Stephen Sent: Thursday, July 16, 2015 10:17 AM To: Burritt, Arthur Cc: Pickett, Douglas

Subject:

RE: Indian Point Head Leak OpE Oh ... sorry about that hyperlink issue.

This link should work https:/ / adamsxt. n rc.gov/WorkplaceXT/getContent?objectStoreName=Main . .Library& vs Id=% 78728 BC F23-D94C-402 D-AD D5-6AOOOE788 3E7% 7D&id=%7B8FD9 F7BB-B F 62-493C-B569-2869B 181 C026% 7D&objectType=document

Thanks, Steve From: Burritt, Arthur Sent: Thursday, July 16, 2015 10:03 AM To: Pannier, Stephen

Subject:

RE: Indian Point Head Leak OpE Thanks, but the first SER involves a BWR so not really applicable and the second link went to a part 21 issue at Vogtle From: Pannier, Stephen Sent: Thursday, July 16, 2015 9:53 AM To: Burritt, Arthur; Pickett, Douglas

Subject:

Indian Point Head Leak OpE Hi Folks, I found a few events related to operation after failure of both reactor vessel head seals. As you can see there isn't much available from the NRC, as these events are not typically reportable under 50.72 and / or 50.73. What I found is included in INPO documents ... through ICES and INPO reports.

Here you go .....

2

SER 3-09, "Unrecognized Reactor Pressure Vessel Head Flange Leak." This is Browns Ferry Unit 1 reactor vessel head leak. Also discussed in this report is a reactor vessel head leak which occurred at Dresden. In December 2001, Dresden Station Unit 3 was removed from power operation to allow personnel to identify and repair a leak in the drywell. A reactor vessel pressure test identified a reactor head/flange leak. Inspections of the vessel and head flange sealing surfaces identified multiple areas that had experienced steam cutting . The reactor head 0-rings had not been adequately compressed at these locations. The depth of the seating grooves was deeper than the dimensions specified by the vendor as-built drawing.

The INPO report related to the Dresden report is attached to this message.

INPO SER 5-03, "Operational Decision-Making.' (Event summary starts on page 23 - The unit is Sizewell B NPP in the UK. Sizewell B is a 1, 188-MWe, Westinghouse pressurized water reactor with two turbine generator sets that began commercial operation in 1995).

Thanks, Steve Pannier Note: The two INPO documents, consisting of 27 IOEB pages, have been withheld in their entirety under FOIA exemption 4.

3

Official Use Only Sensitive Internal Information Internal NRG Qnly RIV NEWS Thursday, 06/14/2012 Sup ort Issues RA NSTR Plant Status BRANCH A Non Responsive Record STP Unit 2 100% Experienced three temperature spikes over the last 24 hrs between the two o-rings in the vessel head (temp currently normal). Has occurred before - licensee not Non Responsive Record BRANCH B Non Responsive Record BRANCH C Non Responsive Record BRANCH D INon Responsive Record Official Use Only Sensitive Internal Information lntenial r~RC Only

From: Reinert, Dustin Sent: Friday, November 17, 2017 9:41 AM To: You, David; Peabody, Charles; Miller, Geoffrey; Dixon, John; Lingam, Siva; Choate, Jackson Cc: Rein ert, Dusti n

Subject:

FYI: Palo Verde Plant Status 11/17/2017 Attachments: Rx Vessel flange leakage diagram.pdf Common Unit

  • None Palo Verde Unit 1 Mode/Power Level/Planned Power Changes: M ode: 1 / Full Power TS Action Statements:
  • Non Responsive Record Events/Conditions of Interest:
  • Non Responsive Record
  • Pressure has been observed to be increasing in the area between t he inner and out er o-rings. This indicates leakby pas.t the inner vessel head o-ring. Pressure currently is about 600#. This has occurred at least t wice in other units over t he past few years. Historically, t he licensee has allowed pressure in this space to increase to full RCS pressure

(-2250) and them implemented a T-mod which changes the alarm setpoint t o 2150 psig and lowering. This would give an indication of the outer o-ring having failed. Containment entry anticipat ed for Monday t o isolate the reactor vessel head leakoff line upstream manua l isolation valve V-211 (see attached figure) ...the solenoid valves downstream that can be used for venting this space are known to !leak by. Below I've pasted the status from the most recent time t his inner o-ring leakage occurred in Unit 3.

From Unit 3 in 2014-2015

  • 2/27/15: ODMI alarm setpoint of 1950 psig was reached yesterday for the reactor vessel head a-ring annulus decreasing pressure. Pressure is decreasing about 2# per hour. Licensee has not observed any changes in the reactor drain tank conditions, containment radiation monitors, or the unidentified leakage

surveillance. Action directed by the ODMI is to reconvene the ODMI team. Licensee is planning for a containment entry and will hold an ODMI challenge meeting today.

  • 5/19/14: Friday afternoon, low pressure alarm received. Pressure lowered to 1395 psig. Containment entry Friday night to close leakoff line upstream manual isolation valve V-211. Pressure immediately increased. Currently at 2250 psig. Licensee believes seat leakage across solenoid valve V-403 was the cause of the lowering pressure.
  • 5/12/14: Reactor vessel a-ring annulus pressure increased and stabilized at 2250 psig on Friday 5/9. Temporary modification installed which changes the alarm setpoint to 2150 psig and lowering. Operators are monitoring count rates from containment rad monitor RU-1 for early indication of any leakage from outer a-ring.
  • 4/12/2014: Operators observed PT-118 increasing Non Responsive Record 2

Non Responsive Record Non Responsive Record INon Responsive Record Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Office: 623-393-3737 Cell: !(b)(6) I Email: dustin.reinert@nrc.gov 3

REACTOR VESSEL i---c;r I

- - -- - V21 7 fPT\ _l __ _(PsH\ CR

[X}-~ ~

(Hs\B04

-~ TO REACTOR 1---- - -- . DRAIN TANK HV403 MAIN 0 -RING POSITION PACKUP 0-RING POSITION REACTOR

- - VESSEL FLANGE REACTOR TO PRESSURE SWITCH VESSEL _.

(LEAK.A.GEDETECTION)

RC009-98 Figure 2 - 6 Reactor Vessel 0 -Ring Leak Detection Reactor Vessel Head Seal D rain Valve Control (HS-403)

A two positio n (OPEN/CLOSE) control switch is provided on B04 in the main cont ro l room fo r control of the reactor vessel head seal drain valve. It is used w hen the pressure of the reacto r vessel 0-rings ind icate high enough to have some leakage. In the OPEN position, coolant is drained to the reactor drain tank. In the C LOSE position, the reactor vessel head seal drain valve closes, isolating the drain line.

Closure Head Attachments The top portion of the closme head contains ninety eight penetrations in w hich are attached housings for 89 operable control e lement drive mechanisms (CEDMs) a nd one 3/4 inch vent line with a butt weld connection. Of the remaining 8 penetrations, unit I contains CEDM housings and u nits 2 & 3 contain pressure housings. Two of the 8 spare hous ings are utilized for the reactor vessel level monitoring system (RVLM S). These 2 housings arc located symetrically opposite from each other and approximately 90° 03/05/99 24 RC STM / Volume 39 / Rev. 5

From: Reinert, Dustin Sent: Friday, November 17, 2017 9:41 AM To: You, David; Peabody, Charles; Miller, Geoffrey; Dixon, John; Lingam, Siva; Choate, Jackson Cc: Rein ert, Dusti n

Subject:

FYI: Palo Verde Plant Status 11/17/2017 Attachments: Rx Vessel flange leakage diagram.pdf Common Unit

  • None Palo Verde Unit 1 Mode/Power Level/Planned Power Changes: M ode: 1 / Full Power TS Action Statements:
  • Non Responsive Record Events/Conditions of Interest:
  • Non Responsive Record
  • Pressure has been observed to be increasing in the area between t he inner and out er o-rings. This indicates leakby pas.t the inner vessel head o-ring. Pressure currently is about 600#. This has occurred at least t wice in other units over t he past few years. Historically, t he licensee has allowed pressure in this space to increase to full RCS pressure

(-2250) and them implemented a T-mod which changes the alarm setpoint t o 2150 psig and lowering. This would give an indication of the outer o-ring having failed. Containment entry anticipat ed for Monday t o isolate the reactor vessel head leakoff line upstream manua l isolation valve V-211 (see attached figure) ...the solenoid valves downstream that can be used for venting this space are known to !leak by. Below I've pasted the status from the most recent time t his inner o-ring leakage occurred in Unit 3.

From Unit 3 in 2014-2015

  • 2/27/15: ODMI alarm setpoint of 1950 psig was reached yesterday for the reactor vessel head a-ring annulus decreasing pressure. Pressure is decreasing about 2# per hour. Licensee has not observed any changes in the reactor drain tank conditions, containment radiation monitors, or the unidentified leakage

surveillance. Action directed by the ODMI is to reconvene the ODMI team. Licensee is planning for a containment entry and will hold an ODMI challenge meeting today.

  • 5/19/14: Friday afternoon, low pressure alarm received. Pressure lowered to 1395 psig. Containment entry Friday night to close leakoff line upstream manual isolation valve V-211. Pressure immediately increased. Currently at 2250 psig. Licensee believes seat leakage across solenoid valve V-403 was the cause of the lowering pressure.
  • 5/12/14: Reactor vessel a-ring annulus pressure increased and stabilized at 2250 psig on Friday 5/9. Temporary modification installed which changes the alarm setpoint to 2150 psig and lowering. Operators are monitoring count rates from containment rad monitor RU-1 for early indication of any leakage from outer a-ring.
  • 4/12/2014: Operators observed PT-118 increasing Non Responsive Record 2

Non Responsive Record Non Responsive Record Non Responsive Record Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Offic'1'*~ ~=~u....E.L...,

Cell: (b)(6)

Emai:

3

From: Reinert, Dustin Sent: Monday, December 18, 2017 9:25 AM To: You, David; Peabody, Charles; Miller, Geoffrey; Dixon, John; Lingam, Siva; Choate, Jackson; O'Banion (Watford), Margaret Cc: Rein ert, Dustin

Subject:

FYI: Palo Verde Plant Status 12/18/2017 Common Unit

  • M et tower data restored late Friday afternoon.

Palo Verde Unit 1 Mode/Power Level/Planned Power Changes: Mode: 1 / Full Power TS Action Statements:

o None Events/Conditions of Interest:

  • Non Responsive Record
  • 11/30: Pressure in the area between the inner and outer o-rings currently at 700#.

o 11/21: Valve V-211 was shut yesterday. Current ly pressure is at 608# and slowly rising.

o 11/17: Pressure has been observed to be increasing in the area between the inner and outer o-rings. This indicates leakby past the inner vessel head o-ring. Pressure currently is about 600#. This has occurred at least twice in other units over the past few years. Historically, the licensee has allowed pr essure in this space to increase to full RCS pressure (-22so) and them implemented a T-mod which changes the alarm setpoint to 2150 psig and lowering. Th is would give an indication of the outer o-ring having failed. Containment entry anticipated for Monday to isolate the reactor vessel head leakoff line upstream manual isolation valve V-211 (see attached figure) ...the solenoid valves downstream that can be used for venting this space are known to leak by. Below I've pasted the status from the most recent time this inner o-ring leakage occurred in Unit 3.

Non Responsive Record

Non Responsive Record Non Responsive Record Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Office: 623-393-3737 Cell: (b)(6)

Email: nrc.gov 2

From: Reinert, Dustin Sent: Thursday, December 21, 2017 9:03 AM To: You, David; Peabody, Charles; Miller, Geoffrey; Dixon, John; Lingam, Siva; Choate, Jackson; O'Banion (Watford), Margaret Cc: Rein ert, Dustin

Subject:

FYI: Palo Verde Plant Status 12/21//2017 Common Unit

  • None Palo Verde Unit 1 Mode/Power Level/Planned Power Changes: Mode: 1 / Full Power TS Action Statements:

0 !Non Responsive Record Events/Conditions of Interest:

o 12/19: Pressure in the area between the inner and outer a-rings currently at "'600#. No change in the past few weeks. 0-ring may have reseated.

o 11/21: Valve V-211 was shut yesterday. Currently pressure is at 608# and slowly rising.

o 11/17: Pressure has been observed to be increasing in the area between the inner and outer o-rings. This indicates leakby past the inner vessel head o-ring. Pressure currently is about 600#. This has occurred at least twice in other units over the past few years. Historically, the licensee has allowed pressure in this space to increase to full RCS pressure ("'2250) and them implemented a T-mod which changes the alarm setpoint to 2150 psig and lowering. This would give an indication of the outer o-ring having failed. Containment entry ant icipated for Monday to isolate the reactor vessel head leakoff line upstream manual isolation valve V-211 (see attached figure) ...the solenoid valves downstream that can be used for venting this space are known to leak by. Below I've pasted the status from the most recent time this inner o-ring leakage occurred in Unit 3.

Non Responsive Record Non Responsive Record

Non Responsive Record INon Responsive Record Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Office: 623-393-3737 Cell: (b)(6)

Emai : us 1n.reinert nrc.gov 2

From: Hagar, Bob Sent: Tuesday, April 21, 2015 11:24 AM To: Reinert, Dustin

Subject:

RE: Follow- up to reactor vessel fl ange issue Thank you, Dustin!

From: Reinert, Dustin Sent: Tuesday, April 21, 2015 11:09 AM To: Hagar, Bob Cc: Peabody, Charles; You, David; Brandt, Lindsay

Subject:

Follow-up to reactor vessel flange issue

Bob, Near the start of each outage, the licensee does an as-found inspection of the inner and outer reactor vessel flange o-rings. The inner o-ring (the one which had been leaking for several months) had what I would call a small dimple near stud hole #22 (see attached picture) which was indicative of a leakage pathway.

This past weekend while reflooding the refueling cavity prior to core reload, the licensee performed a visual inspection of the vessel flange and identified a small pit on the flange at this same location.

The second page of each of the attached newsletters describes the repair strategy and the second newsletter shows a picture of the as-left repaired area.

Note that new o-rings are installed before each operating cycle.

Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station g: 1%)?l?-393-3731 Email: dustin.reinert@nrc.gov

From: Hagar, Bob Sent: Tuesday, April 21, 2015 11:24 AM To: Reinert, Dustin

Subject:

RE: Follow- up to reactor vessel fl ange issue Thank you, Dustin!

From: Reinert, Dustin Sent: Tuesday, April 21, 2015 11:09 AM To: Hagar, Bob Cc: Peabody, Charles; You, David; Brandt, Lindsay

Subject:

Follow-up to reactor vessel flange issue

Bob, Near the start of each outage, the licensee does an as-found insp ection of the inner and outer reactor vessel flange o-rings. The inner o-ring (the one which had been leaking for several months) had what I would call a small dimple near stud hole #22 (see attached picture) which was indicative of a leakage pathway.

This past weekend while reflooding the refueling cavity prior to core reload, the licensee performed a visual inspection of the vessel flange and identified a small pit on the flange at this same location.

The second page of each of the attached newsletters describes the repair strategy and the second newsletter shows a picture of the as-left repaired area.

Note that new o-rings are installed before each operating cycle.

Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Offic

  • 7 Cell: (b)(6)

@nrc.gov

From: Reinert, Dustin Sent: Friday, February 27, 2015 9:17 AM To: You, David; Peabody, Charles; Parks, Brian; Hagar, Bob; Tice, Jan; O'Banion (Watford),

Margaret; Hay, Michael; Brandt, Lindsay

Subject:

FYI: PVNGS Status 2/27/2015 Attachments: Rx Vessel flange leakage diagram.pdf Non Responsive Record Non Responsive Record Palo Verde Unit 3 TS Action Statements:

Day 5 of Diesel Generator A "super-outage" in progress. Expected to complete final retest overnight.

  • 3.8.1 Condition B (10 days) EOG "A" inoperable for planned maintenance Events/conditions of interest:

Reactor Vessel Head Flange Leak Detection Update. See attached figure also.

  • 2/2 7/15: ODMI alarm setpoint of 1950 psig was reached yesterday for the reactor vessel head o-ring annulus decreasing pressure. Pressure is decreasing about 2# per hour. Licensee has not observed any changes in the reactor drain tank conditions, containment radiation monitors, or the unidentified leakage surveillance. A,ction directed by the ODMI is to reconvene the ODMI team. Licensee is planning for a containment entry and will hold an ODMI chaillenge meeting today.
  • 5/19/14: Friday afternoon, low pressure alarm received. Pressure lowered to 1395 psig. Containment entry Friday night to close leakoff line upstream manual isolation valve V-211. Pressure immediately increased. Currently at 2250 psig. Licensee believes seat leakage across solenoid valve V-403 w as the cause of the lowering pressure.
  • 5/12/14: Reactor vessel o-ring annulus pressure increased and stabilized at 2250 psig on Friday 5/9. Temporary modification installed which changes the alarm setpoint to 2150 psig and lowering. Operators are monitoring count rates from containment rad monitor RU-1 for early indication of any leakage from outer o-ring.
  • 4/12/2014: Operators observed PT-118 increasing Other 11
  • 1 rain predicted for Sunday
  • DNMS exited yesterday afternoon with no issues.

Resident Coverage: Charley Peabody Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Offic)<.*......w.1.w...w.>t.1-..i..w.1..,

Cell: (b)(6) nrc.gov 2

From: Reinert, Dustin Sent: Friday, May 16, 2014 9:02 AM To: Baquera, Mica; Brown, Tony; Hagar, Bob; Parks, Brian; Rankin, Jennivine; Taylor, Nick; Tice, Jan; Reinert, Dustin

Subject:

PVNGS Status May 16 2014 Non Responsive Record Non Responsive Record Palo Verde Unit 3 TS Action Statements:

  • None Events/conditions of interest:
  • 5/12: Reactor vessel o-ring annulus pressure increased and stabilized at 2250 psig on Friday 5/9. Temporary modificat ion installed which changes the alarm setpoint to 2150 psig and lowering. Operators are monitoring count rates from containment rad monitor RU-1 for early indication of any leakage from outer o-ring.
  • Dustin has weekend coverage.

Resident Coverage: Dustin Reinert 1-888-455-9242, passcode!(b)(6) j!

Dustin Reinert Resident Inspector Palo Verde Nuclear Generating Station Office: 623-393-3737 Cell: (b)(6) 2

From: Egli, Richard To: Taylor Nick; Burritt Arthur Cc: Q"Keefe Neil: Tindell Brian: Correll Brian; Jobin Margaret; Pixon lobo

Subject:

RE: Unit 1 RCS Leakage ODMI Date: Tuesday, July 21, 201511:44:1 1 AM Note: The first attachment , consisting of two pages, is Attachments: Turkey Point o-Ring o&MR ons.odf withheld in its entirety under FO IA exemption 4. The imageoo1 oog 2nd attachment is the NRC logo.

Attached occurred at Turkey point circa mid-80's From: Taylor, Nick Sent: Tuesday, July 21, 2015 11:30 AM To: Burritt, Arthur Cc: OKeefe, Neil; Tindell, Brian; Correll, Bria n; Egli, Richard; Tobin, Ma rgaret; Dixon, John

Subject:

FW: Unit 1 RCS Leakage ODMI

Art, You were looking for OE last week on operations with both inner and outer O rings failed. I hadn't discovered any previous occurrences. But we got ANO's ODMI today and the licensee lists a number of previous occurrences both at ANO, Browns Ferry 1 and Wolf Creek.

Hope this helps, Nick Taylor Senior Project Engineer Division of Reactor Project s n:::::tv USNRC Region IV r /

R From: Correll, Brian Sent: Tuesday, July 21, 2015 9:11 AM To: OKeefe, Neil; Taylor, Nick; Farina, Thomas; Tindell, Brian; Tobin, M argaret Cc: Egli, Richard

Subject:

Unit 1 RCS Leakage ODM I Unit 1 RCS Leakage ODMI is attached. Pages 5 and 6 are the Thresholds and Actions to be taken if the thresholds are exceeded.

Brian

From: Tobin, Margaret To: Newman Garrett Subj ect: RE: Unit 1 RCS Leakage ODMI Date: Monday, July 27, 2015 10:00:00 AM Note: The attached image is the NRC Attachments: image001.png logo.

Thanks Garrett!

From: Newman, Garrett Sent: Tuesday, July 21, 2015 1:04 PM To: Tobin, M argaret

Subject:

FW: Unit 1 RCS Leakage OD M I Didn't know you left HQ to become a resident too. Welcome !

From: Burritt, Arthur Sent: Tuesday, July 21, 2015 1:42 PM To: Pinson, Brandon; Newman, Garrett; Rich, Sara h; Setzer, Thomas; Stewart, Scott

Subject:

FW: Unit 1 RCS Lea kage ODMI FYI From: Egli, Richard Sent: Tuesday, July 21, 2015 12:44 PM To: Taylor, Nick; Burritt, Arthur Cc: OKeefe, Neil; Tindell, Brian; Correll, Bria n; Tobin, Margaret; Dixon, John

Subject:

RE: Unit 1 RCS Leakage ODMI Attached occurred at Turkey point circa mid-80's From: Taylor, Nick Sent: Tuesday, July 21, 2015 11:30 AM To: Burrit t, Arthur Cc: OKeefe, Nei l; Tindell, Brian; Correll, Bria n; Egli, Richard; Tobi n, Margaret; Dixon, John

Subject:

FW: Unit 1 RCS Leakage OD M I

Art, You were looking for OE last week on operations with both inner and outer O rings failed. I hadn't discovered any previous occurrences. But we got ANO's ODMI today and the licensee lists a number of previous occurrences both at ANO, Browns Ferry 1 and W olf Creek.

Hope this helps, Nick Taylor Senior Project Engineer Division of Reactor Projects USNRC Region IV 0 : (817) 200-1520 C:!(b)(6)  !

E: nick.taylor@nrc.gov From: Correll, Brian Sent: Tuesday, July 21, 2015 9:11 AM

To: OKeefe, Nei l; Taylor, Nick; Farina, Thomas; Tindell, Brian ; Tobin, Margaret Cc: Eg li, Richard

Subject:

Unit 1 RCS Leakage ODMI Unit 1 RCS Leakage ODMI is attached. Pages 5 and 6 are the Thresholds and Actions to be taken if the thresholds are exceeded.

Brian

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