ML17060A146

FOIA/PA-2017-0110 - Resp 1 - Final, Agency Records Subject to the Request Are Enclosed. Part 1 of 6

ML17060A146
Person / Time
Site:	Wolf Creek
Issue date:	02/10/2017
From:	NRC/OCIO
To:
Shared Package
ML17060A140	List: ML17060A143 ML17060A146 ML17060A149
References
FOIA/PA-2017-0110
Download: ML17060A146 (381)

Text

{{#Wiki_filter:From: Taylor, Nick Sent: 15 Sep 2016 15:45:31-0500 To: Drake, James;Anchondo, Isaac; Kopriva, Ron;Werner, Greg;Graves, Sa muel;Alley, David

Subject:

FW: Wolf Creek Pictu res Attachments: DSC04761.jpg, DSC04765.jpg, DSC04766.jpg, DSC04764.jpg, DSC04747.jpg, DSC04720.jpg, DSC04719.jpg, DSC04714.jpg

All, A few pictures from Doug Dodson's tour of containment last week. Shows some corrosion products on the head in addition to boron. Will be really interesting to see what they find when the insulation comes off.

Nick From: Janicki, Steven Sent: Thursday, September 15, 2016 9:03 AM To: Taylor, Nick <Nick.Taylor@nrc.gov> Cc: Proulx, David <David.Proulx@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

Subject:

Wolf Creek Pictures

Nick, I found some of the pictures that show the rust that Doug was talking about. I have alll the pictures downloaded and will place them in the branch folder so you can quickly scan them if you would like at some point (vice having to click them all individually on Certrec).

Steve. Respectfully, Steve Janicki ~ Nuclear Regulatory Commission RIV - Division of reactor Projects Branch B PE (0) 817-200-1457 (C)l(b)(6) I Steven.janicki@nrc.gov

From: Tsao, John Sent: 15 Sep 2016 14:14:49 -0400 To: Alley, David;Hoffman, Keith;Kalikian, Roger Cc: Hsu, Kaihwa;li, Yong

Subject:

RE: Mechanical Clamp in ASME Section Ill I think that we should call the contraption installed on the canopy seal at Wolf Creek as a "mechanical joint", not as a "mechanical clamp". This is because NB-3671.7 permits the installation of mechanical joints (see Keith's email below). NB-3671.7 Sleeve Coupled and Other Patented Joints. Mechanical joints, for which no standards exist, and other patented joints may be used provided the requirements of (a), (b), and (c) below are met. (a) Provision is made to prevent separation of the joints under all Service Loadings. (b) They arc accessible for maintenance, removal, and replacement after service. (c) Either of the following two criteria are met. ( I) A prototype joint has been s ubjected to performance tests to determine the safoty of the joint under simulated service conditions. When vibration, fatigue, cyclic conditions, low temperature, thermal expansion, or hydraulic shock is anticipated, the applicable conditions shall be incorporated in the tests. T he mechanical joints shall be sufficiently leak tight to satisfy the requirements of the Design Specifications. (2) Joints are designed in accordance with the rules of NB-3200. A "mechanical clamp" as per ASME Section XI, Appendix IX or Appendix W, is not permitted to be installed on Class 1 piping and has a limited service time period (to the next refueling outage). From: Alley, David Sent: Thursday, September 15, 2016 1:26 PM To: Hoffman, Keith <Keith.Hoffman@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Cc: Hsu, Kaihwa <Kaihwa.Hsu@nrc.gov>; Li, Yong <Yong.li@nrc.gov>

Subject:

FW: M echanica l Clamp in ASME Section Ill Based on Robert's view, below, it appears that the clamp is acceptable per the construction code. This would appear to make the use of the clamp a code repair as it is in accordance with the construction code. This would appear to mean that the plant can install and leave the clamps on forever and that we have no regulatory hook (other than, potentially, the condition of the threads for this instance based on the extent of leakage). Any thoughts? Dave From: Hsu, Kaihwa Sent: Thursday, September 15, 2016 8:25 AM To: Alley, David <David.Alley@nrc.gov> Cc: Li, Yong <Yong.Li@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill Dave: I don't see any problem for the patented clamp to be used over top of original joint as long as the repair meets ASME Section Ill Code criteria.

Robert From: Li, Yong Sent: Thursday, September 15, 2016 7:29 AM To: Hsu, Kaihwa <Kaihwa.Hsu@nrc.gov>

Subject:

FW: Mechanical Clamp in ASME Section Ill Please respond to Dave. From: Alley, David Sent: Wednesday, September 14, 2016 8:19 PM To: Hoffman, Keith <Keith .Hoffman@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>; Li, Yong <Yong.Li@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill

Keith, Very wise. When you don't know the answer, make is someone else's problem

Yong, As Keith points out below, we could use some section Ill help. NB-3671.7 Sleeve Coupled and Other Patented J oints says that you can use sleeve coupled and other patented joints. Doesn't seem like there is much rigor in their qualification. That aside, the real question is that we have an instance where Wolf Creek appears to be installing a mechanical clamp over a threaded joint/canopy seal based on this code paragraph. It is pretty apparent to me that the patented joint could be used in place of the threaded connection/seal weld. It is not quite so apparent that such a joint is permitted to be used over the top of another type of joint. Your thoughts?

Dave From: Hoffman, Keith Sent: Wednesday, September 14, 2016 8:08 PM To: Alley, David <David.Alley@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill I believe that is probably a question for Yong Li's branch. Obviously the licensees and ABB - CE/Westinghouse believe it can be and that is what the patent says it was designed to do. From : Alley, David Sent: Wednesday, September 14, 2016 4:19 PM To: Hoffman, Keith <Keith.Hoffman@nrc.gov>; Tsao, John <John .Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill 3671.7 seems to allow almost joint that has had a mockup made and tested. However, it is in a section for nonwelded pipe joints. In my mind this would allow such a joint in place of a threaded joint. Does it allow the application of such a joint over an existing joint?

Dave From : Hoffman, Keith Sent: Wednesday, September 14, 2016 6:39 AM To: Tsao, John <John .Tsao@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill This was the section that was referenced as the applicable Section Ill paragraph in one of the documents I looked at yesterday. Specifically NB-3671.7 was referenced. NB-3671.3 shows the threaded joint and the requirement for the weld. NB-3670 SPECIAL PIPING R EQUI REMENTS NB-3671 Selection and Limitat ion of Nonwelded Piping Joints The type of piping joint used shall be suitable for the Design Loadings and shall be selected with consideration of joint tightness, mechanical strength, and the nature of the fluid handled. Pipingjoints shall conform to the requirements of this Subsection with leak tightness being a consideration in selection and design of joints for piping systems to satisfy the requirements of the Design Specifications. NB-3671.1 Flanged J oints. Flanged joints are permitted. NB-3671.2 Expanded Joints. Expanded joints shall not be used. NB-3671.3 T hreaded Joints. Threaded joints in which the threads provide the only seal shall not be used. If a seal weld is employed as the sealing medium, the stress analysis of the joint must inc.ludc the stresses in the weld resulting from the relative deflections of the mated parts. NB-3671.4 F lared, Flareless, and Compression Joints. Flared, flareless, and compression type tubing fittings may be used for tubing sizes not exceeding 1 in. 0.D. (25 mm) within the limitations of applicable standards and specifications listed in Table NCA-7100- 1 and requirements (b) and (c) below. In the absence of such standards or specifications, the Designer shall determine that the type of fitting selected is adequate and safe for the Design Loadings in accordance with the requirements of(a), (b), and (c) below. (a) The pressure design shall meet the requirements ofNB-3649. (b) Fittings and their joints shall be suitable for the tubing with which they are to be used in accordance with the minimum wall thickness of the tubing and method of assembly recommended by the manufacturer. (c) Fittings shall not be used in services that exceed the manufacturer's maximum pressure-temperature recommendations. NB-3671.S Caulked Joints. Caulked or leaded joints shall not be used. NB-3671.6 Brazed and Soldered Joints. (a) Brazed Joints ( 1) Brazed joints of a maximum nominal pipe size of I in. (DN 25) may be used only at dead end instrument connections and in special applications where space and geometry conditions prevent the use of joints permitted under NB-3661.2, NB-3661.3, and NB-3671.4. The depth of socket shall be at least equal to that required for socket welding fittings and shall be of sufficient depth to develop a rupture strength equal to that of the pipe at Design Temperature (N B-4500). (2) Brazed joints that depend upon a fillet rather than a capillary type filler addition arc not acceptable. (3) Brazed joints shall not be used in systems containing flammable fluids or in areas where fire hazards are involved. (b) Soldered Joints. Soldered j oints s hall not be used. NB-3671.7 Sleeve Coupled and Other Patented Joints. Mechanical joints, for which no standards exist, and other patented joints may be used provided the requirements of(a), (b), and (c) below are met. (a) Provision is made to prevent separation of the joints under all Service Loadings. (b) They are accessible for maintenance, removal, and replacement after service. (c) Either of the following two criter ia are met. (1) A prototype joint has been subjected to performance tests to determine the safety of the joint under simulated service conditions. When vibration, fatigue, cyclic conditions, low temperature, thermal expansion, or hydraulic shock is anticipated, the applicable conditions shall be incorporated in the tests. The mechanical joints shall be sufficiently leak tight to satisfy the requirements of the Design Specifications. (2) Joints are designed in accordance w ith the rules of NB-3200. Keith M. Hoffman Materials Engineer

NRR/DE/EPNB (301)415-1294 From: Tsao, John Sent: Tuesday, September 13, 2016 6:00 PM To: Alley, David <David.Alley@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

Mechanical Clamp in ASME Section Ill I did a word search of "Clamp" in NB and NC sections of the 2007 ed ition and 2013 edition of the ASME Code, Section Ill. There were 4 hits in both ed itions. NB-1132.1- the clamp in this article is related to pipe attachment (the clamp used for pipe supports) NB-3411.l(d) pump clamp NB-3651.3 pipe clamp as in pipe supports such as hangers or snubbers that use clamps. NB-4231---clamps used in welding operations (when welding 2 pieces of pipe, welders use clamps) So ASME SEction Ill does not have requirements or specification for the mechan ical clamp application that was used on the CRDM canopy sea l at Wolf Creek.

From: Collins, Jay Sent: 6 Sep 2016 20:34:47 +0000 To: Drake, James

Subject:

RE: Wolf Creek Boric Acid Leaking on Head Sorry, Dave thought it was you. Thanks for passing it on. Jay From: Drake, James Sent: Tuesday, September 06, 2016 4:34 PM To: Collins, Jay

Subject:

RE: Wolf Cree k Boric Acid Leaking on Head Thank you for the reminder Jay. I will pass this on to Ron. He is the inspector for this outage. Jim From: Collins, Jay Sent: Tuesday, September 06, 2016 3:32 PM To: Drake, James <James.Drake@nrc.gov>

Subject:

Wolf Creek Boric Acid Leaking on Head Greetings, Catching this issue from the sidelines, but I thought I would put a bee in your ear to remind you about the problems we had with Fort Calhoun and the cleaning of their head last year. I dona't know the inspection requirements for Wolf Creek this refueling outage, but I figure they are at least going to have to clean the head for a VT-2 inspection. Cleaning the head in too aggressive of a manner can invalidate the visual inspection and may then trigger a vollumetric inspection. Just a heads up for a problem they may not be thinking about, using lessons learned that Region IV caught earlier at Fort Calhoun. This was lsaaca's issue at Fort Calhoun, so I am sure he has all the fine details. Just trying to be helpful, Jay

From: Hoffman, Keith Sent: 6 Sep 2016 15:36:41 -0400 To: Alley, David;Collins, Jay;Tsao, John;Davis, Robert

Subject:

FW: OpE - RCS LEAKAGE RESULTS IN TECHNICAL SPECIFICATIONS SHUTDOWN Leak identified? RFO started. Keith M. Hoffman Materials Engineer NRR/DE/EPNB (301)415-1294 From: Pannier, Stephen Sent: Tuesday, September 06, 2016 3:33 PM To: Hoffman, Keith; Alley, David

Subject:

OpE - RCS LEAKAGE RESULTS IN TECHNICAL SPECIFICATIONS SHUTDOWN The following EN is provided for your information. EN 52218 - WOLF CREEK-TECHNICAL SPECIFICATllON REQUIRED SHUTDOWN While operating in MODE 1 at 100 percent rated thermal power and placing Excess Letdown in service for Reactor Coolant System (RCS) leak detection, RCS operational leakage exceeded 1 gpm [gallon per minute] unidentified leakage as identified by performing RCS Water Inventory Balance using the Nuclear Plant Information System Computer. This required the entry into Technical Specification (TS) Limiting Condition of Operation (LCO) 3.4.13 Condition Bat 0808 [CDT] on 9/2/16. The associated action is to place the unit into Mode 3 in 6 hours. Trending of containment sump level indicates the leakage is inside containment with the exact location within containment unknown. Containment inspection is being performed to try and identify the source of Reactor Coolant System leakage. NRC Resident Inspector has been notified. Re-alignment of the Letdown System back to its normal arrangement has subsequently reduced RCS leak rate to 0.521 gpm at 0652 CDT on 9/2/16. Unusual or Not Understood - Leak Location is not known at this time. Maximum leak rate recorded was 1.358 gpm. The leak was first discovered at 08/31/16 at 1519 CDT. Safety Related Equipment not operational - Reactor Vessel Level Indicating System (TS 3.3.3). From the Region IV Daily Safety Call: The licensee made a containment entry and eventually found the source of the unidentified leakage. While looking down on the vessel head the licensee identified signs of a boric acid leak over a mirrored insulation panel. After removing the panel and using a camera the licensee saw a plume in the area of several penetrations. The licensee was able to determine that the leak was on a core exit thermocouple nozzle threaded connection. The licensee also determined that this was not pressure boundary leakage. In addition, the licensee identified that excess letdown made the leak rate seem worse than the actual value. The leak rate was eventually quantified at around 0.6 gpm. Without being pressure boundary leakage and since the leak rate was less than 1 gpm, the licensee was able to exit the LCO. The licensee has decided to go into their planned refueling outage and will perform some pre-outage surveillances before cooling down to MODE 5. The leak will be repaired during the refueling outage while the head is on the stand.

From: Poehler, Jeffrey Sent: 8 Sep 2016 13:09:42 -0400 To: Alley, David;Butcavage, Alexander Cc: Collins, Jay

Subject:

RE: HAVE YOU HEARD ANY DETAILS On THIS JEFF??? Tha nks, Jeff From : Alley, David Se nt: Thursday, September 08, 2016 1:05 PM To: Poehler, Jeffrey; Butcavage, Alexander Cc: Collins, Jay Subje ct: RE: HAVE YOU HEARD ANY DETAILS On THIS JEFF??? I am pretty sure I started on a response to this yesterday and didna't get it sent. Leak is in a canopy seal weld. Pressure boundary is a threaded joint. Weld is just for leak tightness. Weld is stainless to stainless using stainless fil ler. Wolf Creek was going to go into an outage in a couple weeks. They just are starting the outage early. Initial thoughts are to use a mechanical clamp but region IV will be looking into that approach. They apparently have something like 10 mechanical clamps already installed. Dave From: Poehler, Jeffrey Se nt: Wednesday, September 07, 2016 2:09 PM To: Butcavage, Alexander <Alexander.Butcavage@nrc.gov> Cc: Collins, Jay <Jay.Collins@nrc.gov>; Alley, David <David.Alley@nrc.gov>

Subject:

RE: HAVE YOU HEARD ANY DETAILS On THIS JEFF??? No, I had not. Copied folks from the branch that handles upper head penetrations. Jeff From : Butcavage, Alexander Sent: Wednesday, September 07, 2016 1:09 PM To: Poehler, Jeffrey <Jeffrey.Poehler@nrc.gov> Subje ct: HAVE YOU HEARD ANY DETAILS On THIS JEFF??? NRC Says Wolf Creek Leak Caused By Reactor Vessel Head Penetration Nozzle Issu e. WTBW-TV Topeka, KS (9/6, Palmer, 78K) reported on its website that NRC Public Affairs Officer Victor Dricks acetold 13 NEWS the leak was not caused by a bad weld as some reports have indicated. Dricks said a~The leak was in a penetration nozzle at the top of the reactor vessel.a'a D But the issue was not related to the weekend ea11hquake affecting seven states, nor did any radiation escape the facility from the water leak. ace0n a related topic, Wolf Creek officials are meeting with Nuclear Regulatory officials in Arlington, Texas On Sept. 21 to discuss what they called, ae ... an apparent violation in ma intaining emergency diesel generators at the plant.a'a D Officia ls acesaid that a generator failed 3 hours into a 24-hour run due to a fau lty electrical component, however, there was no danger, because other means were available to supply emergency power to the plant if needed.a Ll The Texas meeting w ill determine whether add itional inspections and oversight are needed.

From: Tsao, John Sent: 12 Sep 2016 13:39:06 -0400 To: Collins, Jay

Subject:

FW: Wolf Creek CSC."A""'""'

                                                  'a" ' - - - - - - - - - - - - - - - .

Attachments: DCP 012962.docx I ~45-page WCNOC Change Pkg.

                                              .0 12962 withheld in full under ex4.

Fyi leaking at the CROM canopy seal weld at wolf creek. They used a mechanical clamp and 50.59 evaluation. Region IV has questions on the licenseea's 50.59 evaluation From : Drake, James Se nt: Monday, September 12, 2016 1:27 PM To: Lyon, Fred ; Alley, David ; Werner, Greg Cc: Anchondo, Isaac; Kopriva, Ron ; Tsao, John ; Hoffman, Keith ; Taylor, Nick; Dodson, Douglas ; Thomas, Fabian

Subject:

RE : Wolf Creek CSCA'a Attachment would help. Jim From : Drake, James Se nt: Monday, September 12, 2016 12:13 PM To: Lyon, Fred <Fred .Lyon@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Werner, Greg

 <Greg.Werner@nrc.gov>

Cc: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Tsao, John

 <John.Tsao@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>;

Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian .Thomas@nrc.gov>

Subject:

Wolf Creek CSCA'a I highlighted so interesting statements. Looks like the Westinghouse analysis was only for stresses on the nozzle. The licensee appears to be treating the entire head adapter as the nozzle. The first comment in the document will take you to the 50.59 screening the licensee performed. Jim .James f:. :Drafi.e James F. Drake Office phone: 817-200-1558 Cell Phone:l(b)(B) I

From: Tsao, John Sent: 12 Sep 2016 14:40:31 -0400 To: Collins, Jay

Subject:

FW: Wolf Creek CSCA'a Attachments: WCNOC 30-day response to BL 2002-01.pdf, WCNOC 60-day response to BL 2002-01.pdf, NRC closeout of BL 2002-01 to RVH Inspection Order. pdf, NRC response to WCNOC BL 2002-01.pdf See listing of Records Already Ava ilable f yi From: Lyon, Fred Sent: Monday, September 12, 2016 2:38 PM To: Drake, James Cc: Alley, David; Werner, Greg; Tsao, John; Hoffman, Keith ; Kopriva, Ron; Anchondo, Isaac Subje ct: RE: Wolf Creek CSCA'a Other relevant documents attached. I havena't run down the Order trail yet. From: Lyon, Fred Se nt: Monday, September 12, 2016 1:40 PM To: Drake, James <James.Drake@nrc.gov> Cc: Alley, David <David .Alley@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Tsao, John

 <John.Tsao@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Kopriva, Ron
 <Ron.Kopriva@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>

Subject:

RE: Wolf Creek CSCA'a Only found one CROM canopy seal weld repair using a mechanical clamp - Salem in 1988. The Turkey Point one is containment boundary leakage repaired with a mechanical clamp (though the title says pressure boundary). Saw plenty use of MNSAs. I included the Robinson and Seabrook examples simply because the licensee originally considered mechanical clamps, but then did weld overlays. From: Drake, James Se nt: Monday, September 12, 2016 1:27 PM To: Lyon, Fred <Fred.Lyon@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Werner, Greg

 <Greg.Werner@nrc.gov>

Cc: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Tsao, John

 <John.Tsao@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>;

Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov> Subje ct: RE: Wolf Creek CSCA'a Attachment would help. Jim From: Drake, James Se nt: Monday, September 12, 2016 12:13 PM To: Lyon, Fred <Fred.Lyon@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Werner, Greg

 <Greg.Werner@nrc.gov>

Cc: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron .Kopriva@nrc.gov>; Tsao, John

 <John.Tsao@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>;

Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

Subject:

Wolf Creek CSCA'a I highlighted so interesting statements . Looks like the Westinghouse analysis was only for stresses on the nozzle. The licensee appears to be treating the entire head adapter as the nozzle.

The first comment in the document will take you to the 50.59 screening the licensee performed. Jim $ rmes 'f. :15rafe James F. Drake Office phone: 817-200-1558 Cell Phonej(b)(6) I

From: Tsao, John Sent: 12 Sep 2016 13:39:57 -0400 To: Collins, Jay

Subject:

FW: N-733 Attachments: ML13263A372.pdf, ML073240650.pdf See listing of Records A lready Available to the Public for these attachments. Background info on Wolf Creek. Pis see email below F rom: Drake, James Sent: Monday, September 12, 2016 9:09 AM To: Alley, David Cc: Tsao, John ; Hoffman, Keith

Subject:

FW: N-733

Dave, As soon as I get some additional information I will be giving you a call to discuss.

Jim F rom : Anchondo, Isaac Sent: Thursday, September 08, 2016 3:31 PM To: Drake, James <James.Drake@nrc.gov> Cc: Werner, Greg <Greg.Werner@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov> Subj ect: RE: N-733

All, I think before we can say that the CC is not applicable, we need to get an understanding on how the licensee is classifying this connection in terms of the ASME Code. Subsequently, we need an agency position on the applicability of Section XI for this particular joint. Given that this is not a weld, none of the weld "Examination Categories" are applicable other than B-P, "Pressure Retaining" components which requires a VT-2. Remember that a CC is an alternative to Section XI, and therefore, the licensee should be able to tell us the requirement needing an alternative.

With that said, the following is provided under the general IWA requirements for mechanical joints (2001 Edition), IWA-4321, "Class 1 Mechanical Joints" (c) Threaded joints in which the threads provide the onjy seal shall not be used in Class 1 piping systems. If a seal weld is em loyed as the sealing medium, the

stress analysis of the joint shall include the stresses in the weld resulting from the relative deflections of the mated parts. This is identical to the construction requirements for Class 1 threaded connections in piping (2001 Edition). NB-3671.3 Threaded Joints. Threaded joints in which the threads provide the only seal shall not be used. If a seal weld is employed as the sealing medium, the stress analysis of the joint must include the stresses in the weld resulting from the relative deflections of the mated parts. So the licensee should have a stress analysis for the seal welds to meet Section XI requirements, and in my opinion, CC N-733 is not an applicable alternative to this requirement. I think there needs to be further discussions with the licensee and HQ in regards to the applicability of the Code Case. Attached are a couple of relief request submitted to the NRC in regards to canopy seal weld leakage that used an actual weld overlay because repairs of the canopy seal weld are required by the Code. P.S. An additional concern would be that the seal weld provides the seal function while the treated connection provides the structural integrity of the joint. It appears that the licensee is assuming that the seal weld failed but what if the threaded connection is degraded. The clamp does not provide a structural integrity function as stated in the CC. Mechanical connection assemblies are permitted only for nozzles on which there are substantially no piping reactions, such as pressurizer heater penetrations and openings for instrumentation. he mechanical connection assembly and the vessel or piping location where the mechanical connection assembly is installed shall be designed taking no structural credit for the existing Category D or branch connection partial penetration weld and shall be based on the stress and fatigue requirements of NB-3200. Let me know if anybody has any questions.

Thanks, Isaac F r om : Drake, James Sent: Thursday, September 08, 20 16 I :54 PM To: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>

Cc: Werner, Greg <Greg.Werner@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>

Subject:

N-733

Isaac, I reread the Code case. I think you are correct, Code Case N-733 may not be applicable. In the reply it states: a mechanical connection modification that replaces the Category D or branch connection partial penetration weld and provides the primary pressure sealing function of the existing nozzle may be used to mitigate flaws in NPS 2 (ON 50) and smaller nozzles and nozzle partial penetration welds in vessels and piping originally constructed in Section Ill, Class 1 or Class A or 8 31. 7 Class 1, provided the following requirements are met: a) Mechanical connection assemblies are permitted only for nozzles on which there are substantially no piping reactions, such as pressurizer heater penetrations and openings for instrumentation. The mechanical connection assembly and the vessel or piping location where the mechanical connection assembly is installed shall be designed taking no structural credit for the existing Category D or branch connection partial penetration weld and shall be based on the stress and fatigue requirements of NB-3200. Another concern: Per MANDATORY APPENDIX IX, MECHANICAL CLAMPING DEVICES FOR CLASS 2 AND 3 PIPING PRESSURE BOUNDARY ARTICLE IX-1000 GENERAL (a) Mechanical clamping devices used as piping pressure boundary may remain in service only until the next refueling outage, at which time the defect shall be removed or reduced to an acceptable size. (b) These clamping devices may be used on piping and tubing, and their associated fittings and flanges, and the welding ends of pumps, valves, and pressure vessels, except as prohibited by (c) below. (c) Clamping devices shall not be used on the following : (1) Class 1 piping; (2) portions of a piping system that forms the containment boundary; (3) piping larger than NPS 2 (DN 50) when the nominal operating temperature or pressure exceeds 200°F (95°C) or 275 psig (1 900 kPa);

(4) piping larger than NPS 6 (ON 150). (d) A Repair/Replacement plan shall be developed in accordance with IWA-4150, and shall identify the defect characterization method, design requirements , and monitoring requirements. (e) Welding performed as part of the fabrication and installation of the clamping device shall be in accordance with the requirements of IWA-4400. Welding shall be documented on an NIS-2 Form. (f) The records required by IWA-6000 shall be maintained by the Owner until the clamping device is removed . I do not know what authorization the licensee used to install these other clamps. This may be another case like the "seal weld enclosures" at STP. I have asked for any OE on RVH penetration canopy seal weld leaks of this magnitude. Jim James F. Drake Office phone: 817-200*-1558 Cell Phone~....(b-)(6_) _ _ __.

Exelon Generation 630 c,-7 2000 011oce RS-13-234 10 CFR 50.55a September 19, 2013 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457

Subject:

Relief Request 13R-11 Associated with Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CROM) Canopy Seal Welds In accordance with 10 CFR 50.55a, "Codes and standards,* paragraph (a)(3)(i), Exelon Generation Company, LLC (EGC), is requesting NRC approval of a proposed alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda for Braidwood Station, Units 1 and 2. The proposed alternative would permit the use of an alternative method of repair and nondestructive examination for control rod drive mechanism (CROM) canopy seal welds. The CROM assemblies were designed and fabricated to the ASME B&PV Code, Section Ill, 1974 Edition through Summer 1974 Addenda. During boroscopic inspection of the reactor head assembly during the Braidwood Station, Unit 1 2013 fall refueling outage (i.e., A1R17), white residue was observed on the CROM canopy seal welds for reactor head penetrations 41, 49, 61, 65 and 73 indicating the potential of past reactor coolant system pressure boundary leakage at one or more of these locations. The locations of the leakage is suspected to be the omega seal welded threaded connection on one or more of these CROM penetrations. IWA-4000 of Section XI requires that repairs be performed in accordance with the original construction Code of the component or system, or later editions and addenda of the Code. The canopy seal weld is described in Section Ill and a repair to this weld would require: 1) an excavation of the rejectable indication(s); 2) a surface examination of the excavated area; 3) re-welding and restoration to the original configuration and materials; and 4) a final surface examination. An alternative to the Code repair process exists that provides an acceptable level of quality and safety, consistent with 10 CFR 50.55a(a)(3)(i). The alternative method also significantly reduces the projected occupational radiation dose when compared to the Code required repair method.

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381 -2000

     ~PR 0 1 \999 U. S . Nuclear Regulatory Commission ATTN : Document Control Desk Washington , D. C .                      20555 Gentlemen :

In the Matter of Docket No. 50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) CODE REQUIREMENTS - CONTROL ROD DRIVE MECHANISM (CRDM) CANOPY SEAL WELDS - FRACTURE MECHANICS ANALYSIS The purpose of this letter is to provide the fracture mechanics analysis on the CROM canopy seal weld repair . In TVA' s letter dated March 19, 1999, r equesting approval of an alternative weld repair and examination method to the ASME Code requirements, TVA corrunitted to provide the fracture mechanics analysis that was to be performed to support the alternative weld repair and examination method . Structural Integrity Associates , Inc. ( SIA) performed the evaluation for TVA. SIA h as indi cated that the f r acture mechanics analysis report SIR-97-089, " Design and Analysis of a We ld Overlay Repair for the watts Bar CROM Lower canopy Seal welds , " Revision 0, previously prepared for t he canopy seal repair during WBN \ Unit 1 Cycle 1 refueling outage, is directly applicable to the canopy seal weld for the GS penetration . The geometry and materia l s are identical to those previously repaired canopy seal welds. Based o n I the comparison, the overlay design for the canopy seal weld on the GS penetration is identical to the welds discussed in the Report SIR Perti~e~~' ~::o rmation in that report can be used ( \ ,~ 1 2 08 9, Revis ion 0 . 9904150261 990407 PDR ADOCK 05000390

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0 PDR

U . S. Nuclear Regulator y Corrunission Pp..,.q_e 2

~AfR      0 7 1999 for the G5 canopy seal weld including heat inputs and weld overlay dimensions . TVA submitted report SIR-97-089 o n October 10, 1997 . The NRC approved the relief request for the previously repaired welds on February 12, 1998 .

The repair of the canopy seal weld , G5, has been completed . An examination of the weld repair was performed by quality con trol personn el using a remote video camera with approximately BX magnification. The e x amination was also verified by a TVA Le vel III examiner and witnessed by TVA' s Authorized Nuclear Inservice Inspector. The required visual examination described in the request for relief , with the exception below, is documented in the associated work order. TVA had indicated in the "Justification For the Granting of Relief" section of the March 19 , 1999 , letter that the process of the repair would be recorded on video tape. This area of the process was not implemented . Video taping was a recorrunendation by SIA for record and review purposes only and not a requirement of the process. Therefore , TVA has enclosed a revision to request for relief , l-RR-2, deleting that statement . In sununary, TVA has completed the weld repair as discussed in the revised Request for Relief. The fracture mechanics analysis submitted to NRC October 10 ,

• 1997, for the WBN Unit 1 Cycle 1 canopy seal weld repairs , has been evaluated and determined to be directly applicable to the canopy seal weld repair for the G5 penetration identified in WBN Unit 1 Cycle 2 refueling outage . Therefore, an additiona l analysis is not being submitted . If you should have any questions concerning this matter , please telephon~ me at (423) 365-1824 .

s~::_ P . L. Pace Manager, Licensing and Industry Affairs Enclosure cc: See page 3

U. S . Nuclear Regulatory Commission r>age 3

  .APR  a* 7 1999 cc (Enclosure) :

NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. Robert E. Martin , Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 U. S . Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth st ., sw , suite 23T85 Atlanta , Georgia 30323 l

ENCLOSURE WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l-RR-2 , Revision 1

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l -RR- 2, Revision 1 Summary : During the Watts Bar Nuclear Plant Unit 1 Cycle 2 refueling outage activity of disassembl ing the reactor vessel , boric acid residue was noticed on the control rod drive mechanisms (CRDMs) (See Attachments 1 and 2} . Further inspection shows that one CRDM has started leaking at the lower canopy seal weld (See Coordinates on Attachment 3) . The ASME Code requires the defects be removed and the configuration of the material be reproduced in order to restore the canopy seal to its original design condition. Due to the physical space limitations and in consideration o f radiation exposure , Watts Bar proposes as an alternative to perform a weld buildup over the leaking canopy seal weld (See proposed design in Attachment 4) rather than removing the defect and performing a weld repair. Also, an enhanced visual examination is proposed as an alternative to the l iquid penetrant examination r equired by the original construction code for the final weld buildup. TVA's proposed alternative seal weld repair and examination methods have been previously implemented at Watts Bar , TVA' s Sequoyah Nuclear Plant Unit 1, and other utilities and provides a n acceptable level of quality and safety . TVA requests authorization t o use these alternatives in accordance with 10CFR50 . 55a(a) (3) (i). Unit : 1 System: Reactor Coolant - System 68 Component : Control Rod Drive Mechanism Code Class: 1 Function : Vertically position a control rod in the nuclear core by raising or lowering an interconnecting drive shaft . Code Requirement : ASME Section XI, 1989 Edition, IWA- 4110(a), " This Article provides rules and requirements for repair of pressure retaining components and their supports , including appurtenances, subassemblies, parts of a component, and core support structures, by welding , brazing , or metal removal . u Code Requirements From Which Relief is Requested : For repair of the defect, relief is requested from the following ASME Section XI, 1989 Edition, IWA-4000, Repai r Procedure requirements: E- 1

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAI R AND REPLACEMENT REQUEST FOR RELIEF l - RR-2 , Revision 1

a. Paragraph I WA-4120(a) , "Repairs shall be performed in accordance with the Owner ' s Design Specification and the original Construction Code of the component or system .

• Later Editions and Addenda of the constructi on Code or*of Secti on III, either in their entiret y or portions thereof, and Code Cases may be used . I f repair welding cannot be performed in accordance with these requirements, the applicable alternative requirements of IWA-4500 and the following may .be used :

( 1) IWB-4000 for Class* 1 components ." b . Paragraph I WA-4130(a) (2) , "Repair operations shall be performed in accordance with a program delineating essential requirements of the complete repair cycle inducting ... ( 2) .. . below : ( 2 ) the flaw removal method, method of measurement of the cavity created by removing the flaw, and dimensional requirements for reference points during and after the repair;" c . Subarticle IWA-4300, " Defect Removal," in its entirety . For examination, relief is requested f rorn the following ASME Section I I I , 1971 Edition, through Winter 1971 Addenda, Paragraph NB-5200, " Examination of Weld" requirements : d . Paragraph NB-5271 , "Welds of this type (welds of specially designed seals , i.e ., canopy seal welds) shall be examined by either the magnetic particle or liquid penetrant me thod." Basis for Relief : During the Unit 1 Cycle 2 Refueling Outage (U1C2 RFO) activity of disassembling the reactor vessel, boric acid residue was noticed on a control rod drive mechanism (CROM) . Further inspection shows that one CROM has started leaking at the l ower canopy seal weld . See Attachments 1 , 2 and 3 for configuration and location of the CRDM and canopy seal weld . The CRDMs are part of the nuclear steam supply sys t em procured from Westinghouse Electric Corporation under Contract 54114 . The CROM housings, as part of the reactor vessel , are within ' the reactor coolant system as defined by the Final Safety Analysis Report (FSAR). The 1971 Edition, Addenda through Winter 1972 of ASME Section I I I establish the design specification and the construction code for the CRDMs . The 1 971 Edition , Addenda through

• winter 1971 of ASME Section I I I establish the design specificati on and the construction code for the reactor vessel .

E-2

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l - RR-2, Revision 1 The CRDMs are fabricated in sections wi th thr eaded joi nts providing the pressure- retaining capabilities. Since the threaded joint provides pre ssure retention , the canopy seal weld is not pressure retaining and is for leakage control . The 1971 Edition, Addenda through Winter 1972 of ASME Section III does not allow threaded joints as the only seal as described in Paragraph NB- 3671 . 3 . Paragraphs NB-3227.7 and NB-4360 address the design of canopy seal welds and qualification requirements for welding specially designed welded seals , respectively . Paragraph NB-5271 requires that seal welds receive either a magnetic particle or liquid penetrant examination. Due to physical space limitations and in consideration of the need to keep worker dose as low as reasonably

• achievable (750 - 800 millirem per hour on contact and 100

               - 150 millirem per hour general area), removal a nd repair of the defect is not the most favorable method of repair .
             . In addition , if the defect was removed , it would be

impossible. to reproduce the configuration of the canopy seal to its original design condition as required by IWA-400 0 .

Alternative Repair Requirements : WBN will apply the following alternative weld overlay repair requirements:

a. A weld over.lay repair designed under the requirements of ASME Section XI, 1989 Edition, Paragraph IWB-3640, "Evaluation Procedures and Acceptance Criteria for Austenitic Piping, " and Appendix c, "Evaluation of Flaws in Austeni tic Piping," will be used as an alternative repair method. Guidance will also be taken from ASME Section XI Code Case N-504, "Alternative Rules for Repair of Class 1 , 2, and 3 Austenitic Stainless Steel Piping," and NUREG-0313, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, Final Report," Revision 2 .

WBN will apply the following alternative examination requirements: b . An enhanced visual examination using a remote video camera with a magnification of approximately BX will be used t o monitor the repair and to perform a visual examination of the final weld at the enhanced magnification . E- 3

r ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l-RR-2, Revision 1 Justif icati on For The Granting Of Relief: TVA ' s Code of Record for Repairs and Replacements is ASME Section XI , 1989 Edition . IWB- 3640 and Appendix C of the 1989 Edition of ASME Section XI will be u sed t o perform t he requi red fracture mechanics a nd to design a weld overlay repair of the flawed canopy seal weld . Portions of Code Case N-504 are also used for guidance . Code Case N-504 allows repair by addition of weld material without removal of the underlying defect to be considered as a code repair . *code Case N-504 is endorsed by the NRC in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability ASME Section XI Division," Revision 11 . IWB- 3640 provides criteria for acceptance of flaws without

repair in ductile , austenitic materials . The basis for

                  ; such accep~ance is the evaluation of the structural
                  , adequacy of the flawed component after considering the
                  ; predicted flaw growth over the evaluation period . The acceptance criteria is based upon the net section collapse (limit load) criteria which are defined in detail in Appendix C of Section XI.* Also , NUREG-0'313 1 "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping , Final Report , "

Revision 2 , is used for guidance. The use of NUREG-0313 will result in the repair design of the canopy seal we l d to be based upon conservative treatment of applied stresses, and includes allowance for continued flaw growth, as required by Section XI . The material used for the repair is Inconel 625 weld mate r ial which has a tensile strength of appro~irnately 110 kips per square inch (ksi) . The Inconel weld material is stronger than the underlying base material (304 stainless steel ) with a tensile strength of 75 ksi , more resistant to degradation mechanisms such as stress corrosion cracking, and is highly ductile. The load carrying

• capability of the repaired l ocation will be greater than

the original component .

Liquid penetrant examinations that are required by NB-527 1

i.  : wi l l not be performed because of space limitations, which prevent examiners the needed access to successfully perform the examination and in consideration of maintaining worker dose as low as reasonably achievabl e .

As an alternative, TVA will use a remote video camera with a magnif i cation of approximately BX and perform a visual examination of the final weld at the enhanced magnification . The basis for this approach is that post-weld liquid penetrant exami nations are surf ace examinations , and provides minimal assurance of repair i nt egrit y when compared to an enhanced visua l e xaminat ion. E- 4

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l-RR- 2 , Revision 1 Additionally, fracture mechanics analyses have been performed for other plants which demonstrates that the critical flaw size (i . e ., the flaw size , which would lead to the incipient collapse of the repair under code allowable applied stress conditions) is significantly larger than a flaw that would be rel iably detected by the enhanced visual examination . The fracture mechanics analysis assumes that an initial defect is completely through the repair membrane. Thus , there is large margin of safety in the analysis . TVA considers the fracture mechanics analysis , coupled with the enhanced visual examination, suitable to provide an acceptable alternative to the code required liquid

• penetrant examination .

           '.TVA has performed a demonstration examination for the

Au.thorized l Nuclear Inspector using the remote video l equipment at Sequoyah Nuclear Plant's (SQN) Unit 1. That demonstration was performed prior to its use for examination of repair of canopy seal welds at SQN and the results documented in a letter to the NRC dated Apr il 3 ,

1996. The demonstration was performed using a machi nist scale to determine if a 1/32 of an inch graduation could be distinguished and was found acceptable . The proposed alternative weld overlay repair and visual examination requirements will be implemented in a work order using the repair and replacement program requirements in Standard Programs and Processes (SSP)-9.1, Part D, "Repairs/Replacements of ASME Section XI Components." This repair and replacement program includes requirements for delineating the weld procedure and post weld heat treatment and nondestructive examination to be used after the repair per Paragraph IWA-4130(a) (3);

             "Inspection" per Subarticl e IWA-4140; "Material" per Subarticle IWA-4200; "Welding and Welde r Qualificati ons"
           ,per Subarticle IWA- 4400; and "Records" per Subarticle IWA-

4800. The design of the weld overlay repair and the

• safety evaluation per 10 CFR 50 . 59 , is documented in a

Design Chahge Notice (DCN) in accordance with SPP-9.3 ,

"Plant Modifications and Design Change Control." The nondestructive examination method which revea led the flaw and the description of the flaw , and a suitability evaluation of the repair meeting the requirements of Paragraphs IWA- 4130(a) (1) and (4) is considered within the DCN.

== Conclusion:== Based on the above di scussion, the alternati ve we ld overlay repair and visual examination provide an acceptable level of quality and safety. Authorization to implement the proposed alternatives is requested in accordance with 10CFR50.55a(a) (3) (i). E-5

ENCLOSURE 1 ATTACHMENT 1 FULL LENGTH CONTROL ROD DRIVE MECHANISM

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CAIL£ to*ECTIOI OPEUTIH COIL STAtl USEMll.Y ~ IYl IOD USEMtl l 0Isc:o**EtT 10 location of lower canopy seal weld ElAl-1

ENCLOSURE 1 ATTACHMENT 2 TYPICAL LOWER CANOPY SEAL WELD DETAIL ElA2 - 1

ENCLOSURE 1 ATTACHMENT 3 LOCATION OF CRDMs TO BE REPAIRED A p N M L I( J H-<f. 1eo* G F' E 0 c B A t5161Ji211I098 76S4 321 CRDMs with Lower Canopy Seal Weld Overlay Repairs CRDM # I Location 13 l G5 ElA3-l

ENCLOSURE 1 ATTACHMENT 4 PROPOSED WELD DESIGN E1A4 - l

From: Tsao, John Sent: 13 Sep 2016 12:36:14 +0000 To: Drake, James

Subject:

Accepted: Wolf Creek canopy seal leak repair

From: Tsao, John Sent: 13 Sep 2016 22:32:31 +0000 To: Alley, David;Hoffman, Keith;Kalikian, Roger

Subject:

another thing about the wolf creek clamps

Dave, At the end of phone call with Wolf Creek today, you asked that from the safety aspect whether permitting the existing 10 clamps to stay in place on the CROM housing at Wolf Creek is okay as opposed to removing the existing 10 clamps and repairing the degraded seal welds.

I think that permitting the clamps to stay in place provides more protection than removing the clamps and repairing the degraded seal welds. First, depending on the size of the existing mechanical clamps and how they are designed at Wolf Creek, a mechanical clamp provides a stronger support than a seal weld. Simply put, a clamp has more metal than a sea l weld. The clamps are similar to the strongbacks that can support lots of loads. A seal weld probably has only two weld passes. Secondly, the existing 10 clamps are installed on existing seal welds, albeit degraded some what. The degraded seal welds do provide some load support even though ASME Code does not take credit for that. A combination of a clamp on top of a degraded seal weld provides more protection than a new sound seal weld without the clamp. the on ly concern i have is whether Wolf Creek analyzed the weight of the mechanical clamp in terms of earthquakes. I do not know if the top of the CROM housing is supported. The bottom of the CROM housing/nozzle is supported at the RPV head. If the top of the CROM housing is supported also then i have no concern on the impact of the weight of the mechanical clamp on the CROM housing stresses due to a seismic event. If the top of the CROM housing is not supported then the CROM housing would be like a cantilever beam and the weight of the clamp may affect the stresses in the CROM housing.

From: Hoffman, Keith Sent: 15 Sep 2016 16:01:32 -0400 To: Alley, David

Subject:

NRR Position on the use of Canopy Seal Clamp Assemblies Attachments: 2008 10_15_08 Material Engineering Counterpart Call Summary.doc, 2008 10_15_08 Material Engineering Counterpart Call Summary - Attachment - CSCA.pdf As we all suspected this could not have been the first time the issue of Canopy Seal Clamp Assemblies (CSCA) has come up because we have found many instances of licensees that are using or have used them in the past. We have found that the issue came up back in 2008 and just as we did the NRC Staff struggled with whether the use of the CSCA was acceptable. The issue was discussed at a Materials Engineering Counterparts Call on 10/15/2008. The attached PDF file shows a document that describes the CSCA and its design and usage. The WORD file shows a summary of the discussion and a position on the use of the CSCA that was developed by Ted Sullivan. The decision they reached in 2008 was that using Appendix J of Section XI the use of the CSCA was a maintenance activity that did not require a Repair/Replacement Plan. Keith M. Hoffman Materials Engineer NRR/DE/EPNB {301)415-1294

SUMMARY

OF THE OCTOBER 15th 2008 MATERIALS ENGINEERING COUNTERPART CONFERENCE CALL On October 15th, 2008, the staff of the Division of Component Integrity of the Office of Nuclear Reactor Regulation (NRR) participated in a material's engineering conference call with staff from Regions 1, 2, 3, the Office of Nuclear Regulatory Research (RES), and the Office of New Reactors (NRO) regarding ongoing material issues and related activities at nuclear plants. In support of the call, the conference participant leads provided summary inputs, see below: CRDM Clamp Repair Method (CSCA) [George Hopper] R2 discussed the need to document the use of the CRDM clamps as an acceptable repair method for use on canopy seal weld leaks/flaws. This device has been used by the industry for a while and we had to reinvent the wheel regarding its use at Harris. A legacy document would be helpful for future encounters. Ted Sullivan will look into this. [Ted Sullivan] Canopy Seal Clamp Assemblies (CSCA) In August a leak was found in a control rod drive mechanism (CRDM) canopy seal welld at Shearon Harris. Harris personnel decided to address the leak by installing a mechanical canopy seal assembly. During the October 2008 Counterparts phone call , Region II asked NRR staff to document its position with respect to installation of CSCAs on CRDMs. The write-up below is in response to that request. Although this is not a new method for addressing leaks through CRDM canopy seal welds, the NRC staff addressing this issue did not have prior experience with this application and raised questions in the areas of whether this application was a non-code repair that needed relief from ASME code requirements and whether the licensee would have to do anything about the actual Class 1 leaking boundary (i.e., the threaded connection). The following represents a summary of the basis provided to the staff regarding the questions raised.

• A threaded connection is a pressure retaining component, but because no stru ctural credit was taken for the seal welds in the design of the CRDMs, the CRDM seal welds were not designed to be a Code component.
• The decision process presented in Non-Mandatory Appendix J of Section XI , "Guide to Plant Maintenance Activities and Section XI Repair/Replacement Activities," leads to the conclusion that installing the clamp assembly is not an activity within the scope of the repair/replacement requirements of Section XI , IWA -4000. The installation of a CSCA does not affect the pressure retaining portion of a code item and the activity comes under maintenance. This activity does not affect tests or examinations or records of tests or examinations that would be required under Section XI.
• The Engineering Change process requires the licensee to ensure that under this activity they continue to meet their design basis, which is the ASME Code, Section Ill. Harris personnel ensured that they met the requirements of Section Ill, NB-3671 .3 for threaded joints.
• AN ll involvement is not required , although Harris provided the AN ll with a courtesy review. No NIS-2 form was prepared.
• Harris conducted a VT-2 leakage examination prior to returning the plant to service, in accordance with the normal requirements of IWA-5000.

The staff concluded that this basis for application of the CSCA at Harris was reasonable. In addition, the staff searched on this topic in ADAMS. While a number of documents discussing CSCAs were found, no relief requests on this topic were identified .

Attachment:

Westinghouse Nuclear Services/Engineering Services Canopy Seal Clamp Assembly (CSCATM) Brochure Catawba Repair Issue R2 continues to follow-up on the Catawba service water pipe repair issue. The licensee used a wooden plug to stop the leak when they did the repair by modification. They welded a pipe with cap over the hole and left the plug in. Therefore, no pressure test was performed, since there was no way to verify the plug dislodged or allowed water to pass. The ANll refused to sign off on the repair. Time Requirements RIS-2005-20 R2 discussed the frustration with dealing with the expectations regarding time requirements for prompt determinations of operability (RIS-2005-20) and what is actually occurringcr at the sites. Farley took over 10 days to evaluate the structural integrity of a leaking service water pipe and finalize the prompt determination of operability to justify continued operation. We may want to look at gathering data on how long these are taking in each region and then move to change the wording in the RIS, or communicate the expectation via another document. DC Cook Turbine/Fire Suppression System Failure On September 20, 2008 at DC Cook Unit 1, experienced a turbine failure and generator hydrogen fire. Many blades in the low pressure side of the turbine were damaged or failed. Additionally, the fire protection header outside the turbine building ruptured shortly after the turbine failure causing loss of the fire protection system and severely damaged two of the 3 fire pumps due to continued pump operation without water in the system. Region Ill is conducting a special inspection to investigate this event. Perry Re-Review of UT Data At Perry, during the licensee re-review of UT data on the N6A and N6C RHR nozzle-to-SE dissimilar metal welds, the licensee identified Code rejectable flaws. Licensee is completing ASME Code Section XI IWB-3600 flaw evaluation to accept welds for continued service (Plant was at power during this review and remained at power). Perry High Cycle Fatigue Failure At Perry, licensee experienced a high cycle fatigue failure causing a thru-wall leak on a 6" reactor water cleanup system line just downstream of the rejenative HX. Licensee maintained this non-safety related system in operation for more than 5 days without providing a basis for functionality. No NRC policy in this area. FFS Fitness-For-Service API 579-1/ASME FFS-1, JUNE 5, 2007

The subject recommended practice is a compendium of analysis techniques that can be applied to inform run/repair/replace decisions for pressurized components. The document was adapted from an earlier API petrochemical industry document and it borrows liberally from a wide variety of pressure vessel codes and standards. It is currently published under a committee that includes ASME sponsorship, so it is an ASME recommended practice. It may be accessed through the IHS Codes and Standards link on the NRC website under the information resources page. In order to find it in IHS, you would search for the term "FFS-1." The document addresses problems of a general nature and warns users that particular circumstances require review of applicable codes, standards and regulations. The applicability section states:

         "The assessment procedures in this Standard can be used for Fitness-For-Service assessments and/or rerating of equipment designed and constructed to the following codes:

a) ASME 8&PV Code, Section VIII, Division 1 b) ASME 8&PV Code, Section VIII, Division 2 c) ASME 8&PV Code, Section I d) ASME 831 .1 Piping Code e) ASME 831.3 Piping Code f) AP/ 650 g) AP/ 620" This document is over 1100 pages long and it includes a large number of fairly detailed approaches for assessing susceptibility to, or damage caused by, common degradation mechanisms. It is arranged by degradation mechanism, so there is a chapter on addressing brittle fracture, one on pitting, and chapters on creep damage, fire damage, general wastage, weld misalignment and shell distortions, laminations, crack-like flaws, gouges, etc. Each chapter includes some standard assessment techniques, describes acceptable general approaches and includes acceptance criteria. Overall, the document provides a wide variety of useful, standard engineering tools and approaches for making run/repair/replace assessments. We have not reviewed this document. It is not clear that licensees could use it to assess systems, structures or components that are designed to Codes or Standards that are not included in the applicability section. It is possible that a licensee could use the guidance in FFS-1 to assess non-safety related system components, piping systems constructed to B31.1 , or for performing functionality (not operability) assessments of degraded ASME components. The example we discussed during the Counterparts Call was the Comanche Peak hot leg boric acid corrosion assessment. The licensee could have elected to use the guidance in the "Assessment of General Metal Loss" chapter to determine that the reactor coolant system remained functional (or not) before they completed the actions required to make their operability determination. They are not constrained to use the ASME Code for a functionality determination. We have not yet collectively discussed or decided how this document would fit ointo our regulatory framework.

Nuclear Services/Engineering Services Canopy Seal Clamp Assembly (CSCA') Background Description Small leaks in PWR head penetrations can The canopy seal is a weld between the reactor vessel prevent a return to power and cause head control rod drive mechanism (CROM ) expensive delays until a fix is devised. An penetration and the mating part. This weld has a increasing number of plants are reporting tendency to develop cracks as a result of stress primary coolant leaks in the field-welded corrosion cracking (SCC) and/or original weld canopy seal area. Westinghouse offers a full defects. These cracks spread through the walls until there's leakage. range of products and services to control these kinds of leaks, including a unique The Westinghouse CSCA provides a non-welded mechanical clamp assembly, the CSCA. mechanical method of stopping leaks in the canopy seal weld. The CSCA seals the leaking weld and introduces a compressive load into the canopy seal, which tends to close and arrest the crack propagation. The CSCA seals the leaking weld by compressing a graphite seal over the entire canopy seal weld area. The CSCA consists of a housing, a top plate, seal carrier halves, a split graphite seal, cap screws, and Belleville washers. The general arrangement is the same for spare capped CROM penetrations, full-length active CRDM penetrations, and for core exit therrnocoup[e (CET) CROM penetrations. 'fo be repaired, the housing is lowered over the penetration, b elow the elevation of the canopy seal weld. The first seal carrier half with graphite is lowered and placed into the housLng; the second seal carrier half is installed the same way. The (.)Westinghouse

housing is then raised until the graphite seal contacts At several plants, the installations were done on an the canopy seal and the top plate, and the cap screws emergency basis. Usually, in this situation, t he CSCAs with the Belleville washers are installed to provide the are installed and the leak is eliminated within four to necessary loading forces. five days. after notification by the utility. Trained Westinghouse crews will implement the Benefits installations with special tooling designed for this unique The average radiation dose for all installations was application. We provide full engineering and support Jess than 300 millirem per CSCA, comparing favo r-services, including design specification, stress report, and ably with about l.5 man-rem per penetration for certifications to Section III of the ASME Code. weld overlay repairs, and up to 10 to 20 man-rem per penetration for other "cut-and-cap" or welding optio ns. CSCA installation can be done without draindown of the reactor cooling system (RCS)-a distinct advantage over weld repair options. A leak at o ne plant wasn't discovered until the outage was over and the unit was returned to power; the CSCA was installed with the RCS in Mode 5 at abo ut 300 psi and at l 70°F. At another plant, the CSCA was installed during a system heat-up to about 200°F.

• The CSCA works o n leaking or non-leaking canopy seal welds, and can also be modified for installation on previously overlay-repaired welds. Applying the CSCA on a non -leaking seal weld is a good preventative measure against continued sec. which is a primary cause of failure.

From: Taylor, Nick Sent: 15 Sep 2016 14:48:20 -0500 To: Alley, David;Anchondo, Isaac; Drake, James;Werner, Greg Cc: Graves, Samuel;Dodson, Douglas;Thomas, Fabian

Subject:

RE: FW: NRC Questions regarding Penetration 77

All, FYI - I just got off the phone with the EDO's office, who had called to ask about the "ASME code non-compliance" and "improper repairs to the vessel head" at Wolf Creek (as relayed to them by our regional management after the morning meeting here yesterday I think).

I let them know the following: That there is no cu rrent safety issue given that the plant is shut down and won't see power operation for at least 2 months, and that we will have ample opportunity to inspect the head and the licensee's actions before restart Regarding the code issue - I let them know that the staff is still having dialogue about what the code requires for these penetration leaks, and that we are still waiting for the licensee to answer some questions before we determine whether or not a compliance issue exists Please let me know if I've communicated anything in error. I look forward to more discussion on this as the facts become more clear. DRP management has asked that we provide status periodically as the issue develop.

Thanks, Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV 0:

C: (8!7) l(b)( 200-1141 I E: nick.taylor@nrc.gov I R From : Alley, David Sent: Thursday, September 15, 2016 2:12 PM To: Anchondo, Isaac <lsaac.An chondo@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Taylor, Nick

<Nick.Taylor@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Cc: Graves, Samuel <Samuel.Graves@nrc.gov>

Subject:

RE: FW: NRC Questions regarding Penetration 77

No apology necessary. Your focus is exactly in the right spot which is the safety significant spot. Dave From : Anchondo, Isaac Sent: Thursday, September 15, 2016 3:07 PM To: Alley, David <David.Alley@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Taylor, Nick

   <Nick.Taylor@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Cc: Graves, Samuel <Samuel.Graves@nrc.gov> Subje ct: RE: FW: NRC Questions regarding Penetration 77 Dave, I apologize for forgetting to acknowledge that the question was already out there. I was just trying to stress that our regulations, intent of the code, etc, point to ade uate ressure retainin capabilities which is the threaded 'oint not the seal weld. (b)(5) (b)(5) Look forward to Keith's conclusion! Isaac From : Alley, David Sent: Thursday, September 15, 2016 1:52 PM To: Anchondo, Isaac <lsaac.An chondo@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Taylor, Nick

   <Nick.Taylor@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Cc: Graves, Samuel <Samuel.Graves@nrc.gov> Subje ct: RE: FW: NRC Questions regarding Penetration 77 All, Keith Hoffman is working diligently to come to a conclusion regarding our opinion on the code compliance of the clamp. He may get done this PM. We probably will still want to have the licensee go through their basis for code compliance, irrespective of Keith's findings. Isaac, I can't remember whether you were on the phone call last Saturday. If so you may remember that I asked them about their basis, given the amount of leakage, for saying that the threads were ok. In my opinion we have already asked the question that you wish to pursue and that we absolutely should follow up on that question. At this point, I am not proposing that the threads are bad, only that it is a worthwhile question. Dave From : Anchondo, Isaac Sent: Thursday, September 15, 2016 2:39 PM To: Drake, James <James.Drake@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Alley, David

   <David.Alley@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Cc: Graves, Samuel <Samuel.Graves@nrc.gov>

Subject:

RE: FW: NRC Questions regarding Penetration 77

(b)(5) For reference, here's the technical rationale (in part) of NUREG-0800 in regards to threaded fasteners (Class 1, 2, and 3) and therefore a light on the intent of mechanical connections. GDCs 1 and 30 require that SSCs important to safety be designed, fabricated, erected, tested and inspected to quality standards commensurate with the importance of the safety functions to be performed. GDC 14 requires that the RCPB be designed, fabricated, erected, and tested in a manner that provides assurance of an extremely low probability of abnormal leakage, rapidly propagating failure, or gross rupture. The RCPB, provides a barrier to fission products, a confined volume for the inventory of reactor coolant, and flow paths to facilitate core cooling. Threaded fasteners and mechanical joints form an integral part of maintaining pressure boundary integrity and are essential for withstanding normal loading and any transient load created during abnormal or accident conditions. The failure of fasteners in a system could result in loss of fluid in the system and jeopardize safe operation of the plant. Conformance with criteria of the ASME Code, Section Ill and the regulatory positions of RG 1.65 satisfies, in part, the requirements of GDC 1, 14, and 30 by providing assurance that threaded fasteners will be designed, fabricated, and tested to established and proven standards and, thereby, minimizing the likelihood of failure of the pressure boundary. GDC 31 requires that the RCPB be designed with sufficient margin to ensure that when stressed under operating, maintenance, testing, and postulated accident conditions the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized. 10 CFR Part 50, Appendix G establishes fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to ensure that there are adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Threaded fasteners and mechanical joints are integral to the design of the RCPB. Application of the requirements of Appendix G ensures that threaded fasteners in the RCPB will behave in a nonbrittle manner, minimizing the probability of rapidly propagating fracture and thereby satisfying the requirements of GDC 31. I agree with having a call with the licensee, and in addition to Jim's points, we would have to get a clarification on the intent of the CSCA as far as pressure retaining function.

From: Drake, James Sent: Thursday, September 15, 2016 11:46 AM To: Taylor, Nick <Nick.Taylor@ nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Alley, David

<David.Alley@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Subject:

RE: FW: NRC Questions regarding Penetration 77

Nick, Right now we do not have enough information to come to a more aligned regulatory position.

There are several potential approaches on the issue and there may be other documents out there that we have not found. I think we need to have the meeting with the licensee to have them explain in detail how they determined that the CSCA's are Code compliant. Once we have that information, we can eva luate and come to a regulatory position. The CSCA's do not appear to be a safety issue, they are designed to Class 1 standards, they have the required strength, and we are not aware of any problems with leakage from the clamps. Westinghouse completed the stress analysis and there is no problem. However, we have not verified the results. Where we are currently at is: Is the use of Canopy Seal Clamp Assemblies allowable by Code and has Wolf Creek complied with all regulatory requirements when they installed them. Until we have Wolf Creek's position on the CSCA and all associated documents, we will be making assumptions and won't be able to make an informed decision. Jim From: Taylor, Nick Se nt: Thursday, September 15, 201611:17 AM To: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Alley, David

<David.Alley@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Subje ct: RE: FW: NRC Questions regarding Penetration 77

All, I would like to see us have a meeting to get more aligned on code applicability, etc prior to engaging with the licensee or having another substantial discussion at the morning meeting.

This issue has come up 3 days in a row now at the morning meeting, and there are a llot of opinions out there on what the code requires, but not a lot of facts from the licensee. I'd like to see us all on the same page prior to communicating with management or the licensee on whether or not the licensee improperly repaired the head, etc.

Thanks, Nick From: Anchondo, Isaac Se nt: Thursday, September 15, 2016 10:56 AM To: Drake, James <James.Drake@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Werner, Greg

<Greg.Werner@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>

Cc: Lyon, Fred <Fred .Lyon@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Tsao, John

  <John.Tsao@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Proulx, David
  <David.Proulx@nrc.gov>; Thomas, Fabian <Fabian .Thomas@nrc.gov>; Kopriva, Ron
  <Ron. Kopriva@n re.gov>

Subje ct: RE: FW: NRC Questions regarding Penetration 77 All, I would like to suggest coming up with an agreeable regulatory path as far as how we are interpreting this issue (i.e, ASME vs TS vs CAP, etc). The end game will have to be whether we agree that the licensee can use the CSCA, and if so, do they need relief to do so. (b)(5) Criterion 14-Reactor coolant pressure boundary. The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and o f gross rupture. I would suggest approaching this issue in terms of corrective actions rather than simply ASME Code compliance. Previous to this leak, the licensee opted to perform a "basic cause evaluation" as part of the approval to install the CSCA. Two statements caught my attention:

          "There have been no industry reports of degradation of canopy seal welds resulting in significant leakage flow rates (Ref. 3). Considering the head adapter flange design, leakage through a crack in the non-pressure boundary seal weld would be expected to be limited by the load carrying component, the flange connection threads."
          'T he Westinghouse hardware failure analysis also included examination of some threaded joints that were removed along with the lower canopy seal welds. There was no evidence of corrosion or cracking on any of the threaded joints that were examined."

(b)(5)

I"' Any comments are welcome. Thanks, Reactor Inspector U.S. Nuclear Regulatory Commission I Region IV Division of Reactor Safety I Engineering Branch 2 (817) 200- 1152 From: Drake, James Sent: Thursday, September 15, 2016 7:59 AM To: Alley, David <David.Alley@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Taylor, Nick

  <Nick.Taylor@nrc.gov>

Cc: Lyon, Fred <Fred .Lyon@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Tsao, John

  <John.Tsao@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Dodson, Douglas
  <Douglas.Dodson@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Thomas, Fabian
  <Fabian.Thomas@nrc.gov>; Kopriva, Ron <Ron .Kopriva@nrc.gov>

Subject:

FW: FW: NRC Questions regarding Penetration 77 Importance: High Interesting responses From: M uilenburg Will iam T [1] Sent: Thursday, September 15, 2016 7:53 AM To: Drake, James <James.Drake@nrc.gov>

Subject:

[External_Sender) FW: NRC Questions regarding Penetration 77 Importance: High Jim, Below are the answers I got to the questions you and Isaac gave me earlier. I've attached the CR associated with the final question as well. Bill Questions from 9/12 phone cal l - Everyone, NRC Inspectors Jim Drake and Isaac Anchondo called this morning with the following questions related to our vessel head and penetration 77. From Jim Drake

1. What code was used to construct the head, B31.7 or Section Ill? If Section Ill, what year?

ASME Section Ill, 1971 Edition through Winter 1972 Addenda

2. What Code, for ISi, is Wolf Creek currently committed to?

ASME Section XI 2007 Edition through 2008 Addenda

3. What Class of piping is this penetration?

The vessel penetration is ASME Class 1. The CRDM housing is a Class 1 component; there is no piping involved. The pressure boundary connection is a threaded mechanical connection with a non-pressure boundary seal weld outside of the threaded connection.

4. Is Code Case N-733 applicable to this condition?

No, Code Case N-733 is not applicable because this Code Case is applicable to vessel penetration welds and the leak is on the seal weld of the threaded mechanical connection. From Isaac Anchondo

1. It is noted that there are 10 other penetrations that have these repairs made, was Code Case N-733 applied to those efforts?

No, see above response. The canopy seal clamp assembly was a mod ification designed to ASME Section Ill requirements, so no Code Case was needed for the clamp assembly. The clamp assembly was designed to ASME Section Il l to assure the stresses in the clamp assembly and the RV and CRDM threaded connect ions as a result of applying the clamp assembly, did not exceed those allowed for Class 1 components, not because it was sealing a leak of the non-pressure boundary canopy seal weld.

2. Is there any root cause/apparent cause documents associated with these previous repairs?

CR 93697, HFAR MA 92-008, WCAP 12088, MED-PCE-11788 Please let me know when the answers to any of these are available so that I can provide a response to the NRC as quickly as possible.

Thanks, Bill Muilenburg, ext. 4511

From: Tsao, John Sent: 15 Sep 2016 18:14:48 +0000 To: Alley, David;Hoffman, Keith;Kalikian, Roger Cc: Hsu, Kaihwa;li, Yong

Subject:

RE: Mechanical Clamp in ASME Section Ill I think that we should call the contraption installed on the canopy seal at Wolf Creek as a "mechanical joint", not as a "mechanical clamp". This is because NB-3671.7 permits the installation of mechanical joints (see Keith's email below). NB-3671.7 Sleeve Coupled and Other Patented Joints. Mechanical joints, for which no standards exist, and other patented joints may be used provided the requirements of (a), (b), and (c) below are met. (a) Provision is made to prevent separation of the joints under all Service Loadings. (b) They arc accessible for maintenance, removal, and replacement after service. (c) Either of the following two criteria are met. ( I) A prototype joint has been s ubjected to performance tests to determine the safoty of the joint under simulated service conditions. When vibration, fatigue, cyclic conditions, low temperature, thermal expansion, or hydraulic shock is anticipated, the applicable conditions shall be incorporated in the tests. T he mechanical joints shall be sufficiently leak tight to satisfy the requirements of the Design Specifications. (2) Joints are designed in accordance with the rules of NB-3200. A "mechanical clamp" as per ASME Section XI, Appendix IX or Appendix W, is not permitted to be installed on Class 1 piping and has a limited service time period (to the next refueling outage). From: Alley, David Sent: Thursday, September 15, 2016 1:26 PM To: Hoffman, Keith <Keith.Hoffman@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Cc: Hsu, Kaihwa <Kaihwa.Hsu@nrc.gov>; Li, Yong <Yong.li@nrc.gov>

Subject:

FW: M echanica l Clamp in ASME Section Ill Based on Robert's view, below, it appears that the clamp is acceptable per the construction code. This would appear to make the use of the clamp a code repair as it is in accordance with the construction code. This would appear to mean that the plant can install and leave the clamps on forever and that we have no regulatory hook (other than, potentially, the condition of the threads for this instance based on the extent of leakage). Any thoughts? Dave From: Hsu, Kaihwa Sent: Thursday, September 15, 2016 8:25 AM To: Alley, David <David.Alley@nrc.gov> Cc: Li, Yong <Yong.Li@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill Dave: I don't see any problem for the patented clamp to be used over top of original joint as long as the repair meets ASME Section Ill Code criteria.

Robert From: Li, Yong Sent: Thursday, September 15, 2016 7:29 AM To: Hsu, Kaihwa <Kaihwa.Hsu@nrc.gov>

Subject:

FW: Mechanical Clamp in ASME Section Ill Please respond to Dave. From: Alley, David Sent: Wednesday, September 14, 2016 8:19 PM To: Hoffman, Keith <Keith .Hoffman@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>; Li, Yong <Yong.Li@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill

Keith, Very wise. When you don't know the answer, make is someone else's problem

Yong, As Keith points out below, we could use some section Ill help. NB-3671.7 Sleeve Coupled and Other Patented J oints says that you can use sleeve coupled and other patented joints. Doesn't seem like there is much rigor in their qualification. That aside, the real question is that we have an instance where Wolf Creek appears to be installing a mechanical clamp over a threaded joint/canopy seal based on this code paragraph. It is pretty apparent to me that the patented joint could be used in place of the threaded connection/seal weld. It is not quite so apparent that such a joint is permitted to be used over the top of another type of joint. Your thoughts?

Dave From: Hoffman, Keith Sent: Wednesday, September 14, 2016 8:08 PM To: Alley, David <David.Alley@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill I believe that is probably a question for Yong Li's branch. Obviously the licensees and ABB - CE/Westinghouse believe it can be and that is what the patent says it was designed to do. From : Alley, David Sent: Wednesday, September 14, 2016 4:19 PM To: Hoffman, Keith <Keith.Hoffman@nrc.gov>; Tsao, John <John .Tsao@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill 3671.7 seems to allow almost joint that has had a mockup made and tested. However, it is in a section for nonwelded pipe joints. In my mind this would allow such a joint in place of a threaded joint. Does it allow the application of such a joint over an existing joint?

Dave From : Hoffman, Keith Sent: Wednesday, September 14, 2016 6:39 AM To: Tsao, John <John .Tsao@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

RE: Mechanical Clamp in ASME Section Ill This was the section that was referenced as the applicable Section Ill paragraph in one of the documents I looked at yesterday. Specifically NB-3671.7 was referenced. NB-3671.3 shows the threaded joint and the requirement for the weld. NB-3670 SPECIAL PIPING R EQUI REMENTS NB-3671 Selection and Limitat ion of Nonwelded Piping Joints The type of piping joint used shall be suitable for the Design Loadings and shall be selected with consideration of joint tightness, mechanical strength, and the nature of the fluid handled. Pipingjoints shall conform to the requirements of this Subsection with leak tightness being a consideration in selection and design of joints for piping systems to satisfy the requirements of the Design Specifications. NB-3671.1 Flanged J oints. Flanged joints are permitted. NB-3671.2 Expanded Joints. Expanded joints shall not be used. NB-3671.3 T hreaded Joints. Threaded joints in which the threads provide the only seal shall not be used. If a seal weld is employed as the sealing medium, the stress analysis of the joint must inc.ludc the stresses in the weld resulting from the relative deflections of the mated parts. NB-3671.4 F lared, Flareless, and Compression Joints. Flared, flareless, and compression type tubing fittings may be used for tubing sizes not exceeding 1 in. 0.D. (25 mm) within the limitations of applicable standards and specifications listed in Table NCA-7100- 1 and requirements (b) and (c) below. In the absence of such standards or specifications, the Designer shall determine that the type of fitting selected is adequate and safe for the Design Loadings in accordance with the requirements of(a), (b), and (c) below. (a) The pressure design shall meet the requirements ofNB-3649. (b) Fittings and their joints shall be suitable for the tubing with which they are to be used in accordance with the minimum wall thickness of the tubing and method of assembly recommended by the manufacturer. (c) Fittings shall not be used in services that exceed the manufacturer's maximum pressure-temperature recommendations. NB-3671.S Caulked Joints. Caulked or leaded joints shall not be used. NB-3671.6 Brazed and Soldered Joints. (a) Brazed Joints ( 1) Brazed joints of a maximum nominal pipe size of I in. (DN 25) may be used only at dead end instrument connections and in special applications where space and geometry conditions prevent the use of joints permitted under NB-3661.2, NB-3661.3, and NB-3671.4. The depth of socket shall be at least equal to that required for socket welding fittings and shall be of sufficient depth to develop a rupture strength equal to that of the pipe at Design Temperature (N B-4500). (2) Brazed joints that depend upon a fillet rather than a capillary type filler addition arc not acceptable. (3) Brazed joints shall not be used in systems containing flammable fluids or in areas where fire hazards are involved. (b) Soldered Joints. Soldered j oints s hall not be used. NB-3671.7 Sleeve Coupled and Other Patented Joints. Mechanical joints, for which no standards exist, and other patented joints may be used provided the requirements of(a), (b), and (c) below are met. (a) Provision is made to prevent separation of the joints under all Service Loadings. (b) They are accessible for maintenance, removal, and replacement after service. (c) Either of the following two criter ia are met. (1) A prototype joint has been subjected to performance tests to determine the safety of the joint under simulated service conditions. When vibration, fatigue, cyclic conditions, low temperature, thermal expansion, or hydraulic shock is anticipated, the applicable conditions shall be incorporated in the tests. The mechanical joints shall be sufficiently leak tight to satisfy the requirements of the Design Specifications. (2) Joints are designed in accordance w ith the rules of NB-3200. Keith M. Hoffman Materials Engineer

NRR/DE/EPNB (301)415-1294 From: Tsao, John Sent: Tuesday, September 13, 2016 6:00 PM To: Alley, David <David.Alley@nrc.gov>; Hoffman, Keith <Keith.Hoffman@nrc.gov>; Kalikian, Roger

<Roger.Kalikian@nrc.gov>

Subject:

Mechanical Clamp in ASME Section Ill I did a word search of "Clamp" in NB and NC sections of the 2007 ed ition and 2013 edition of the ASME Code, Section Ill. There were 4 hits in both ed itions. NB-1132.1- the clamp in this article is related to pipe attachment (the clamp used for pipe supports) NB-3411.l(d) pump clamp NB-3651.3 pipe clamp as in pipe supports such as hangers or snubbers that use clamps. NB-4231---clamps used in welding operations (when welding 2 pieces of pipe, welders use clamps) So ASME SEction Ill does not have requirements or specification for the mechan ical clamp application that was used on the CRDM canopy sea l at Wolf Creek.

From: Drake, James Sent: 19 Sep 2016 11:27:05 -0500 To: Alley, David

Subject:

RE: Response to questions on Wolf Creek penetration Thank you Dave. Isaac is running with the issue for now. I just got to Grand Gulf to start a heat exchanger inspection. Jim From: Alley, David Se nt: Sunday, September 18, 2016 7:17 PM To: Drake, James <James.Drake@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Taylor, Nick

<Nick.Taylor@nrc.gov>

Cc: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <D.avid.Proulx@nrc.gov>; Lyon, Fred

<Fred.Lyon@nrc.gov>

Subje ct: RE: Response to questions on Wolf Creek penetration Jim For question 1. I buy their argument about the lack of steam cutting. If they can pass over 3 GPM through threads in good condition , given that we were less than 3 GPM, they appear to have given us a reasonable basis to not require further inspection of the threads immaterial of past OpE. The concept that it should be possible to pass more than 3 GPM through good threads seems excessive to me - not sure why - it just does. One option would be to stop here and declare victory based on the Westinghouse work and their explanation of it. The other option would be to review the Westinghouse work. Not sure what that would accomplish (other than completeness) as it should be a reasonably straight forward calculation. For question 3 we could quit here as we have figured out that they are OK in code space for two different reasons or we could go back to them and say that there answer is insufficient (which it is) and that we want them to tell us by which paragraphs of the code the clamps were analyzed and by which paragraphs of the code their use is permitted. Absolutely nothing to be gained with respect to safety by putting them through this exercise, however, the concept of letting them think that the answer that they provided was acceptable is somewhat unpalatable. This is just my two cents worth. You note that I haven't actually given you a straight answer. That is intentional. These seem to me to be decisions that fall within the Region's purview. I will support whatever you folks decide. Dave From: Drake, James Se nt: Saturday, September 17, 201610:33 PM To: Werner, Greg <Greg.Werner@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Taylor, Nick

<Nick.Taylor@nrc.gov>

Cc: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Lyon, Fred

<Fred.Lyon@nrc.gov>

Subject:

Fw: Response to questions on Wolf Creek penetration Here is the reply to the questions we asked the licensee. Jim James F. Drake Email: James.Drake@nrc.gov Office phone:817-276-6558 Cell phone:l(b)(6) I From: Muilenburg William T <wimuile@WCNOC.com> Sent: Friday, September 16, 2016 5:59 PM To: Drake, James

Subject:

[External_Sender] Response to questions on Wolf Creek penetration

Jim, Here are the answers I got on your questions last week.

We could have a call with you at either 10 or 11:00 AM on Monday 9/19. Let me know what works for you please and I'll confirm.

Thanks, Bill

1. How are we verifying the structural integrity of the joint? His interest here is increased by the size of this leak.

No verification of the structural integrity of the joint is required. Westinghouse has calculated that the maximum leakage flow for one canopy seal is about 3.5 gpm. The observed leakage was less than the maximum value. The design of the mechanical connection is that the canopy seal weld is a specially designed seal between the housing (i.e. Control Rod Drive Mechanism (CROM), head adapter plug or CET) and the reactor vessel head adapter flange. The sole function of the canopy seal and seal weld is to provide RSC leakage control. The threaded connection between the adapter flange and the housing, independent of the canopy seal, provides the structural integrity for the pressure boundary items of the connection under all service loadings. With the failed seal weld, the leakage does not affect the threaded connection since the mechanical connection is pressurized by the RCS and leakage past the threads is not a failure of the pressure boundary. With the RCS at normal operating temperature and pressure, water in and around the threads are essentially at the same pressure as the RCS and the leakage from the failed weld flashes to steam once beyond the outer surface of the canopy seal (or across the flaw). In this condition, the water does not flash to steam until the failed surface or beyond so there is no steam cutting of the threads. Therefore, no impact on the structural integrity of the joint will occur.

2. What is our plan to repair the penetration?

Install a canopy seal clamp on the leaking penetration.

3. If we intend to use the canopy seal clamp again, what is our basis and code that we intend to apply?

The CROM Seal Clamp Assembly is analyzed to ASME B&PV Code, Section 111, Division 1, 1986 Edition (No addenda). The Design Specification for the CROM Seal Clamp Assembly is cert ified to ASME B&PV Code, Section Ill, Division 1, 1971 Ed. up to and including the Winter 1972 Addenda and the 1974 Ed. The Design Report for the CROM Seal Clamp Assembly is cert ified to ASME B&PV Code, Sect ion Ill, Division 1, 1986 Edition (No addenda).

4. Comment on Head Inspection. Jim urged us to use a forensic approach to examining and cleaning the head. He indicated that Ft. Ca lhoun had had a similar problem and through power washing the head destroyed any evidence that could have contributed to analysis of the defect.

Mark Barraclough is aware of the need for this as he is considering the impact on his programmatic inspections.

From: Ross-Lee, MaryJane Sent: 20 Sep 2016 15:19:03 -0400 To: Alley, David;McHale, John

Subject:

FW: Wolf Creek Status - 09/20/16 Attachments: DSCF3798.jpg, DSCF3797.jpg, DSCF3792.jpg, DSCF3789.jpg, DSCF3788.jpg, DSCF3786.jpg Mary Jone Ross-Lee (MJ) Deputy Director, Division of Engineering Off ice of Nuclear Reactor Regulation OWFN 9Hl US Nuclear Regulatory Commission ~ Office: 301-415-3298

1 e-mail: maryjane.ross-lee@nrc.gov From : W ilson, George Sent: Tuesday, September 20, 2016 1:08 PM To: Dean, Bill <Bill.Dean@nrc.gov>; McDermott, Brian <Brian.McDermott@nrc.gov>; Evans, Michele

<Michele.Evans@nrc.gov>

Cc: Boland, Anne <Anne.Boland@nrc.gov>; Benner, Eric <Eric.Benner@nrc.gov>; Lubinski, John

<John.Lubinski@nrc.gov>; Ross-Lee, MaryJane <MaryJane.Ross-Lee@nrc.gov>

Subject:

FW: Wolf Creek Status - 09/20/16 See the attached pictures of the uncleaned RVH at WCGS, it does show some degradation. RIV sent these to the EDO coordinator. Outage information from RIV The licensee's RPV Visual Inspection procedure/activity is scheduled to complete at - 9:00 PM tonight (09/20/16) according to the Outage Schedule in the OCC. However, based on the delayed start of this activity on yesterday, I am almost positive that this will slip by at least 24 hours. Based on my conversation with the dayshift OCC Manager, the head condition will not be fully assessed until the RPV inspection activity is completed. George Wilson Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation USN RC 301-415-1711 Office 08E4 From : Lyon, Fred Sent: Tuesday, September 20, 2016 10:02 AM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>

Cc: Wilson, George <George.Wilson@nrc.gov>

Subject:

FW: Wolf Creek Status - 09/20/16 See the attached pictures of the uncleaned RVH at WCGS. RIV sent these to the EDO coordinator, so if questions come down, no, it's not D-B. From : Thomas, Fabian Sent: Tuesday, September 20, 2016 9:45 AM To: Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Lyon, Fred

<Fred.Lyon@nrc.gov>; Janicki, Steven <Steven.Janicki@nrc.gov>

Cc: Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

Subject:

Wolf Creek Status - 09/20/16 Wolf Creek Status for September 20, 2016

• Outage Parameters:

Mode-5 RCS Temperature -Approximately 100 F RCS Pressure - Depressurized (<1 psig) Inventory - 607 .6 inches -RCS Drain-down still in progress RCS Time to Boil: -45 minutes SFP Time to Boil: -67 hours Containment Status - Open Fuel Moves - None in progress Upcoming Activities: Decontamination of RX Vessel Head (ongoing) Moving equipment into containment (ongoing) RCS drain down to the flange (09/21/16 on dayshift) Mode 6 is scheduled for 09/21/16 & Stud detensioning (02:00); According to OCC Manager, this is not expected until 09/22/16

• Plant Shutdown Risks: All Green Reactivity Management - Green Core Decay Heat Removal - Green Spent Fuel Decay Heat Removal - Green RCS Inventory - Green Electrical Power Sources - Green Containment Closure - Green Radiation Monitoring and Ventilation - Green Attached are pictures of the east side of the reactor head vessel, taken last night after the removal of the blanket (NUCON) insulation and side mirror insulation.

The licensee's RPV Visual Inspection procedure/activity is scheduled to complete at -9:00 PM tonight (09/20/16) according to the Outage Schedule in the OCC. However, based on the delayed start of this activity on yesterday, I am almost positive that this will slip by at least 24 hours. Based on my conversation with the dayshift OCC Manager, the head condition will not be fully assessed until the RPV inspection activity is completed.

Also, Mode 6 has slipped due to the aforementioned delays in the insulation removal and issues encountered with the Large Equipment Lift outside of containment (delaying placement of head detensioning equipment). Mode 6 is scheduled for 02:00 on 09/21/16, but the OCC Manager is sure that it will slip to 09/22/16.

• TS LCOs: None.
• Other Work Activities:

The Rose Hill transmission line (one of the three 345 KV transmission lines) will be out of service until 09/23/16 for breaker testing, which will be conducted in the switchyard. The residents have reviewed the outage risk assessment for this condition-in accordance with the risk process Wolf Creek will be implementing risk mitigation actions when work is occurring in the switchyard (supervisory oversight) to maintain "Electrical Power Sources" risk green during this evolution.

• Items of Interest:
• Other Inspections/Audits: Ron Kopriva expected to arrive onsite today.
• Significant Forecasted Weather: None
• Coverage and Other Visits: Fabian is onsite, and in the area. Doug is in the region for Regulatory Conference, returning to the office on 09/22/16.

0 c

From: Tsao, John Sent: 4 Sep 2016 15:12:18 +0000 To: Alley, David

Subject:

Re: Status 3 PM Saturday 9/3: TS Shutdown at Wolf Creek (RCS leakage TS 3.4.13)

Dave, thanks. I am glad that the phone ca ll with Wolf Creek went well. I am working on my appraisals today and hopefully wi ll finish it by tomorrow.

John F rom: Alley, D avid Sent: Sunday, September 4, 2016 8:50 AM To: Tsao, John; Lyon, Fred

Subject:

RE: Status 3 PM Saturday 9/3: TS Shutdown at Wolf Creek (RCS Leakage TS 3.4.13)

John, Phone call went well. You were not needed. Looks like this issue will just move up the start of their refueling outage (which was scheduled for later this month. It does not appear that this event will be time sensitive. I doubt that there will be any more that needs to be done on this this weekend.

Dave From: Tsao, John Sent: Saturday, September 03, 2016 10:27 PM To: Lyon, Fred <Fred. Lyon@nrc.gov> Cc: Alley, David <David.All ey@nrc.gov>

Subject:

Re: Status 3 PM Saturday 9/3: TS Shut down at Wolf Creek (RCS Leakage TS 3.4.13)

Fred, I am sorry that i missed your call to my home and I missed the 2 Pm phone call with the licensee.

I was out t oday and did not check my home phone for your voice mail until now (Saturday 10:15PM). If you need me tomorrow Sunday or Monday my cell phone number isl.... <b-)(6_J_ _ ___, John From: Lyon, Fred Sent: Saturday, September 3, 2016 2:53 PM To: Pascarelli, Robert; Alley, David; Tsao, John; 'pascarelli1991@verizon.net '

Cc: W ilson, George; Boland, Anne; Klein, Alex; Woodyatt, Diana; Grover, Ravinder

Subject:

RE: St at us 3 PM Saturday 9/3: TS Shutdown at Wolf Creek (RCS Leakage TS 3.4.13) We completed our 2 PM phoncon with the licensee. The licensee is in M3 at NOP/T. They are restoring the auxiliary boiler to service, which was OOS for maintenance, and then will go to MS and essentially go into the RFO that was scheduled to begin on 9/24. They will decide over the next 3 weeks what can be brought forward in the outage. In a nutshell, the licensee determined that the leak is about 0.5 gpm from the canopy seal weld on penetration 77 (CET). The licensee exited the TS last evening upon identifying the leakage. The licensee intends to cool down, put the RVH on the stand, and use a mechanical clamp to repair the leak. Westinghouse has an available clamp, and these have a long history of use at WCGS and in the industry. They considered doing a weld repair, but that appears to involve much more risk of failure to get a good weld or recurrence of leak. The licensee must also verify that the leak is actually a canopy seal weld leak and that the penetration threads (the pressure boundary) are not adversely impacted . No licensing actions are necessary; the licensee will do the work under their design modification process. More detailed information will follow from the licensee, and Nick will send out a summary email shortly. From: Lyon, Fred Sent: Saturday, September 03, 2016 10:10 AM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; 'pascarelli199l@verizon.net' <pascarelli1991@verizon.net> Cc: Wilson, George <George.Wilson@nrc.gov>; Boland, Anne <Anne.Boland@nrc.gov>; Klein, Alex <Alex.Klein@nrc.gov>; Woodyatt, Diana <Diana.Woodyatt@nrc.gov>; Grover, Ravinder

<Ravinder.Grover@nrc.gov>

Subject:

RE: Status 10 AM Saturday 9/3: TS Shutdown at Wolf Creek (RCS Leakage TS 3.4.. 13) Importance: High The licensee determined that the leak is from penetration 77, a core exit thermocouple canopy seal. Below is an excerpt from an IR documenting a similar leak on penetration 20 during the 2015 outage that Jim Drake inspected for RIV. Contrary to our previous belief, there was no relief request associated with it. During refueling outage RF20, a visual examination (VT-2) of the reactor pressure vessel head was performed. The examination was in accordance with Code Case N 729-1 Table 1, Item B4.20. An indication of primary water stress corrosion cracking was identified on the canopy seal weld for CRDM penetration 20. The CRDMs were fabricated in sections with threaded joints providing the pressure-retaining capabilities. Since the threaded joint provides pressure retention, the canopy seal weld is not pressure retaining and is for leakage control. The licensee installed a mechanical clamp on the canopy seal weld to restore leakage control.

There will be a phoncon at 2 PM ET today to discuss the licensee's repair plans. I've forwarded the scheduler to you. Anne, George, Diana, Alex, Ravi: I don't t hink it' s necessary for you to call in, unless you feel you need a piece of this pie. From: Lyon, Fred Sent: Friday, September 02, 2016 11:30 PM To: Pascarell i, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, Joh n <John.Tsao@nrc.gov>; 'pascarelli199l @verizon.net ' <pascarelli1991@verizon.net> Cc: Wilson, George <George.Wilson@nrc.gov>; Boland, Anne <Anne.Boland@nrc.gov>; Klein, Alex <Alex.Klein@nrc.gov>; Woodyatt, Diana <Diana .Woodyatt@nrc.gov>; Grover, Ravinder

<Ravinder.Grover@nrc.gov>

Subject:

Status 11 PM Friday 9/2: TS Shutdown at Wolf Creek (RCS Leakage TS 3.4.13) Importance: High The licensee identified t he source of the leak as t he RVH, specifically, a nozzle that was repaired with a mechanical clamp last outage (spring 2015). I've no other informat ion yet; t he licensee is gathering drawings, taking photos, et al., to provide to RIV. Nick Taylor, the RIV DRP BC, will set up a call about midday t omorrow, specific time TBD, wit h the licensee to discuss whether t his is pressure boundary leakage and how they intend to repair it. I'll provide the information when it is available. Dave, John: Jim Drake, who apparently was involved in the inspection of t he repair last outage, recommended you listen in on the call for consult to RIV. Perhaps he consulted you on the issue? I did an ADAM S search back to 11/2014, but did not find any relief request that might apply. Thanks, Fred From: Lyon, Fred Sent: Friday, Sept ember 02, 2016 12:10 PM To: Pascarell i, Robert <Robert.Pascarelli@nrc.gov> Cc: Wilson, George <George.Wilson@nrc.gov>; Boland, Anne <Anne.Boland@nrc.gov>; Klein, Alex <Alex.Klein@nrc.gov>; Woodyatt, Diana <Diana.Woodyatt@nrc.gov>; Grover, Ravinder

<Ravinder.Grover@nrc.gov>

Subject:

Status: TS Shutdown at Wolf Creek (RCS Leakage TS 3.4.13) Importance: High On the 10 AM CT RIV/licensee phoncon, the licensee discussed the indications they have so far and plans go ing forward (all times CT).

1. Shutdown to M3 is in progress and will be done about 2 PM. They will remain at NOP/NOT to enter the bioshield and search for the leak source. If they find the source, the leak will then be identified leakage, and the TS limit is 1O gpm; so they will be able to exit the current TS required action. Then , if possible, they would remain in M3 to conduct repairs.

If they are unable to locate the leak, then they will re-assess. It would be easier to troubleshoot at NOP/NOT, but the TS action is to be in M5 within 36 hours. It is highly unlikely that NRC would grant a NOED for them to remain at NOP/NOT in M3 for troubleshooting activities.

2. Inside containment: Rad monitors are stable just above MDA. Sump pumpdown rate is slightly up, but the sump is on the other side of containment from where the most likely leak sources are located. VCT level decrease rate is slightly increased. No substantive temperature/humidity changes; there is a small steam leak on a SGBD valve (secondary side).

The leakrate is over the TS limit of 1 gpm when excess LD is in service (1.35 gpm). With normal LD in service, the leakrate is below the TS limit of 1 gpm (0.54 gpm). Technically, they are below the TS limit right now (0.54 gpm with normal LO in service). Rad monitor background level is slightly elevated from 1E-11 to 1E-9 over the past week. Rad monitor filter levels are slightly elevated over the past week from 1E-11 to 1E-10, mostly 1-131 and 1-133; no short-lived isotopes have been detected. There are no known fuel defects at WCGS. Licensee has not been able to correlate the indications that they have within containment to a particular system or location. Licensee has contacted Callaway, WSI, and Areva in case repair support is needed. The licensee will status call RIV/DRP/RPB-B/BC (Nick Taylor) late day shift today after they are able to enter containment and do an initial search. From: Lyon, Fred Sent: Friday, September 02, 2016 10:15 AM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov> Cc: Wilson, George <George.Wilson@nrc.gov>; Boland, Anne <Anne.Boland@nrc.gov>; Klein, Alex <Alex.Klein@nrc.gov>; Woodyatt, Diana <Diana.Woodyatt@nrc.gov>; Grover, Ravinder

<Ravinder.Grover@ nrc.gov>

Subject:

UPDATE: Tech spec shutdown at Wolf Creek this morning Importance: High The licensee's last leakrate determination, at about 8:00 AM CT, was 0.52 gpm; however, they do not trust it as much as the earlier determination of 1.35 gpm. Therefore, they are continuing with TS action to be in M3 within 6 hours (from a start time of 8:08 AM CT today). They will begin shutdown at 10 AM CT. They've made 2 containment entries so far but have not located the leak. A 4-hour report per 10 CFR 50.72 will follow (due between 10 AM -2 PM CT today). The next action level, if the leakage were to go to 1O gpm, would be a NOUE. The licensee has a phoncon scheduled with RIV at 10 AM CT today that I will be listening to, and I will update you afterwards. From: Lyon, Fred Sent: Friday, September 02, 2016 8:12 AM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>

Cc: Wilson, George <George.Wilson@nrc.gov>; Boland, Anne <Anne.Boland@nrc.gov>

Subject:

FYI: Tech spec shutdown likely at Wolf Creek this morning Importance: High I'll get more information on my morning call at 9:45. From: Taylor, Nick Sent: Friday, September 02, 2016 8:10 AM To: Pruett, Troy <Troy. Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>; Vegel, Anton

<Anton.Vegel@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>; Kennedy, Kriss
<Kriss.Kennedy@nrc.gov>; Morris, Scott <Scott.Morris@nrc.gov>

Cc: Warnick, Greg <Greg.Warnick@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Janicki, Steven <Steven.Janicki@nre.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Lyon, Fred <Fred.Lyon@nrc.gov>

Subject:

Tech spec shutdown likely at Wolf Creek this morning Importance: High Good morning everyone, As has been discussed in the morning status meetings this week, Wolf Creek continues to struggle with unidentified RCS leakage. That leakage trend worsened through the day yesterday and overnight. The resident inspector spent much of last evening at the site monitoring the site's actions. We have learned that this morning at 0408, the station recorded an unidentified leak rate of 1.35 gpm, in excess of the tech spec allowed limit for unidentified leak rate (TS 3.4.13). Action statement A requires them to reduce the leakage to within limits within 4 hours or enter action statement B, which would require the plant to be in Mode 3 within 6 hours. Wolf Creek started preparations for shutdown last night. I spoke with the plant general manager, Steve Smith, a few minutes ago. He expressed to me that their plans are for the control room to commence a normal plant shutdown at about 0745 if they are unable to find and isolate the leak. Their plans are to proceed to mode 3 as quickly as possible and use the allowed time in tech specs to look for the leak in containment with the plant shut down but still at pressure. He does not expect them to have success finding the leak based on the actions last night. Fabian has been in near-continuous dialogue with the control room through the morning. Plant conditions (aside from leakage) are stable. Containment humidity and temperatures are steady & normal for these plant conditions. Pressurizer level is in the normal range and not trending (around 58%). The licensee is in their off-normal procedure for excessive RCS leakage, which has scram actions should pressurizer level near 6%, but it does not appear that is a likely outcome. Fabian also reported that the most recent leak rate this morning is down to about 0.52 gpm, but their plans remain to conduct a shutdown in order to locate the leak. David Proulx and Steve Janicki are in the Region IV office today. I will continue to be involved via cell phone. We have a call scheduled for 1000 this morning with their plant management to discuss their plans (this call was already on the books based on discussions yesterday).

Please call me on cell phone if you have questions. Otherwise I will communicate regularly through the day with David and Steve.

Thanks, Nick Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV 0 : 817 200-1141 C: (bJ(6J E: nick.taylor@nrc.gov

From: Collins, Jay Sent: 6 Sep 2016 20:32:14 +0000 To: Drake, James

Subject:

Wolf Creek Boric Acid Leaking on Head Greetings, Catching this issue from the sidelines, but I thought I would put a bee in your ear to remind you about the problems we had with Fort Calhoun and the cleaning of their head last year. I dona't know the inspection requirements for Wolf Creek this refueling outage, but I figure they are at least going to have to clean the head for a VT-2 inspection. Cleaning the head in too aggressive of a manner can invalidate the visual inspection and may then trigger a vollumetric inspection. Just a heads up for a problem they may not be thinking about, using lessons learned that Region IV caught earlier at Fort Calhoun. This was lsaaca's issue at Fort Calhoun, so I am sure he has all the fine details. Just trying to be helpful, Jay

From: Tsao, John Sent: 3 Oct 2016 17:28:52 +0000 To: Alley, David

Subject:

RE: Internal communications at Wolf Creek re head corrosion I plans Dave, Yes we should be on the call with Wolf Creek tomorrow From: Alley, David Sent: Monday, October 03, 2016 1:21 PM To: Tsao, John <John.Tsao@nrc.gov>

Subject:

FW: Internal communications at Wolf Creek re head corrosion I plans John Please take a look at this Greg Just tried to call - no answer. I am tied up for a while this PM. Might be good for us to be on the call tomorrow Dave From: Werner, Greg Sent: Monday, October 03, 2016 1:08 PM To: Alley, David <David.Alley@nrc.gov> Cc: Taylor, Nick <Nick.Taylor@nrc.gov>

Subject:

FW: Internal communications at Wolf Creek re head corrosion/ plans FYI. Just giving you a heads up in case WC asks for relief. NO OTHER information other than what is in the attached file, which is part of a CR and an internal WC newsletter. We are planning an informational call with WC sometime tomorrow, would you like to be included on the appointment? We are trying to find out the status of the head cleaning, information on potential relief requests, and how they selected the other 4 penetrations for the clamps. Greg Werner From: Taylor, Nick Sent: Monday, October 03, 2016 11:50 AM To: Werner, Greg <Greg.Werner@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov> Cc: Pruett, Troy <Troy.Pruett@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>; Lantz, Ryan

<Ryan.Lantz@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>; Proulx, David <David .Proulx@nrc.gov>;

Janicki, Steven <Steven.Janicki@nrc.gov>

Subject:

Internal communications at Wolf Creek re head corrosion I plans

All, I'm still working on setting up a call with the licensee tomorrow. But Doug provided the attached today from the licensee's CAP and internal outage newsletters. I added the red comment boxes.

Thanks, Nick

From: Sydnor, Christopher Sent: 4 Oct 2016 20:10:02 -0400 To: Alley, David;Anchondo, lsaac;Baquera, Mica;Bloodgood, Michael;Bozga, John;Brand, Javier;Bu rket, Elise;Butcavage, Alexander;Carrion, Robert;Case, Michael;Chaudhary, Suresh;Clayton, Kelly;Collins, Brendan;Collins, Jay;Cumblidge, Stephen;Drake, James;Dykert, Jason;Farnholtz, Thomas;Floyd, Niklas;Gavula, James;Gray, Harold;Hills, David;Hiser, Allen;Ho lmberg, Mel;Honcharik, John;Jayroe, Peter;Jones, William;Jose, Benny;Karwoski, Kenneth; Kaufman, Paul;Kopriva, Ron;Lupold, Timothy;Makor, Shiattin;Meghani, Vijay;Mitchell, Matthew;Modes, Michael;Neurauter, James;Nove, Carol;Poehler, Jeffrey;OHara, Timothy;Reichelt, Eric; Reinert, Dustin; Rezai, Ali; Rivera Ortiz, Joel;Rudland, David;Sanchez Santiago, Elba;Sengupta, Abhijit;Shaikh, Atif;Sifre, Wayne;Taylor, Robert;Wallace, Jay;Williams, Robert;Young, Matt; Huang, John;Tsao, John;Thomas, Brian; Dunn, Darrell;Tregoning, Robert;Jandovitz, John;Davis, Robert;Widrevitz, Dan;Diaz-Colon, Marioly X;McMurray, Nicholas;Gray, Mel;Johnson, Andrew;Vitto, Steven;Yeshnik, Andrew;Lin, Bruce;Oberson, Greg; Focht, Eric;Smith, Laura;Turilin, Andrey;Walker, Shakur;Shuaibi, Mohammed; Kulp, Jeffrey;Pettis, Robert;Werner, Greg;Li, Yong;Kalikian, Roger;Fernandez, Edison;Domke, Matthew;Raynaud, Patrick;Hovanec, Christopher;Cheruvenki, Ganesh;Fairbanks, Carolyn;Sheng, Simon;Young, Austin;Jenkins, Joel;Dijamco, David;Cooper, Paula

Subject:

September 2016 MECC Meeting Minutes

Dear MECC Call Participants,

Please see the following meeting minutes for the items that were discussed on the September 21, 2016 MECC Call. Please correct any inaccuracies in the below and revise/supplement as required. These will go into the SharePoint meeting summary.

1. Wolf Creek Leakage through Threaded Connections (Isaac Anchondo & Greg Warner.

RIV, DRS): Control room operators calculated unidentified reactor coolant system leakage of 1.358 gallons per minute, in excess of the TS 3.4.13 limit of 1 gallon per minute. The licensee entered containment and discovered a non-pressure boundary leak from a canopy seal weld on a reactor vessel head penetration that serves one of the core exit thermocouples. The licensee has opted to use a CROM Seal Clamp Assembly as a repair method. The licensee has justified the structural integrity of the threads with a Westinghouse calculation showing a maximum allowable leak of 3.5 GPM. Region IV staff is still in the process of reviewing this issue with the help of NRR. During the call, RIV staff indicated that the main concern is that actual thread degradation is causing the leak. The licensee is trying to justify that there's no thread degradation, the leak is due to the canopy seal weld, and that the repair using the CROM Seal Clamp Assembly is adequate as a corrective action. Staff emphasized the concern that the licensee isn't addressing the potential root cause of thread degradation, nor adequately justifying why the threads are still intact. The Westinghouse calculation is currently being reviewed, but there are concern s. EPNB staff proposed that if it can be established that only a small leak rate could be expected for the seal weld with undamaged threads, then the NRC would be in a good position to ask the licensee to provide a rigorous evaluation of the structural integrity of the threads. Conversely, if undamaged threads can, in fact, leak 1.358 GPM then the NRC may not necessarily be in the best position to request a more rigorous evaluation. Other points that were raised on this pertained to boric acid issues as a result of the leakage, whether there's other industry OE with leakage resulting from thread damage, and design criteria.

2. Peening, Region IV/DRS and NRR/DE: Staff discussed the EPRI Materials Reliability Program Topical Report, MRP-335, "Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement," and associated staff SE. The outgoing NRR SE is publically available at ADAMS Accession No. ML16208A485 and the NRR/DE/EPNB non-concurrence document is publically available at ML16187A319.

The NRR contacts for this are Stephen Cumblidge, Jay Collins, David Alley, and John Tsao.

3. Use of UT in Lieu of RT for Component Repair/Replacements (Steve Cumblidge, NRR/DE/EPNB): Staff discussed two Code Cases related to this issue: CC N-818- not getting endorsed by the staff; CC N-831 - looking more favorable but not yet in RG 1.147 (please correct these Code Case Numbers if they're not accurate). Staff also discussed a Millstone Code Alternative to implement UT in lieu of RT and mentioned that all plants proposing to implement this require NRC authorization of ASME Code Alternatives, per 50.55a(z). It was also mentioned that some plants are using UT in lieu of RT for non-safety-related piping.

4. NRC Generic Letter 90-05 (Isaac Anchondo, RIV/DRS): There was question regarding licensee Relief Request submittals based on GL 90-05. Cooper had this issue with piping less than the minimum wall thickness. Per GL 90-05, licensees must perform Code repairs or request NRC to grant relief for temporary non-Code repairs on a case-by-case basis regardless of pipe size for all ASME Code Class 1, 2, and 3 piping. The GL provides specific guidance for acceptable Code Relief for performing the temporary non-Code repair of ASME Code Class 1, 2, and 3 piping. It was emphasized that they all must submit the Relief Request for non-Code repairs.

I'll send out the scheduler for the next MECC call on Wednesday, October 19.

Thanks, Chris Christopher R. Sydnor Materials Engineer Vessels and Internals Integrity Branch NRR/Div. of Engineering USNRC (301) 41 5-6065 Office: 0-9Hl 4 Christopher.Sydnor@nrc.gov

From: Drake, James Sent: S Oct 2016 11:11:10 -OSOO To: Werner, Greg;Taylor, Nick Cc: Alley, David;Anchondo, Isaac

Subject:

RE: Relief request coming from Wolf Creek Code requires the seal weld for threaded connections on Class 1 component. From : Werner, Greg Sent: Wednesday, October OS, 2016 11:10 AM To: Drake, James <James.Drake@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>

Subject:

RE: Relief request coming from Wolf Creek If it is just thread leakage, what is the flaw? The seal weld is only for housekeeping - nothing structurally. Does the code require a seal weld? Don't remember? From: Drake, James Sent: Wednesday, October OS, 2016 11:07 AM To: Werner, Greg <Greg.Werner@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov> Cc: Alley, David <David .Alley@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>

Subject:

RE: Relief request coming from Wolf Creek Code requires the flaw be removed or restored to an acceptable condition as part of the repair, unless allowed by a Code case. Since there isn't a Code Case to allow leaving the repair that is probably the purpose of the relief request. Jim From: Werner, Greg Sent: Wednesday, October OS, 2016 11:04 AM To: Taylor, Nick <Nick.Taylor@nrc.gov> Cc: Alley, David <David .Alley@nrc.gov>; Drake, James <James .Drake@nrc.gov>; Anchondo, Isaac

<lsaac.Anchondo@nrc.gov>

Subject:

RE: Relief request coming from Wolf Creek I suspect it is because they are going to install a CSCA clamp, so it really doesn't matter, as long as you believe all of the boron came from the leaking canopy seal and no leakage was from the J-groove weld for that penetration. Pure speculation on my part. Greg From : Taylor, Nick Sent: Wednesday, October OS, 2016 10:22 AM To: Singal, Balwant <Balwant.Singal@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Kopriva, Ron

<Ron.Kopriva@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian
<Fabian.Thomas@nrc.gov>; Allley, David <David.Alley@nrc.gov>

Cc: Proulx, David <David.Proulx@nrc.gov>; Janicki, Steven <Steven.Janicki@nrc.gov>; Pruett, Troy

<Troy.Pruett@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>

Subject:

Relief request coming from Wolf Creek

All, I just got off the phone with the reg affairs manager at Wolf Creek (Cindy Hafenstine). She was calling to correct one thing they told us yesterday. They have apparently decided to request relief from performing the volumetric inspection on Penetration 77 only (not all 12). She did not know the basis for the request, nor did she know when they would be ready to submit the req uest. This was an early heads up that it is coming.

I'll share any information I receive on this as soon as I get it.

Thanks, Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV 0: (817) 200-1141 C: (b)(6)

E: nick.ta lor ov

From: Tsao, John Sent: 5 Oct 2016 13:33:33 -0400 To: Collins, Jay

Subject:

FW: Pictures of WC penetration Attachments: DSC00024.jpg, DSC00026.j pg Fyi wolf creek CROM boric acid deposits From : Drake, James Sent: Wednesday, October OS, 2016 11:45 AM To: Alley, David <David.Alley@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Hoffman, Keith

<Keith.Hoffman@nrc.gov>

Subject:

Pictures of WC penetration These are pictures of Penetration 77. Will try to get copies of additional pictures. Jim ~rmes 'f". :Drafe James F. Drake Office phone: 817-200-1558 Cell Phone: l{b){6) I

From: Anchondo, Isaac Sent: 11 Oct 2016 15:01:15 -0500 To: Tsao, John;Collins, Jay

Subject:

WC Call - Item of Note Greetings, I was just thinking, what happens if Nozzle 77 & 78 are included in the nozzles to be UT/Leakpath given that they are also requesting relief from the examination volume for those two penetrations? I just wanted to point that out as food for thought since we didn't ask them on the call.

Thanks, Reactor Inspector U.S. Nuclear Regulatory Commission I Region JV Division of Reactor Safety I Enginee1ing Branch 2 (817) 200-1152

From: Collins, Jay Sent: 11 Oct 2016 20:49:54 +0000 To: Anchondo, Isaac

Subject:

RE: WC Call - Item of Note Attachments: DSCF3798.jpg, DSCF3797.jpg, DSCF3792.jpg, DSCF3789.jpg, DSCF3788.jpg, DSCF3786.jpg Yes. Hey could you confirm that these are pictures from Wolf Creek this outage? Jay From: Anchondo, Isaac Se nt: Tuesday, October 11, 2016 4:43 PM To: Collins, Jay

Subject:

RE: WC Call - Item of Note Strictly my opinion (not the branch), I think that if we hold them to the same cleaning limitations as FCS, there doesna't seem to be a way for Cooper to clean it without having acerelevant indicationsaL left in place. But isna't this the reason they are performing the volumetric examinations? From: Collins, Jay Sent: Tuesday, October 11, 2016 3:35 PM To: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>

Subject:

RE: WC Call - Item of Note Well I have some pictures, in my mind from the discussion on the phone call, there is some areas of significant masking. The cleanliness that we got at Fort Calhoun seems like it would be difficult, without their power washing. From: Anchondo, Isaac Se nt: Tuesday, October 11, 2016 4:20 PM To: Collins, Jay <Jay.Collins@nrc.gov>

Subject:

RE : WC Call - Item of Note la'm not the inspector on-site. Would you like me to ask Ron Kopriva to give you a call sometime tomorrow? Isaac From: Collins, Jay Sent: Tuesday, October 11, 2016 3:16 PM To: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>

Subject:

RE: WC Call - Item of Note They are ones that they would have to perform the inspection on. The volumetric leak path is performed on the nozzle above the weld. The limitation to inspection coverage is below the weld. Therefore, not a specific concern for these locations. I would very much appreciate your impression of the cleanliness of that head though. Any thoughts, or perhaps a conversation tomorrow sometime, would be useful. Jay From: Anchondo, Isaac Sent: Tuesday, October 11, 2016 4:01 PM To: Tsao, John <John .Tsao@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov> Subje ct: WC Call - Item of Note Greetings,

I was just thinking, what happens if Nozzle 77 & 78 are included in the nozzles to be UT/Leakpath given that they are also requesting relief from the examination volume for those two penetrations? I just wanted to point that out as food for thought since we didna't ask them on the call.

Thanks,

~~ Reactor Inspector U.S. Nuclear Regulatory Commission I Region IV Division of Reactor Safety I Engineering Branch 2 (817) 200-1152

From: Collins, Jay Sent: 11 Oct 2016 16:41:29 -0400 To: Tsao, John;Alley, David; Hoffman, Keith

Subject:

FW: W olf Creek RX Vessel Head Pictur es - As-Found during M irror Insulation Removal Attachments: DSCF3798Jpg, DSCF3797Jp& DSCF3792Jp& DSCF3789Jp& DSCF3788Jp& DSCF3786.jpg These were reported to be from Wolf Creek, and also part of my concern. I do not know if you guys have seen these. Jay

From: Collins, Jay Sent: 11 Oct 2016 14:00:33 +0000 To: Singal, Balwant

Subject:

Accepted: Wolf Creek Relief Request - Outage Support

From: Singal, Balwant Sent: 11 Oct 2016 11:37:22 -0400 To: Alley, David;Collins, Jay;Tsao, John; Muilenburg William T Cc: Lingam, Siva;Smith Stephen L;Reasoner Cleve O;Taylor, Nick;Hafenstine Cynthia R;Werner, Greg;Anchondo, lsaac;Drake, James; Kopriva, Ron

Subject:

Wo lf Creek Relief Request - Outage Support Attachme nts: [Externa l_Sender] Wolf Creek Relief Request Anticipated, [External_Sender] Wolf Creek Relief Requests 14R-03 & 04 Revised to attach e-mail with copy of the reliefrequest. No response needed. Thanks. Please see the attached e-mail from WCNOC requesting a phone call in support of the reliefrequest in support of the current refueling outage. The forma l relief request is to be submitted by noon today. I will pass on the copy of the relief request to everyone after receipt. Please use the following Bridge No. for this call. 866-624-3402 Passcode j(b)(B) j# Thanks.

From: Muilenburg William T Sent: 7 Oct 2016 21:17:37 +0000 To: Singal, Balwant

Subject:

[External_Sender] Wolf Creek Relief Request Anticipated

Balwant, I wanted to give you advance notice that on Monday morning (10/10) Wolf Creek will be sending a Relief Request for review concerning reactor vessel head inspections. We need to perform supplemental exams on certain penetrations and we have two concerns.

First, one penetration is one where we have had relief on before because of access concerns and we will need to request the same relief again (ML12353A241 provided NRC Safety Assessment of the request), and second, we will be asking to perform an alternate exam v. that specified in code case N-729. Can you help us assemble the right people to have a phone call regarding this request on Monday morning? I will be in Saturday and Sunday if there are any questions I cain help answer.

Thanks, Bill Muilenburg 620-364-4186

From: Muilenburg William T Sent: 11 Oct 2016 15:23:50 +0000 To: Singal, Balwant Cc: 'Nicholas.Taylor@NRC.Gov';Dodson, Douglas

Subject:

[External_Sender) Wolf Creek Relief Requests 14R-03 & 04 Attachments: pr222_.PDF See listing of Records A lready Ava ilable to the Pub lic for attachment.

Balwant, Here is the relief requests for discussion this afternoon.

Thanks for arranging today's call, Bill Muilenburg

From: lingam, Siva Sent: 12 Oct 2016 10:59:57 -0400 To: jaknust@WCNOC.com;wimuile@WCNOC.com Cc: Pascarelli, Robert;Singal, Balwant;Alley, David;Tsao, John;Collins, Jay;Taylor, Nick;Groom, Jeremy

Subject:

Wolf Creek--DRAFT RAls for Relief Request 14R-04 CROM nozzle examination Attachme nts: Wolf Creek RAl .. docx Attached please find the draft RAls for the subject relief request. Balwant isl(b)(6) ~ this morning, and will be working this afternoon. Balwant will forward the official RAls later. Siva P. lingam U.S. Nuclear Regulatory Commission Project Manager (NRR/DORL/LPL4-1) Palo Verde Nuclear Generating Station Location: 08-05; Mail Stop: 08-B3 Telephone: 301-415-1564; Fax: 301-415-1222 E-mail address: siva.lingam@nrc.gov

REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST 14R-04 ALTERNATE EXAMINATION OF CONTROL ROD DRIVE MECHANISM NOZZLE PENETRATIONS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 By letter dated October 11 , 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-04 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration numbers 77 and 78. To complete its review, the Nuclear Regulatory Commission (NRC) requests the following additional information.

1. (a) Discuss whether CROM nozzle numbers 77 and 78 will be ultrasonically examined during the current fall 2016 refueling outage as part of the supplement examinations in accordance with Relief Request 14R-03. (b) Discuss the schedule of the future ultrasonic examination of nozzle numbers 77 and 78 during the fourth inservice inspection interval (e.g.,

which refueling outage? What year?) (3) Discuss results of previous inspections of these two nozzles.

2. (a) Explain why the examination distance (0.64 inches) for CROM nozzle number 78 obtained in 2013 was reduced as compared to the examination distance (0.88 inches) obtained in the 2006 inspection as shown in Table 2 on page 5 of Relief Request 14R-04. (b) Discuss whether the examination distance will be reduced further for nozzle numbers 77 and 78 in the future inspections. If yes, the licensee needs to propose an examination distance for nozzle numbers 77 and 78 that it can achieve in the next inspection during the fourth inservice inspection interval.

3. By letters dated July 2, 2012 (ADAMS Accession ML12193A559) with supplement dated October 15, 2012 (ADAMS Accession No. ML12341A228), the licensee proposed alternate examination distances for CROM nozzle numbers 77 and 78 as shown in Relief Request 13R-

07. In Relief Request 13R-07, the licensee proposed examination distances of 0.6 inches and 0.88 inches for nozzle numbers 77 and 78, respectively. These two values were obtained during the 2006 inspection. By letter dated January 4 , 2013 (ADAMS Accession No, ML12353A241 ), the NRC staff approved Relief Request 13R-07 based on the examination distances of 0.6 inches and 0.88 inches. However, as shown in Relief Request 14R-04, the actual examination distances obtained during the 2013 inspection were 0.6 inches and 0.64 inches for nozzle numbers 77 and 78, respectively. Nozzle number 78 did not achieve the examination distance of 0.88 inches that the NRC approved for Relief Request 13R-07 for the 2013 inspection. Please explain the discrepancy.

4. Figure 1 of Relief Request 14R-04 shows the threaded region; however, Figure 1 is not clear regarding the examination distance with respect to the location of the J-groove weld .

Provide sketches (hand sketches are acceptable) of the region below the J-groove weld for CROM nozzle numbers 77 and 78. The sketch should be similar to Figure 2 of ASME Code Case N-729-1 , including the use alphabet to show demarcations. The sketches should include the following information. (a) The total length of the CROM nozzle from the elevation of the toe of the J-groove weld to the bottom of the CROM nozzle numbers 77 and 78. (b) Identify the threaded region which is approximately 1.19 inches as stated in the relief request, (c) Identified the required "a" distance

of 1.0 inch. (d) Identify the inspected distance of 0.6 inches and 0.64 inches for nozzles 77 and 78 that were obtained in the 2013 inspection. (e) Identify the starting poinUlocation of the initial flaw and the approximate location of the final flaw tip at the time of the next inspection. (f) Identify the zone of greater than 20 ksi which is 0.3 inches as stated in the relief request.

5. Figure 3 of Relief Request 14R-04 shows crack growth prediction. (a) Confirm that if a flaw is initiated at 0.15 inches below the toe of the J-groove weld as shown in Figure 3, the upper flaw tip would reach the toe of the J-groove weld after 6 effective full power years. (b) If a fllaw is initiated below the crack growth curve in Figure 3 (e.g., below 0.15 inch location), at what effective full power years will the final flaw tip reach the toe of the J-groove weld?

From: Singal, Balwant Sent: 12 Oct 2016 15:56:23 -0400 To: Alley, David;Tsao, John;Collins, Jay Cc: Pascarelli, Robert;Lingam, Siva

Subject:

Wolf Creek Relief Request - Status

All, Surprisingly, I never heard from anyone from Wolf Creek about the relief request needed in support of the current refueling outage since the discussions yesterday and was unable to reach anyone in their licensing department. I just spoke to their Reg. Affairs Manager, Cindy Hafenst ine and she indicated that they are still working on revising the relief request. We are expected to get it either by end of the day today or early tomorrow morning.

I indicated to her that it w ill be very difficult to process it and provide the verbal by Friday. Thanks for everyone's patience. Balwant K. Singal Senior Project Manager (Diablo Canyon) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwa nt.Si nga l@nrc.gov Tel : (301) 415-3016 Fax: (301) 415-1222

From: Tsao, John Sent: 12 Oct 2016 10:32:30 +0000 To: Collins, Jay;Anchondo, Isaac

Subject:

RE: WC Call - Item of Note

Isaac, I plan to ask the licensee whether they will inspect nozzles 77 and 78 in this outage.

From: Collins, Jay Sent: Tuesday, October 11, 2016 4:16 PM To: Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>

Subject:

RE: WC Call - Item of Note They are ones that they would have to perform the inspection on. The volumetric leak path is performed on the nozzle above the weld. The limitation to inspection coverage is below the weld. Therefore, not a specific concern for these locations. I would very much appreciate your impression of the cleanliness of that head though. Any thoughts, or perhaps a conversation tomorrow sometime, would be useful. Jay From: Anchondo, Isaac Sent: Tuesday, October 11, 2016 4:01 PM To: Tsao, John <John .Tsao@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov>

Subject:

WC Call - Item of Note Greetings, I was just thinking , what happens if Nozzle 77 & 78 are included in the nozzles to be UT/Leakpath given that they are also requesting relief from the examination volume for those two penetrations? I just wanted to point that out as food for thought since we didn't ask them on the call.

Thanks, Reactor Inspector U.S. Nuclear Regulatory Commission I Region IV Division of Reactor Safety I Engineeri ng Branch 2 (8 17) 200- 11 52

From: Tsao, John Sent: 12 Oct 2016 09:59:02 -0400 To: Singal, Balwant Cc: Lingam, Siva;Alley, David;Collins, Jay

Subject:

Wolf Creek--RAI for Relief Request 14R-04 CROM nozzle examination Attachments: Wolf Creek RAl .. docx

Balwant, Attached are my draft RAI questions regarding the subject relief request. Dave Alley has not reviewed my draft RAI questions. I am forward them to you because the urgency of the review.

Please forward my RAI questions to the licensee as "draft" Thanks John

REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST 14R-04 ALTERNATE EXAMINATION OF CONTROL ROD DRIVE MECHANISM NOZZLE PENETRATIONS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 By letter dated October 11 , 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-04 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration numbers 77 and 78. To complete its review, the Nuclear Regulatory Commission (NRC) requests the following additional information.

1. (a) Discuss whether CROM nozzle numbers 77 and 78 will be ultrasonically examined during the current fall 2016 refueling outage as part of the supplement examinations in accordance with Relief Request 14R-03. (b) Discuss the schedule of the future ultrasonic examination of nozzle numbers 77 and 78 during the fourth inservice inspection interval (e.g.,

which refueling outage? What year?) (3) Discuss results of previous inspections of these two nozzles.

2. (a) Explain why the examination distance (0.64 inches) for CROM nozzle number 78 obtained in 2013 was reduced as compared to the examination distance (0.88 inches) obtained in the 2006 inspection as shown in Table 2 on page 5 of Relief Request 14R-04. (b) Discuss whether the examination distance will be reduced further for nozzle numbers 77 and 78 in the future inspections. If yes, the licensee needs to propose an examination distance for nozzle numbers 77 and 78 that it can achieve in the next inspection during the fourth inservice inspection interval.

3. By letters dated July 2, 2012 (ADAMS Accession ML12193A559) with supplement dated October 15, 2012 (ADAMS Accession No. ML12341A228), the licensee proposed alternate examination distances for CROM nozzle numbers 77 and 78 as shown in Relief Request 13R-

07. In Relief Request 13R-07, the licensee proposed examination distances of 0.6 inches and 0.88 inches for nozzle numbers 77 and 78, respectively. These two values were obtained during the 2006 inspection. By letter dated January 4 , 2013 (ADAMS Accession No, ML12353A241 ), the NRC staff approved Relief Request 13R-07 based on the examination distances of 0.6 inches and 0.88 inches. However, as shown in Relief Request 14R-04, the actual examination distances obtained during the 2013 inspection were 0.6 inches and 0.64 inches for nozzle numbers 77 and 78, respectively. Nozzle number 78 did not achieve the examination distance of 0.88 inches that the NRC approved for Relief Request 13R-07 for the 2013 inspection. Please explain the discrepancy.

4. Figure 1 of Relief Request 14R-04 shows the threaded region; however, Figure 1 is not clear regarding the examination distance with respect to the location of the J-groove weld .

Provide sketches (hand sketches are acceptable) of the region below the J-groove weld for CROM nozzle numbers 77 and 78. The sketch should be similar to Figure 2 of ASME Code Case N-729-1 , including the use alphabet to show demarcations. The sketches should include the following information. (a) The total length of the CROM nozzle from the elevation of the toe of the J-groove weld to the bottom of the CROM nozzle numbers 77 and 78. (b) Identify the threaded region which is approximately 1.19 inches as stated in the relief request, (c) Identified the required "a" distance

of 1.0 inch. (d) Identify the inspected distance of 0.6 inches and 0.64 inches for nozzles 77 and 78 that were obtained in the 2013 inspection. (e) Identify the starting poinUlocation of the initial flaw and the approximate location of the final flaw tip at the time of the next inspection. (f) Identify the zone of greater than 20 ksi which is 0.3 inches as stated in the relief request.

5. Figure 3 of Relief Request 14R-04 shows crack growth prediction. (a) Confirm that if a flaw is initiated at 0.15 inches below the toe of the J-groove weld as shown in Figure 3, the upper flaw tip would reach the toe of the J-groove weld after 6 effective full power years. (b) If a fllaw is initiated below the crack growth curve in Figure 3 (e.g., below 0.15 inch location), at what effective full power years will the final flaw tip reach the toe of the J-groove weld?

From: Tsao, John Sent: 12 Oct 2016 06:38:00 -0400 To: Collins, Jay;Alley, David; Hoffman, Keith

Subject:

RE: Wolf Creek RX Vessel Head Pictures - As-Found during Mirror Insulation Removal The wolf creek's RPV head looks like the corroded RPV head at Davis Besse. Do we want to ask wolf creek to send us photos of cleaned RPV head before it can restart so that we can ensure that the RPV head is cleaned to our satisfaction. From: Collins, Jay Se nt: Tuesday, October 11, 2016 4:41 PM To: Tsao, John <John .Tsao@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Hoffman, Keit h

<Keith.Hoffman@nrc.gov>

Subje ct: FW: Wolf Creek RX Vessel Head Pictures - As-Found duri ng Mirror Insulation Removal These were reported to be from Wolf Creek, and also part of my concern. I do not know if you guys have seen these. Jay

From: Alley, David Sent: 14 Oct 2016 00:32:46 +0000 To: Davidson, Evan

Subject:

RE: Input to the DE Periodic Meeting [)ue 10/12 If you are in tomorrow (which will hopefully be today if you are reading this on Friday), lets try to have the meeting we tried to schedule for Wednesday. Dave From: Davidson, Evan Se nt: Wednesday, October 12, 2016 1:54 PM To: Alley, David <David.Alley@nrc.gov>

Subject:

RE: Input to the DE Periodic M eeting Due 10/12 What about your SOI? Any progress or decisions after review ing the applicants? From : Alley, David Sent: Tuesday, October 11, 2016 8:48 AM To: Ross-Lee, M aryJane <MaryJane.Ross-Lee@nrc.gov>; NRR_ DE_DO Distribution

<NRRDEDODistribution@nrc.gov>; Alvarado, Rossnyev <Rossnyev.Alvarado@nrc.gov>

Subject:

RE: Input to the DE Periodic Meeting Due 10/12 EPNB has nothing on the present list which needs to be revised and while we have a bunch of stuff going on, I don't think any of it is quite at the level of the items currently on the list. The closest thing we have is following up on the canopy seal leak at wolf creek. Region is in the lead. There is considerable boric acid on the head. While there is no apparent degradation of the head ala Davis Besse, the inspection process to determine that there are no additional leaks through the nozzles is complicated. There will likely be one or more relief requests. Exact nature of the requests has not yet been determined. Dave From: Ross- Lee, MaryJane Sent: Tuesday, October 11, 2016 8:10 AM To: NRR_DE_DO Distribution <NRRDEDODistribution@nrc.gov>; Alvarado, Rossnyev

<Rossnyev.Alvarado@nrc.gov>

Subject:

FW: Input to the DE Periodic Meeting Due 10/12 Importance: High Reminder that input is due. We meet with Brian on Thursday. Mary Jane Ross-Lee (MJ) Deputy Director, Division of Engineering Off ice of Nuclear Reactor Regulation OWFN 9H1 US Nuclear Regulatory Commission ir Office: 301-415-3298

 ~ e-mail: maryjane.ross-lee@nrc.gov From: Davidson, Evan Sent: Thursday, October 06, 2016 11:23 AM To: NRR_DE_DO Distribution <NRRDEDODistribution@nrc.gov>; Sacko, Fanta <Fanta.Sacko@nrc.gov>

Subject:

Input to the DE Periodic M eeting Due 10/12 The next DE periodic briefing with Brian McDermott is scheduled for October 13. Please take a look at the attached briefing sheet from last month and provide updates for your branch along with suggestions for additions or deletions. If you can get it to me by 10/12 that would be great. I can also stop by to go over the list in person if that's faster. (Today or Wednesday)

Thanks, Evan

From: Collins, Jay Sent: 13 Oct 2016 15:39:56 -0400 To: Tsao, John;Kalikian, Roger

Subject:

FW: Leak Location Attachments: leak location.pdf 1 s-page attachment withheld in full under ex4. Some potential additional nozzles to discuss .... Regional discussion? From : Anchondo, Isaac Sent: Thursday, October 13, 2016 2:41 PM To: Collins, Jay <Jay.Collins@nrc.gov> Cc: Werner, Greg <Greg.Werner@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Kopriva, Ron

<Ron.Kopriva@nrc.gov>; Alley, David <David.Alley@nrc.gov>

Subject:

Leak Location

Jay, I made an attempt to identify the location of the nozzles to be inspected from some of the pictures that Ron took this past Saturday. Unfortunately, it's hard to see the vicinity of some of the nozzles that you question during the call (i.e., 52, 34, 26).

One item of note, the view next to the vent line is a bit confusing where it's hard to figure out which side you are looking at in relation to the drawing provided. Hope this helps. Reactor Inspector U.S. Nuclear Regulatory Commission I Region IV Division of Reactor Safety I Engineering Branch 2 (817) 200-1 152

From: Collins, Jay Sent: 13 Oct 2016 13:53:54 +0000 To: Hoffman, Keith

Subject:

Re: NRC Report for code meetings I wi ll leave that one to Dave. I doubt we are looking at removing 25% sample for all MSIP. I think if any actions are taken it is going to be more focused. In a week or two we will hold another call with Calvert/Westinghouse to discuss the RES weld residual stress analysis. From: Hoffman, Keith Sent: Thursday, October 13, 2016 9:48:06 AM To: Collins, Jay

Subject:

RE: NRC Report for code meetings What about for Calvert Cliffs? This is what I put in the last report. Calvert Cliffs Unit 1 Pressurizer Safety Relief Nozzle to Safe-end Weld LERNo. 3172016002, ADAMS ML16106A304) Ultrasonic (UT) examinations performed at Calvert Cliffs Nuclear Power Plant, Unit 1 identified a change from previous examinations in an axial flaw in a pressurizer safety relief nozzle to safe-end weld that was mitigated by the Mechanical Stress Improvement Process (MSIP) in 2006. Evaluation of the data identified one axially oriented flaw contained within the weld material with a depth measured as 81.6% through-wall including the clad thickness. UT examinations prior to the application of MSIP identified an axial flaw in the same location as the 2016 flaw but a depth of 8% through-wall. UT following MSIP confirmed the flaw was still present at a depth of 8% through-wall. The ISi examinations in 2010 reported essentially no change in the through-wall depth of the indication. Given this information, the NRG is considering rulemaking action to eliminate the allowance of a 25% sample of welds mitigated by MSIP. The NRG is also considering requiring a new baseline examination for welds that have been mitigated by MSIP and have not received an ISi exam in more than ten years. Any recommendations on how I should update this? Keith M. Hoffman Materials Engineer NRR/DE/EPNB (301)415-1294 From: Collins, Jay Sent: Thursday, October 13, 2016 9:43 AM To: Hoffman, Keith

Subject:

Re: NRC Report for code meetings Greetings, The Wolf Creek item should probably include details regarding while this was not a pressure boundary leak, the boric acid leak above the head caused difficulties in performing the head visual inspection. It might be good to wait until after the relief requests are completed t o decide how to write that up.

On the second one, the acceptance criteria is based on the genera l surface exam acceptance criteria as it is a repair and not necessarily a requirement of N-729. Just t he comment. Jay From: Hoffman, Keith Sent: Thursday, October 13, 2016 8:54:12 AM To: Collins, Jay Subject : NRC Report for code meetings Jay I was planning on discussing the two events shown below i n the NRC Report do have any comments on the script below want anything added with regard to these events or any other events. I was going to put this one under operational leakage WOLF CREEK EN 52218 TECHNICAL SPECIFICATION REQUIRED SHUTDOWN While operating in MODE 1 at 100 percent rated t hermal power and placing Excess Letdown in service for Reactor Coolant System (RCS) leak detection, RCS operational leakage exceeded 1 gpm [gallon per minute] unidentified leakage as identified by performing RCS Water Inventory Balance using t he Nuclear Plant Information System Com puter. This requ ired t he Unit to be placed into Mode 3 in 6 hours. Trending of containment sump level indicated the leakage was inside containment with the exact location within containment unknown. The licensee made a containment entry and event ually found t he source of the unidentified leakage. While looking down on the vessel head the licensee identified signs of a boric acid leak over a m irrored insulation panel. After removing the panel and using a camera the licensee saw a plume in the area of several penetrations. The licensee was able to determine that the leak was on a core exit thermocouple nozzle threaded connection. The licensee also determined that this was not pressure boundary leakage. In addition, t he licensee identified t hat excess letdown made t he leak rate seem worse than the actual va lue. The leak rate was eventua lly quantified at around 0.6 gpm. Without being pressure boundary leakage and since the leak rat e was less t han 1 gpm, t he licensee was able to exit the LCO. The licensee has decided to go into their planned refuel ing outage and wi ll perform some pre-outage surveillances before cooling down to MODE 5. The leak will be repaired during the refueling outage while the head is on t he stand. This one I was going to put under RV Head Penetration Inspections BRAIDWOOD 1 (EN 52275)- LIQUID PENETRATION EXAMINATION RESULTS IN INDICATIONS ON REACTOR VESSEL HEAD PENETRATION During the Braidwood Station Unit 1 Refueling outage (A1R19), an inservice Liquid Penetration examination w as performed on the previously repaired control rod drive mechanism (CROM) penetration 69. During the exa mination on the weld build up for CROM penetration 69, two indications were discovered. A 7/32 inch rounded indication was discovered located at 359 degrees on the reactor head portion of the weld buildup, and it is 4 inches from the transition of the head to penetration. A 1/4 inch rounded indication was also discovered located at 200 degrees at the transition of the head to penetration. The transition is the point where t he vertical portion of the penetration meets the horizontal area of the

reactor head. Rounded indications that exceed 3/16 inch are rejectable per ASME Code Case N-729-1. Should they both be under RV Head Penetration Inspections? With a description of what was found @ WC and the difficulties they are having performing the exams and whatever relief they are requesting. Keith M. Hoffman Materials Engineer NRR/DE/EPNB (301)415-1294

From: Singal, Balwant Sent: 13 Oct 2016 16:09:26 -0400 To: Tsao, John;Collins, Jay Cc: Pascarelli, Robert;Alley, David;lingam, Siva

Subject:

FW: WCNOC RV pictures I GUESS THERE WILL BE ANOTHER E-MAIL PROVIDING ACCESS. I WILL REQUEST THEM TO PROVIDE ACCESS TO SIVA AND JAY AS A MINIMUM. THANKS. Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwant.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 From: Good Nicole R [mailto:nilyon@WCNOC.com) Sent: Thursday, October 13, 2016 4:05 PM To: lingam, Siva <Siva.Lingam@nrc.gov> Cc: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

[External_Sender] WCNOC RV pictures I was told you would like pictures of t he penetrations with labels of the penetration number. I have only been able to locate a few pictures, at this point. I have granted you access to the Certrec IMS Sept 2016 Forced Outage. Item #14 has five pictures that may be helpful (DCS00006, DCS00039, DCS00029, DCS00019, and DCS00018). I will need to contact Certrec to get access for Mr. Singal. I will work on getting Mr. Signal access and looking for more pictures tomorrow. Thank you, Nicole Good Licensing nilyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek , Nucleor Operohng Corporol1on

From: Singal, Balwant Sent: 13 Oct 2016 07:32:46 -0400 To: Alley, David Cc: Lingam, Siva; Pascarelli, Robert;Tsao, John;Collins, Jay

Subject:

FW: Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 Attachments: W016-0052R5dt.pdf

Dave, This is a Draft version of the revised relief request and the licensee wants to discuss it with the NRC staff before making it formal. Are we in process to discuss the Draft version and are you ok with it? The licensee wants to have a call at 1.00 PM (Eastern) today. Please confirm your staffs availability so that I can setup the call.

Thanks. Balwant K. Singal Senior Project Manager (Diablo Canyon) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwant.Singa l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222

~ ... ~ .

     ..........)

\~ From: Hafenstine Cynthia R [2] Sent: Wednesday, October 12, 2016 6:11 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>; 'siva.lingman@nrc.gov' <siva.lingman@nrc.gov> Cc: Muilenburg William T <wimuile@WCNOC.com>; Tougaw Dennis E <detouga@WCNOC.com>; Barraclough Richard M <ribarra@WCNOC.com>

Subject:

[External_Sender] Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 Attached is our current draft revision of t he relief request. We have not yet incorporated the questions listed in the draft RAI that you provided. We would like to have a call on Thursday at 1:00 pm Eastern Time/ Noon Central Time. Please let me know if that will work for you. We appreciate your support in getting this document revised to support our request.

Thanks, Cindy Hafenstine Office 620-364-4204

Cleveland Reasoner Site Vice President U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Gentlemen: reactor vessel head penetration nozzles were examine 29-1 and both Wolf Creek Generating Station (WCGS) I nd Boric 1d Programs. A canopy seal weld leak led to the shutdo boric acid accumulation from the canopy seal weld leak he subject of the head inspection. The boric acid te visu inspection of 12 nozzles because the canopy seal weld y bo on from a nozzle leak. Wolf Creek Nuclear Operating dent that the observed deposits were the result of the canopy to nozzle interface areas were obscured such that adequate visual inspections sible on the top side of the head. Because of this, WCNOC will be performing a supple examination of the obscured nozzles from the underside of the head in accordance with C e Case N-729-1. WCNOC is requesting relief from the requirement to perform a surface examination of the partial penetration due to hardship without a compensating increase in the level of quality or safety. Therefore, pursuant to 10 CFR 50.55a(z)(2), WCNOC hereby requests NRC approval of the attached relief request for the WCGS, lnservice Inspection (ISi) Program, fourth ten-year interval. The attachment identifies the affected components, applicable American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME) Code requirements, reason for request, proposed alternative, and basis for proposed alternative. The alternatives are proposed to be applied during Interval 4, which began September 3, 2015 and ends September 2, 2025.

WO 16-0052 Page 2 of 2 The provisions of this relief are applicable to Refueling Outage 21 only. WCNOC will return to the normal inspection protocol for the remainder of ISi Interval 4, which began September 3, 2015 and ends on September 2, 2025 (Reference 1). WC NOC requests approval of this request by October 14, 2016, to support restart from the current refueling outage. In addition, pursuant to 10 CFR 50.55a(z)(2), Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of 10 CFR 50.55a Request Number 14R-04 for the Fourth Ten-Year Interval of WCNOC's lnservice Inspection (ISi) Program. The attached 10 CFR 50.55a Request (14R-04) requests relief from certain ASME Code Case N-729-1 requirements for examination of reactor vessel ead penetration welds. (Attachment 2) This request is similar to that requested in the Ten Year Interval of WCNOC's lnservice Inspection (ISi) that was accepted by ML 12 241 . The Code of Federal Regulations 10 CFR 50.55a(g)(6)(ii)( s that examii nations of the reactor vessel head be performed in accordance with e N-729-1 subject to conditions specified in paragraphs 10 CFR 50.55a (6). The vendor chosen by WCNOC to perform these examinatio d examination coverage below the J-groove weld on two contr etrations. Both of these CROM penetrations are configur s on distance required by N-729-1 cannot be met. Attachment 2 t Request 14R-04, documents the ultrasonic coverage limitations. WCNOC had intended to request this re tions in Refueling Outage 23 but the circumstances described above h

• ion of one of the subject penetrations be performed at this time rin Outa 21 . Therefore, WCNOC requests approval of the attached 10 CFR -04 by October 14, 2016, to support inspection and res om Refuelin is now scheduled to complete November 14, 2016.

questions concerning this matter, please e (620) 364-4204. Sincerely, Cleveland Reasoner COR/rlt Attachments: 1) 10 CFR 50.55a Request Number 14R-03

2) 10 CFR 50.55a Request Number 14R-04 cc: K. M. Kennedy (NRC), w/a B. K. Singal (NRC), w/a N. H. Taylor (NRC), w/a Senior Resident Inspector (NRC), w/a to WO 16-0052 Page 1 of 6 Wolf Creek Nuclear p 10 CFR 50.

ance with (2) to WO 16-0052 Page 2 of 6 10 CFR 50.55a Request Number 14R-03 Relief Requested In Accordance with 10 CFR 50.55a(z)(2) Alternative provides an acceptable level of quality and safety

• ASME Code Component(s) Affected Com anent: Reactor Vessel Closure Head Code Class: Class 1 Examination Cate o 8-P Code Item Number: ative Examination el Upper Heads with ial-Penetration

== Description:==

Size: Material: 2.

                                                            *1er and Pressure Vessel Code (ASME ddenda 50 .55a(g)(6)(ii)(D) 3.

(1) re res that examinations of the reactor vessel head be

                                 'th A    E Code Case N-729-1 subject to the conditions specified a(g)(6)(ii)(D)(2) through (6).

Paragraph -3 The supp/em al examination performed to satisfy -3142.2 shall include volumetric examination f the nozzle tube and surface examination of the partial-penetration weld, (emphasis added) or surface examination of the nozzle tube inside surface, the partial penetration weld, and nozzle tube outside* surface below the weld, in accordance with Fig. 2, or the alternative examination area or volume shall be analyzed to be acceptable in accordance with Appendix I. The supplemental examinations shall be used to determine the extent of the unacceptable conditions and the need for corrective measures, analytical evaluation, or repair I replacement activity. to WO 16-0052 Page 3 of 6

4. Reason for Request

Based on visual examination (VE), deposits resulting from leakage in the canopy seal weld on penetration 77 are on the Reactor Vessel Closure Head. These deposits are dispersed on the reactor head in such a way that it is evident they resulted from the spray pattern, or spray deflection, from the canopy seal weld leak. Other observations noted were: 1) the condition of the head which only had surface rust present rather than wastage; 2) the color and location of these deposits were consistent with spray following the crud burst that was then oxidized by exposure to the atmosphere; 3) there was a layer f white boric acid on top of the deposits in a similar pattern indicating that clean borated ad followed the same path; and 4) no penetrations other than those in the path of spray7deflection show any abnormal indications. ng the condition resulted from the canopy seal weld leak above still obscure the head and prevent the required VE from being ted penetrations. WCNOC will perform supplemental examinations of tions. Twelve penetrations require supplemental examination in accordance with code requirements. Per paragraph -3200(b) of N-729-1 these supplemental examinations

   " ...shall include volumetric examination of the nozzle tube and surface examination of the partial-penetration weld, ... ".

WCNOC does not have the internal resources to conduct the volumetric and surface examinations as required by Code Case N-729 3200(b). A third party vendor has been contracted to perform the examinations. The options for the surface examination of the to WO 16-0052 Page 4 of 6 partial penetration weld are: 1) the dye penetrant technique or 2) the eddy current technique. The dye penetrant technique carries an estimated dose of proximately 1500 mRem (1.5 REM) per nozzle, approximately 18 REM for the entire task. The vendor selected to perform the volumetric examination of the nozzle tube has remotely operated tooling available to perform the surface examination of the partial penetration weld using the Eddy Current technique; however, there are few personnel qualified to operate this equipment. It is estimated that the surface examination of the partial penetration weld using the Eddy Current technique would result in approximately 2.5 Rem of additional exposure. The volumetric examination of the nozzle tube will be performe with remotely operated tooling that is mounted on a manually positioned tool stand rder to perform the supplemental volumetric examination of 12 penetrations, 13 ries under the RV closure head are required . The first entry is estimated to proximately 10 minutes accumulating 408 mRem of exposure. The remain* re estimated to take approximately 2 minutes each yielding 81 mRem of p mRem for a total of 1387 .2 mRem. In order to perform the surface e I penetration weld using the Eddy Current technique, an additional ure head would be required, resulting in a projection of two r leak path assessment and volumetric exam approach. 5. o metric leak path assessment (in addition

                                        *eu of the s    ace examination of the partial penetration erformed in tandem with the nozzle tube aditional dose. This combination (volumetric volumetric leak path assessment) will provide served on the RV closure head were a result of the 1eves that the combination of the volumetric exams and ill pr ide an acceptable level of confidence in the condition of each pe                      ecause, as shown in the figure below, the two examinations will verify there                 ns in the nozzle tube and verify that there has been no leakage in the penetra              VCH interface.

to WO 16-0052 Page 5 of 6 I I I I The volumetric exam is the area fro the interface between the nozzle and weld. Performing the leak path asse groove weld will demonstrate that the head were a result oft opy seal wel ath assessments. The table below lists the eld measured in 2013 using the axially xial coverage was at least 2 inches above 2013 Coverage Obtained Above Weld (Axial Shooting) in inches Combo-2 3.22 OHS 3.64 Combo-2 3.00 Combo-2 3.44 35.2 Combo-2 3.04 35.2 Combo-2 2.76 38.7 Combo-2 2.80 38.7 Combo-2 3.12 44.3 Combo-2 2.60 45.9 Combo-2 2.96 45.9 Combo-2 3.20 77 48.7 OHS 3.32 to WO 16-0052 Page 6 of 6 BE VERY SPECIFIC REGARDING EXAMS TO BE PERFORMED - WHAT PROBES, Zero Degree Etc, and details of coverage - Follow-up comment from our PM The Open Housing Scanner (OHC) uses Type PSC-24 TOFD 5 MHz transducers with a refracted angle of 55° for the circumferential shooting and a refracted angle of 40° for the axial shooting. The Combo-2 blade probes use Type PSC-23.5 TOFD 6.2 MHz transducers with a refracted angle of 57° for the circumferential shooting and a refracted angle of 44° for the axial shooting. In both the OHC and Combo-2 probes, the search units utiliz examination have a nominal frequency of 2.25 MHz. WCNOC has examined the RCVH previously in 2006 n degradation of the RCVH . The data from the e compared to the data from the previous exams. the leak path data. WCNOC believes that the estimated ad REM depending on method used), the added time ace examination of the partial penetration welds fit over the proposed Leak Path Assessment and artial penetration welds does not result in r safety. 6. utilized during WCNOC Refueling Outage 21 only. nor inspection protocol for the remainder of ISi Interval 4, 015 and ends on September 2, 2025. to WO 16-0052 Page 7 of 6

7. Precedent

8. References

1. ASME Boiler and Pressure Vessel Code Case N-729-1 "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Section XI , Division 1"

2. NUREG CR 7142, "Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation

3. WDl-T J-0-03-P, "Ultrasonic Testing of Interference Fit Sam Leak Path Detection (PWROG PA-MSC-0532)"

to WO 16-0052 Page 1of13 Wolf Creek Nuclear to WO 16-0052 Page 2of13 10 CFR 50.55a Request 14R-04 Request for Relief from the Requirements of ASME Code Case N-729-1 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship or Unusual Difficulty Without Compe ting Increase in Level of Quality or Safet

1. ASME Code Components Affected Code Class:

Reference:

Item No.:

Description:

drive mechanism (CROM) J- oove weld that attaches erside of the head for 2. 008 Addenda, as augmented by ASME ative Examination Requirements for PWR Hea ozzles Having Pressure-Retaining Partial-1, Division ,"as amended by 10 CFR 50.55a(g)(6)(ii)(D). 3. (0)(1) requires that examinations of the reactor vessel head be ce with ASME Code Case N-729-1 subject to the conditions s 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6). Paragraph -2 of Code Case N-729-1 states, in part: If obstructions or limitations prevent examination of the volume or surface required by Figure 2 for one or more nozzles, the analysis procedure of Appendix I shall be used to demonstrate the adequacy of the examination volume or surface for each such nozzle. If Appendix I is used, the evaluation shall be submitted to the regulatory authority having jurisdiction at the plant site. to WO 16-0052 Page 3of13 Figure 2 in ASME Code Case N-729-1 , as referenced by paragraph -2500, requires that the volumetric or surface examination coverage d istance below the toe of the J-groove weld (i.e. dimension "a") be 1.5 inches for incidence angle, 8, less than or equal to 30 degrees; 1 inch for incidence angle, 8, greater than 30 degrees; or to the end of the tube, whichever is less. These coverage requirements are applicable to Wolf Creek Generating Station (WCGS) reactor vessel head penetrations as shown in Table 1. Table 1: WCGS Reactor Vessel Head Penetration Coverage Requirements Penetration Numbers *red Coverage, "a" inches 1to29 1.5 30 to 78 1.0

4. Reason for Request styles of ends, referred to gh 73 are Type "Y" that are er diameter and inner diameter.

meter and an internal taper. tion nozzles 74 through 78, referred to as

                                                          , approximately 1.19 inch in length at the re located at the 48.7 degree location. The at th1            1s such that the distance from the lowest point Id to the top of the threaded region could be less than the "a" shown in Figure 2 of ASME Code Case N-729-1.

ired inspection coverage is sought for reactor vessel

                                      , as the required coverage for these two penetrations to WO 16-0052 Page 4 of 13 The table below lists the coverage obtained on nozzles 74-76 during the 2006/2013 exams performed per NRC Order EA-03-009 (2006) and N-729-1 (2013).

Note: The lower measurement in 2006 was performed using circumferential shooting TOFD transducers while the 2013 measurements were accomplished using axial shooting TOFD transducers. While the table below shows different coverage values it is noted in the 201 3 exam report that the "Lower extent comparison using Channel 2 data shows no change from 2006 to 2013 measurements." Penetration 0 (degrees) N-729-1 Required 2013 Inspection No. Exam Coverage Coverage (inches) Obtained inches 74 48.7 1.00 75 48.7 1.08 76 48.7 .00 design weld size or contour is

• side of the peripheral netrations 77 and 78, as weld, resulting in less of the e threads) being available for ssel head penetration welds pelfformed in t was previously submitted for inability to eferences 3 and 4 ). This previous request in For t he examinations performed in 201 3 in

                                     , as conditioned by 10 CFR 50.55a, another similar request fe nces 7, 8, & 9).

5. e volumetric and surface examination coverage requirements "a" in Figure 2 of ASM E Code Case N-729-1, WCGS proposes the ultrasonic examination distances shown in Table 2. The required examination c erage dimension for the other penetrations will be met or exceeded. to WO 16-0052 Page 5of13 Table 2: WCGS Inspection Coverage Obtained for CROM Penetrations Having Limited Coverage Penetration 9 (degrees) N-729-1 2006 Inspection 2013 Inspection No. Required Exam Coverage Coverage Coverage Obtained Obtained inches inches 77 48.7 1.0 .6 78 48.7 1.0 .64 Appendix I of ASME Code Case N-729-1 provides the is procedure for evaluation of an alternative examination area or volume to that Figure 2 of Code Case N-729-1 if impediments prevent examination of the e. Section 1-1000 of ASME Code Case N-729-1 requires, for alter ones below the J-groove weld, that analyses shall be performe analysis method (Section 1-2000) or the deterministic frac d Section !- 3000) to demonstrate that the applicable described in Section 1-2000 were validated in WCAP-1 in WCAP-16589-P was reviewed. The stress a over the entire region outside t 16589-P analysis was compared t it was determined that the require analysis in WCAP-16589-P was also 3000. Since the lternative examination zones that eliminate portions of Figure 2 w the J-groove weld, that 1-1000 requires only the analysis 00 or 1-3000 to be performed Although not required, the mechanics analysis described in Section 1-3000 was also validated 5.1 Stress Analysis in Accordance with ASME Code Case N-729-1 Section 1-2000 Section 1-2000 of ASME Code Case N-729-1 requires that plant-specific analysis demonstrate that the hoop and axial stresses remain below 20 kips per square inch (ksi) (tensile) over the entire region outside the alternative examination zone but within the examination zone defined in Figure 2 of the Code Case. to WO 16-0052 Page 6of13 The distance below the J-groove weld that requires examination, as determined by the point at which the CROM penetration hoop stress distribution for the operating stress levels is less than 20 (ksi) tension, was obtained from Appendix A of Reference 2. Note that hoop stresses during steady state operation are much greater than the axial stresses. The hoop stress distribution plots for penetrations 77 and 78 are provided in Figure 2 of this submittal. The hoop stress distribution plots in Figure 2 indicate that the minimum achievable inspection coverage below the bottom of the J-gr weld insures stresses remain below 20 ksi tensile over the entire region outsid zone but within the examination zone defined in Figure The hoop stress distribution plots display the downhill *sis more limiting. Also, stress distribution plots shown are for the insid surface. Table 3 summarizes the distance from below the toe of t o both the inside and outside surface hoop stre Ciro and 78. Penetration Distance Below Toe of Nozzle No. i Side J-Groove Weld ere Hoop Stress

                                                                           < 20 ksi inch 0.30 5.2 ed and documented in Reference 2.       The ta po           ral crack in the unexamined zone will not grow d prior to the examirnation frequency specified in Table 1 of wa     repared prior to approval to use Code Case N-729-1.

ced EPRI MRP-55 as the source for the crack growth formula used i t Appendix 0 as required by Code Case N-729-1 . However, since the for la for crack growth rate is used in both EPRI MRP-55 and Appendix rs no technical difference, and WCAP-16589-P does meet the technical requr ents for l-3200(a).) (1) The following table provides the dimensions for nozzles 77 and 78 for both the designed and as-built configurations. The actual weld height was measured using the ultrasonic test data and is listed for the as-built dimension. to WO 16-0052 Page 7of13 Penetration Nozzle Number As-designed As-built (inches) (inches) 77 1.46 1.98 78 1.46 2.04 The flaw evaluation in WCAP-16589-P is based on the as-designed J-groove weld dimensions which assumed a smaller weld throat than the as-built condition. Often, the as-built fillet weld dimension on the downhill side of the CRD ozzle is larger than the

                                                                                  'on. When the weld tion nozzle due to a larger below the J-groove weld on similar CRDM ds have a reduced e below the weld weld heights of s profiles input from the previously own as flaw tolerance require inspection coverage. This of the penetration nozzle not before the next inspection. The ented in Figure 3.

tes that a postulated through-wall flaw at xamination zone will not grow to the toe of al of four refueling cycles. The crack growth n six e e full power years (EFPY) of operation required the toe of the weld. Additionally, the assumed initial upper h-wall flaws are conservative based on achievable e assumed upp.er crack extremities are located within of port' nozzle significantly below the J-groove weld is not omena of concern, which include leakage through the J-groove ntial cracking in the nozzle above the J-groove weld. In all cases, erage is adequate to allow WCGS to continue to operate prior to the hypothetical aws reaching the J-groove weld. In accordance with 10 CFR 50.55a(g)(6)(ii)(D) requirements, the next required examination would be completed prior to potential flaw propagation into the J-groove welds. 5.3 Surface Examination 10 CFR 50.55a(g)(6)(ii)(D)(3) states in part that "if a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is to WO 16-0052 Page 8of13 below the toe of the J-groove weld , the surface examination shall be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically." To reduce personnel radiation exposure, the nozzles are typically inspected using remotely operated volumetric examination equipment. Although dye penetrant testing of threaded surfaces is possible, it is not practical. The threaded outside diameter (OD) makes a dye penetrant examination on the lower section of the penetration impractical because of excessive bleed out from the threads. Eddy current examination would similarly not be effective due to the threaded configuration. rrent known radiation levels under the reactor vessel head are 4.5 Rem/hr at th o of 1 CROM nozzle. This could result in an exposure of approximately 1. per nozzle using 4500 mRem/hr and 20 minutes/nozzle. At this time our e f dose rates (based on recent measurements in the area) range from 4.5 R m/hr at the bottom of the CROM nozzles, the expected dose ranges fr 5 Rem to 3.3 Rem per nozzle to perform surface examination. Therefore, no alternative is proposed for coverage below the J-groove weld. 6. The alternati will be applied for the remaining duration of the current ~th I) Interval which ends on September 2, 2025. 7. granted to the following plants: Safety dated December 22, 2009, for San Onofre Nuclear ting S ion, Units 2 and 3, "Relief Request ISl-3-29, Request for Relief e n Requirements of ASME Code Case N-729-1 for Control Element chanism Penetrations (TAC Nos. ME0768 and ME0769)"

                       <:i 1035)

• NRC Safety Evaluation dated March 3, 2011, for Braidwood Station Units 1 and 2, and Byron Station Units 1 and 2, "Relief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds (TAC Nos. ME3510, ME3511 , ME3512 and ME3513)" (ML110590921) to WO 16-0052 Page 9of13
• NRC Safety Evaluation dated January 4, 2013, for Wolf Creek Generating Station, "Wolf Creek Generating Station - Request for Relief No. 13R-07 for the Third 10-Year lnservice Inspection Program Interval (TAC No. ME9078) to WO 16-0052 Page 10 of 13

8. References

1. ASME Code Case N-729-1 , "Alternative* Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section XI, Division 1," March 28, 2006.

2. WCAP-16589-P, Revision 0, "Structural Integrity Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: Wolf Creek," August 2006.

3. WCNOC letter ET 06-0035 from T. J. Garrett, W Request from the First Revised NRC Order E for Nondestructive Examination of Nozzles 2006.

4. WCNOC letter ET 06-0048 from T. C, "Additional Information Related to the Firs egarding Requirements for Nondestructive Ex e J-Groove,"

November 1, 2006.

5. NRC letter from D. Ter uench, WCNOC, "Wolf Creek Generating Station - Req ive Examination of Reactor Pressure Vessel Head Pe evised Order EA-03-009 (TAC NO. MD3210)," Dece

6. evised NRC Order (EA-03-009) equirements For Reactor Pressure Vessel s," February 20, 2004 .

7. . Broschak, WCNOC, to USNRC, "10 CFR Relief from ASME Code Case N-729-1 T 12- 24 from J. P Broschak, WCNOC, to USNRC, "Response Additional Information Regarding 10 CFR 50.55a Request

                              " Relief from ASME Code Case N-729-1 Requirements for eactor Vessel Head Penetration Welds," October 15, 2012.

9. rom M. T. Markley, USNRC, to M. W. Sunseri, WCNOC, "Wolf Creek Gener ng Station - Request for Relief No. 13R-07 for the Third 10 Year lnseNice Inspection Program lnteNal (TAC NO. ME9078)," January 4, 2013.

to WO 16-0052 Page 11of13 Figure 1 WCGS Reactor Vessel Head Penetration Ends

                            ~   0Eii'
                                 .     ..      ','I
                                 ;,_   ;>  I      I

~ /

                                     ' '*'~---'

Set lltl~I_ 'J" 8 ._1,~>'"' / *in'

                                            '""'/ ~,20' 1 :.oo *+/- 0 O"Dl.'1 nd"?: portions of Penetrations 74, 75, 76, 77, and 78 referred to of Penetrations 1 through 73, referred to as Type Y."

to WO 16-0052 Page 12 of 13 Figure 2 Hoop Stress Distribution Downhill Side (48.7° CRDM Peaetratlon Nozzle) 80,000 - - - - - - . . . - - - - - - . . - - - - - - - . . . . . - - - - - - . . . . . - - - - ---. I I 70,000 ************ J *********

• I ** ' *- **
• I '** --- *****--*

I I 60.000 **------*----~---------- - -- *------ - ------~------------~------------- I ' I t t 50,000 *********,- ************r****-*-*-***r*-*********-,******-*- * *** I I I I L*- * ------ -- -~ - ---- * -----

• J- * *--------- *

     =-*  .C0,000         *****

• J ********** * *
• I I t I

    ~
    !U) a 30,000 20,000            ... ,..*.... . . ..

I

                                                                 - ~--- * *- - ** ** **-** * -*

I I ' I

                                                                                                               -- --- ~--- -- - -----**

I 0 0 I I I x 10,000 *** ********* L ************ l ************ t I

                                                                                                                         ~

I I

         *10,000 0
                                                                   .'--*-***---*-*r**----------,***
                                   . . . . - .. - .. - - * * - - * ** * * * - - .. - - - - - * * * - - * - * - - * - * - -4.

I I I

         *20.000                   J-------------L-----*-*****
         *30,000 ..__ ___..__...__,__ _ _ _ _ _ _                   ' _ _ _ _ _ _ _ _ _ _ _ _ _ _ __...._ _ _-1 00               0.5                            10                           1   s                      20                         25 Distance from Bottom of Weld (In)

I-+- Inside -o- outside I to WO 16-0052 Page 13of13 Figure 3 Crack Growth Prediction for WCGS for Through-Wall Longitudinal Flaws Located in the 48.7' Row of CROM Penetrations, Downhill Side {: ! 9 j ___ ,_ 1-TC~"c~1c:.c-cc~-{cc-*ccc:" ~ ~

                                                                                 ~+"~~"'" j            m
                                                                                                          ~

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                                                    ** 1.

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                                                             -1  )"

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From: Singal, Balwant Sent: 13 Oct 2016 13:59:46 -0400 To: Collins, Jay;Tsao, John Cc: Alley, David;Lingam, Siva;Pascarelli, Robert

Subject:

FW: FW: Relief Request for Code Case N-729-1 Attachments: M-706-00009_REACTOR PEN.JPG Finally, the e-mail came through . Balwant K. Singal Senior Project Manager (Diablo Canyon) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwa nt.Si nga l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 From: Hafenstine Cynthia R [mailto:cyhafen@WCNOC.com) Sent: Thursday, October 13, 2016 1:01 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

[External_Sender] FW: Relief Request for Code Case N-729-1 New drawing for the draft relief request... From: Barraclough Richard M Sent: Thursday, October 13, 2016 11:53 AM To: Hafenstine Cynthia R Cc: Tougaw Dennis E

Subject:

Relief Request for Code Case N-729-1 This is the image I had Salvador Ferrara put together for t he relief request R. Mark Barraclough Wolf Creek Nuclear Boric Acid Engineer I Program Owner Fluid Leak Management I Program Owner AOV Engineer 620-364-8831 x8148 I ribarra@wcnoc.com Fax: 620-364-4154

From: Singal, Balwant Sent: 13 Oct 2016 09:33:18 -0400 To: Co llins, Jay;Tsao, John;Lingam, Siva; Pascarelli, Robert;Alley, David;Taylor, Nick;Proulx, David;Drake, James;Werner, Greg;Anchondo, lsaac;Kopriva, Ron Subje ct: Wolf Creek Relief Request Attachme nts : [External_ Sender] Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 Please see the attached e-mail from WCNOC with a draft version of the revised relief request. WCNOC has requested a call to discuss the reliefrequest with the NRC staff before issuing the formal request. The relief is needed in support of the current refueling outage. Bridge No. for the call. 866-624-3402 Passcode:l(b)(G) I# Thanks.

From: Hafenstine Cynthia R Sent: 12 Oct 2016 22:10:42 +0000 To: Singal, Balwant;'siva.lingman@nrc.gov' Cc: Muilenburg William T;Tougaw Dennis E;Barraclough Richard M

Subject:

[External_Sender) Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 Attachme nts: W016-0052R5dt.pdf Attached is our current draft revision of the relief request. We have not yet incorporated the questions listed in the draft RAI that you provided. We would like to have a call on Thursday at 1:00 pm Eastern Time I Noon Central Time. Please let me know if that will work for you. We appreciate your support in getting this document revised to support our request.

Thanks, Cindy Hafenstine Offr 620-364-4204 Cell (bX6J I

From: Tsao, John Sent: 13 Oct 2016 08:14:02 -0400 To: Collins, Jay;Singal, Balwant;Alley, David

Subject:

RE: Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 I am available at 1 Pm also From : Collins, Jay Sent: Thursday, October 13, 2016 8:07 AM To: Singal, Balwant <Balwant.Singal@nrc.gov>; Alley, David <David.Alley@nrc.gov> Cc: Tsao, John <John.Tsao@ nrc.gov>

Subject:

RE: Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 I am available at lpm. I have a call at 2pm. From: Singal, Balwant Sent: Thursday, October 13, 2016 7:33 AM To: Alley, David <David.Alley@nrc.gov> Cc: Lingam, Siva <Siva .Lingam@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov>

Subject:

FW: Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052

Dave, T his is a Draft version of the revised relief request and the licensee wants to discuss it with the NRC staff before making it formal. Are we in process to discuss the Draft version and are you ok with it? The licensee wants to have a call at 1.00 PM (Eastern) today. Please confirm your staff's availability so that I can setup the call.

Thanks. Balwant K. Singal Senior Project Manager (Diablo Canyon) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwant.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 i~ - \~ .... ) From : Hafenstine Cynthia R [3] Sent: Wednesday, October 12, 2016 6:11 PM To: Singal, Balw ant <Balwant.Singal@nrc.gov>; 'siva.lingman @nrc.gov' <siva.lingman@nrc.gov> Cc: Muilenburg William T <wimuile@WCNOC.com>; Tougaw Dennis E <detouga@WCNOC.com>;

Barraclough Richard M <ribarra@WCNOC.com>

Subject:

[External_Sender] Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 Attached is our current draft revision of the relief request. We have not yet incorporat ed the questions listed in the draft RAI that you provided. We would like to have a call on Thursday at 1:00 pm Eastern Time I Noon Central Time. Please let me know if that will work for you. We appreciate your support in getting this document revised to support our request.

Thanks, Cindy Hafenstine Office 620-364-4204 Cell l(b)(6) I

From: Tsao, John Sent: 13 Oct 2016 14:40:23 +0000 To: Singal, Balwant

Subject:

Accepted: Wolf Creek Relief Request

From: Burkhardt, Janet Sent: 14 Oct 2016 06:09:33 -0400 To: FRN_ ReviewRequest Cc: Mendiola, Doris;Lingam, Siva

Subject:

FW: WCGS LT FRN MF8168 Attachments: MF8168-FRN.docx Importance: High Also, can you please remove the red footnote when you return the clean copy? We cannot listserv a copy to the licensee like that when it's signed and dated. Thank you. From : Singal, Balwant Sent: Thursday, October 13, 2016 3:14 PM To: FRN_ReviewRequest <FRN_ReviewRequest@nrc.gov> Cc: Burkhardt, Janet <Janet.Burkhardt@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>

Subject:

FW: WCGS LT FRN Mf8168 Importance: High Attached is the copy of the FRN for early FRN review. Thanks. Balwant K. Singal Senior Project Manager (Diablo Canyon) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwa nt.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222

    ~*

From: Burkhardt, Janet Sent: Thursday, October 13, 2016 3:01 PM To: Singal, Balwant <Balwant.Singal@nrc.gov> Cc: Lingam, Siva <Siva .Lingam@nrc.gov>

Subject:

WCGS LT FRN MF8168 Importance: High

Balwant, Please forward the att ached to FRN ReviewRequest@nrc.gov for early FRN review (cc me and Siva). I'll return the hard copy package to Siva tomorrow after I reprint it .

Please note: if you are not going to be here to sign the FRN, the signatu re block on the FRN (only) will need to be changed to the person signing it.

[7590-01-P] NUCLEAR REGULATORY COMMISSION Docket No. 50-482; NRC-2016-XXXX WOLF CREEK GENERATING STATION Consideration of Approval of Transfer of License AGENCY: Nuclear Regulatory Commission. ACTION: Application for indirect transfer of license; opportunity to comment, request a hearing, and petition for leave to intervene.

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) received and is considering approval of an indirect license transfer application filed by Wolf Creek Nuclear Operating Company (WCNOC) on July 22, 2016. WCNOC is the licensed operator of Wolf Creek Generating Station (WCGS). Kansas City Power and Light Company (KCP&L) and Kansas Gas and Electric Company (KG&E) are two of the three non-operating owner licensees, each holding 47 percent undivided interest in WCGS and 47 percent of the stock of WCNOC. KCP&L is a subsidiary of Great Plains Energy Incorporated (Great Plains) and KG&E is a subsidiary of Westar Energy Incorporated (Westar). The indirect license transfer will result from the proposed merger of Great Plains and Westar, with Westar as wholly-owned subsidiary of Great Plains.

Every attempt has been made to make these templates all-inclusive. However, every FRN may involve case-specific issues related to it, requiring certain modifications of this template. If you believe that a particular part of this template does not apply to your specific rule please check with the ADM/RADS Regulations Specialist assigned to your working group or e-mail FRN ReviewRequest@nrc.gov.

DATES: Comments must be filed by [INSERT DATE 30 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER] . A request for a hearing must be filed by [INSERT DATE 20 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER] . ADDRESSES: You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

• Federal Rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-XXXX. Address questions about NRC dockets to Carol Gallagher; telephone: 301 -415-3463; e-mail: Carol.Gallagher@nrc.gov. For technical questions contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
• E-mail comments to: Hearingdocket@nrc.gov. If you do not receive an automatic e-mail reply confirming receipt, then contact us at 301 -415-1677.
• Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 301-415-1101.
• Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
• Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal workdays; telephone: 301-415-1677.

For additional direction on obtaining information and submitting comments, see "Obtaining Information and Submitting Comments" in the SUPPLEM ENTARY INFORMATION section of this document. 2

FOR FURTHER INFORMATION CONTACT: Balwant K. Singal, Office of Nuclear Reactor Regulation, telephone: 301-415-3016, e-mail: Balwant.Singal@nrc.gov; U.S. Nuclear Regulatory Commission, Washington DC 20555-0001. SUPPLEMENTARY INFORMATION: I. Obtaining Information and Submitting Comments A. Obtaining Information Please refer to Docket ID NRC-2016-XXXX when contacting the NRC about the availability of information for this action. You may obtain publicly-available information related to this action by any of the following methods:

• Federal rulemaking Web Site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-XXXX.
• NRC's Agencywide Documents Access and Management System (ADAMS):

You may obtain publicly-available documents online in the ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select "ADAMS Public Documents" and then select "Begin Web-based ADAMS Search." For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to pdr.resource@nrc.gov. The application for indirect transfer of the license dated July 22, 2016, is available in ADAMS at Accession No. ML16208A250.

• NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room 01-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

3

B. Submitting Comments Please include Docket ID NRC-2016-XXXX in the your comment submission. The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information. If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS. II. Introduction The NRC is considering the issuance of an order under§ 50.80 of title 10 of the Code of Federal Regulations (1 O CFR) approving the indirect transfer of control of WCGS, Facility Operating License No. NPF-42, currently held by WCNOC . WCNOC is the licensed operator of WCGS. KCP&L and KG&E are two of the three non-operating owner licensees, each holding 4 7 percent undivided interest in WCGS and 47 percent of the stock of WCNOC. Kansas Electric Power Cooperative, Incorporated (KEPCo) holds rest of the 6 percent undivided interest in WCGS and 6 percent of the stock of WC NOC. KCP&L is a subsidiary of Great Plains Energy Incorporated (Great Plains) and KG&E is a subsidiary of Westar Energy Incorporated (Westar). The indirect license transfer will result from the proposed merger of Great Plains and Westar, 4

with Westar as wholly-owned subsidiary of Great Plains. The current and intended ownership structure of the facility is depicted in simplified organization chart provided in Figures 1 and 2 of the letter dated July 22, 2016. KCP&L and KG&E will each continue to hold their respective 47.0 percent interests in WCNOC and WCGS. KCP&L and KG&E will continue to operate as separate electric utilities responsible for their pro rata shares of the costs of operating WCGS and entitled to their pro rata shares of the capacity, energy and other energy products produced by WCGS. Great Plains will indirectly own a combined interest in WCGS of 94.0 percent. WCNOC will continue to be the operator of WCGS . The remaining 6.0 percent ownership interest of KEPCo is not affected by the Merger. No physical changes to the WCGS or operational changes are being proposed in the application. The NRC's regulations at 10 CFR 50.80 state that no license, or any right thereunder, shall be transferred, directly or indirectly, through transfer of control of the license, unless the Commission gives its consent in writing. The Commission will approve an application for the indirect transfer of a license, if the Commission determines that the proposed merger of Great Plains and Westar will not affect the qualifications of the licensee to hold the license, and that the transfer is otherwise consistent with applicable provisions of law, regulations, and orders issued by the Commission Ill. Opportunity to Comment Within 30 days from the date of publication of this notice, persons may submit written comments regarding the license transfer application, as provided for in 10 CFR 2.1305. The Commission will consider and, if appropriate, respond to these comments, but such comments 5

will not otherwise constitute part of the decisional record. Comments should be submitted as described in the ADDRESSES section of this document. IV. Opportunity to Request a Hearing and Petition for Leave to Intervene Within 20 days after t he date of publication of this notice, any persons (petitioner) whose interest may be affected by this action may file a request for a hearing and a petition to intervene (petition) with respect to the action. Petitions shall be filed in accordance with the Commission's "Agency Rule*s of Practice and Procedure" in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.309, which is available at the NRC's PDR, located at One White Flint North, Room 01-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The NRC's regulations are accessible electronically from the NRC Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a petition is filed within 20 days, the Commission or a presiding officer designated by the Commission or by the Chief Administrative Judge of the Atomic Safety and Licensing Board Panel, will rule on the petition; and the Secretary or the Chief Administrative Judge of the Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order. As required by 10 CFR 2.309, a petition shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following general requirements: (1) the name, address, and telephone number of the petitioner; (2) the nature of the petitioner's right under the Act to be made a party to the proceeding; (3) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (4) the possible effect of any decision or order 6

which may be entered in the proceeding on the petitioner's interest. The petition must also set forth the specific contentions which the petitioner seeks to have litigated at the proceeding. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases for the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion to support its position on the issue. The petition must include sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the proceeding. The contention must be one which, if proven , would entitle the petitioner to relief. A petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party. Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing with respect to resolution of that person's admitted contentions consistent with the NRC's regulations, policies, and procedures. Petitions for leave to intervene must be filed no later than 20 days from the date of publication of this notice. Requests for hearing, petitions for leave to intervene, and motions for leave to file new or amended contentions that are filed after the 20-day deadline will not be entertained absent a determination by the presiding officer that the filing demonstrates good cause by satisfying the three factors in 10 CFR 2.309(c)(1 )(i) through (iii). A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof, may submit a petition to the Commission to participate as a party under 10 CFR 2 .309(h}(1 ). 7

The petition should state the nature and extent of the petitioner's interest in the proceeding. The petition should be submitted to the Commission by [INSERT DATE 20 DAYS FROM THE DATE OF PUBLICATION IN THE FEDERAL REGISTER] . The petition must be filed in accordance with the filing instructions in the "Electronic Submissions (E-Filing)" section of this document, and should meet the requirements for petitions set forth in this section, except that under 10 CFR 2.309(h)(2) a State, local governmental body, or Federally-recognized Indian Tribe, or agency thereof does not need to address the standing requirements in 10 CFR 2.309(d) if the facility is located within its boundaries. A State, local governmental body, Federally-recognized Indian Tribe, or agency thereof may also have the opportunity to participate under 10 CFR 2.315(c). If a hearing is granted, any person who does not wish, or is not qualified, to become a party to the proceeding may, in the discretion of the presiding officer, be permitted to make a limited appearance pursuant to the provisions of 10 CFR 2.315(a). A person making a limited appearance may make an oral or written statement of position on the issues, but may not otherwise participate in the proceeding. A limited appearance may be made at any session of the hearing or at any prehearing conference, subject to the limits and conditions as may be imposed by the presiding officer. Details regarding the opportunity to make a limited appearance will be provided by the presiding officer if such sessions are scheduled. V. Electronic Submissions (E-Filing) All documents filed in NRC adj udicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene (hereinafter "petition"), and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; August 28, 2007, as amended 8

at 77 FR 46562, August 3, 2.012). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request ( 1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition (even in instances in which the participant, or its counsel or representative, already holds an NRG-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket. Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-he/ple-submittalslgetting-started.html. System requirements for accessing the E-Submittal server are available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/adjudicatorv-sub.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Electronic Filing Help Desk will not be able to offer assistance in using unlisted software. Once a participant has obtained a digital ID certificate and a docket has been created, the participant can then submit a petition. Submissions should be in Portable Document Format (PDF). Additional guidance on PDF submissions is available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the time the documents are submitted through the NRC's E-Filing system. To be timely, an 9

electronic filing must be submitted to the E-Filing system no later than 11 :59 p.m. Eastern Time on the due date. Upon receipt of a transmission, the E-Filing system time-stamps the document and sends the submitter an e-mail notice confirming receipt of the document. The E-Filing system also distributes an e-mail notice that provides access to the document to the NRC's Office of the General Counsel and any others who have advised the Office of the Secretary that they wish to participate in the proceeding, so that the filer need not serve the documents on those participants separately. Therefore, applicants and other participants (or their counsel or representative) must apply for and receive a digital ID certificate before a hearing petition to intervene is filed so that they can obtain access to the document via the E-Filing system. A person filing electronically using the NRC's adjudicatory E-Filing system may seek assistance by contacting the NRC Electronic Filing Help Desk through the "Contact Us" link located on the NRC's public Web site at http:llwww.nrc.gov/site-he/ple-submittals.html, bye-mail to MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The NRC Electronic Filing Help Desk is available between 9 a.m. and 7 p.m., Eastern Time, Monday through Friday, excluding government holidays. Participants who believe that they have a good cause for not submitting documents electronically must file an exemption request, in accordance with 10 CFR 2.302(g), with their initial paper filing stating why there is good cause for not filing electronically and requesting authorization to continue to submit documents in paper format. Such filings must be submitted by: (1) first class mail addressed to the Office of the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) courier, express mail, or expedited delivery service to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11 555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Participants filing a document in this manner are responsible for serving the document on all other participants. Filing is considered 10

complete by first-class mail as of the time of deposit in the mail, or by courier, express mail, or expedited delivery service upon depositing the document with the provider of the service. A presiding officer, having granted an exemption request from using E-Filing, may require a participant or party to use E-Filing if the presiding officer subsequently determines that the reason for granting the exemption from use of E-Filing no longer exists. Documents submitted in adjudicatory proceedings will appear in the NRC's electronic hearing docket which is available to the public at http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the Commission, or the presiding officer. Participants are requested not to include personal privacy information, such as social security numbers, home addresses, or home phone numbers in their filings, unless an NRC regulation or other law requires submission of such information. However, in some instances, a petition will require including information on local residence in order to demonstrate a proximity assertion of interest in the proceeding. With respect to copyrighted works, except for limited excerpts that serve the purpose of the adjudicatory filings and would constitute a Fair Use application, participants are requested not to include copyrighted materials in their submission. The Commission will issue a notice or order granting or denying a hearing request or intervention petition, designating the issues for any hearing that will be held and designating the Presiding Officer. A notice granting a hearing will be published in the Federal Register and served on the parties to the hearing. 11

For further details with respect to this application, see the application dated July 22, 2016. Dated at Rockville, Maryland, this day of October 2016. For the Nuclear Regulatory Commission. Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 12

From: Collins, Jay Sent: 14 Oct 2016 19:39:58 +0000 To: Alley, David

Subject:

Fw: RE: WCNOC RV pictures Greetings, It looks like you will get access. The numbering is a bit confusing. Once you get connected, in the first folder is a list of pictures. The file titled, Pen 67 & 54 DSC00068, seems to include a picture of penetration nozzles 67 and 54. They also appear to be labeled in the picture. I believe this is a view that Isaac gave us previously, but not the same photo. Note that the vent line in the picture between t he two penetration nozzles. Now, if you go to the head map image listed as M -706-00009_REACTOR PEN in the folder, you w ill find that penetration nozzles numbers 67 and 54 are no where near nozzle 77, the source of the spi ll. Instead, number 67 is at approximately 320 degrees and nozzle 54 is at 290 degrees, near the periphery, in the south-w est quad rant of the head. Note also, they are not near the head vent line, which is at about the 45 degree location in the North-west quadrant of the head. I believe nozzle 67 is the nozzle 76 that Isaac circled with a question mark in the images he sent on Thu rsday. Either way, 67 or 76, it has remaining indications in the annulus region and is not listed for volumetric inspection. I will look over the photos this weekend, but i think we will perhaps need an internal discussion on Monday for a bit. Jay From: Good Nicole R Sent : Friday, October 14, 2016 10:25 AM To: Singal, Balwant; Lingam, Siva Cc: Lingam, Siva; Collins, Jay; Tsao, John; Alley, David; Pascarelli, Robert

Subject:

[External_ Sender] RE: WCNOC RV pictures Access has been provided to: Siva Lingam Jay Collins John Tsao Access has been requested for: Balwant Singal David Alley Robert Pascarelli Thank you, Nicole Good From: Singal, Balwant [4] Sent: Thursday, October 13, 2016 3:12 PM To: Good Nicole R; Lingam, Siva

Cc: Lingam, Siva; Collins, Jay; Tsao, John; Alley, David; Pascarelli, Robert

Subject:

RE: WCNOC RV pictures

Nicole, Not clear if we already have the access or will be getting access later. Please have access to following two persons as a minimum:

Siva Lingam Jay Collins Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwant.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222

    ~*

From: Good Nicole R [mailto:nilyon@WCNOC.com) Se nt: Thursday, October 13, 2016 4:05 PM To: Lingam, Siva <Siva.Lingam@nrc.gov> Cc: Singal, Balwant <Balwant .Singal@nrc.gov>

Subject:

[External_Sender] WCNOC RV pictures I was told you would like pictures of the penetrations with labels of the penetration number. I have only been able to locate a few pictures, at this point. I have granted you access to the Certrec IMS Sept 2016 Forced Outage. Item #14 has five pictures that may be helpful (DCS00006, DCS00039, DCS00029, DCS00019, and DCS00018). I will need to contact Certrec to get access for Mr. Singal. I will work on getting Mr. Signal access and looking for more pict ures tomorrow. Thank you, Nicole Good Licensing nilyon@wcnoc.com (620) 364-8831x4557 W oIf Cree k..u...*.,,"""* Nuclear Operofing Corporation

From: lingam, Siva Sent: 14 Oct 2016 21:03:06 -0400 To: Tsao, John;Collins, Jay Cc: Pascarelli, Robert;Alley, David;Singal, Balwant

Subject:

RE: Wolf Creek Relief Requests The licensee wants our approval (obviously verbal) by 10/17/16, as discussed during the phone call on Thursday. From : lingam, Siva Sent: Friday, October 14, 2016 9:00 PM To: Tsao, John <John.Tsao@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov> Subje ct: FW: Wolf Creek Relief Requests Attached please find the revised RRs from the licensee for your review/comment s/evaluation. Thank you. From : Muilenburg William T [5] Sent: Friday, October 14, 2016 6:42 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>; 'nick.taylolr@nrc.gov' <nick.taylolr@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabia11 .Thomas@nrc.gov>; Lingam, Siva

<Siva. Li nga m@nre.gov>

Subject:

[External_Sender) Wolf Creek Relief Requests

Everyone, Attached are Wolf Creek Relief Requests 14R-03 and 04.

Bill Muilenburg, 620-364-4186

From: Alley, David Sent: 17 Oct 2016 16:51:25 +0000 To: Tsao, John

Subject:

FW: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment

John, These are the words Jay put in The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2),

such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. From : Al ley, David Se nt: Monday, October 17, 2016 11:58 AM To: Collins, Jay <Jay.Collins@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Singal, Balwant

<Ba lwant.Singal@nrc.gov>

Cc: Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov> Subje ct: RE: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment John , Jay, Both verbals appear well written . I do have one question . In a normal SE in the first or second paragraph we say that the licensee is proposing its alternative in accordance with 50.55a(z)(1) (or in this case (z)(2)) or words to that effect. In the scripts I need to get all the way to the end to officially find out that these are (z)(2) requests. Do we normally put the authority for the proposed alternative up front? Dave From : Collins, Jay Sent: M onday, October 17, 2016 11:24 AM To: Tsao, John <John .Tsao@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert

<Robert.Pascarelli@nrc.gov>

Subje ct: RE: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment Greetings, 1st Draft of the script for 14R-03. Please provide me comments when you can. Jay From : Tsao, John Se nt: Thursday, October 13, 2016 2:09 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert

<Robert.Pascarelli@nrc.gov>

Subject:

Wolf Creek: Verbal Authorization script for Relief Request 14R-04 Alternate CRDM nozzle examinations Attached is the draft verbal authorization script for Relief Request 14R-04 regarding alternate examinations of CROM nozzles at Wolf Creek. I may change the script slightly after the licensee submits the final version of the relief request. Please review and provide changes. The attached verbal authorization does not include Relief Request 14R-03 which Jay is working on. Jay, if you want I can prepare the script for your relief request, 14R-03.

From: Collins, Jay Sent: 17 Oct 2016 12:01:55 +0000 To: Tsao, John;Cumblidge, Stephen;Alley, David

Subject:

Wolf Creek - Potential additional nozzles require volumetric inspection Attachments: wolfcreekHead2016.pdf One-page attachment withheld in fulll under ex4. Greetings, Attached is a small sized slide package of some pictures of the Wolf Creek Head during the Fall 2016 refueling outage. A leak from above provided significant boric acid deposits on the head surface, which interfered with the licenseea's ability to perform an effective bare metal visual examination. The licensee identified 12 nozzles that would require further volumetric examination. However, in reviewing the attached photos, I believe there are additional nozzles that should be considered, or at least further explanation should be provided by the licensee, if a volumetric examination is not planned to be performed this refueling outage. We are setting up an internal call this morning with the Region to discuss the issue. Jay

From: Collins, Jay Sent: 17 Oct 201617:57:04 +0000 To: Cumblidge, Stephen

Subject:

BMV for upper heads Attachments: BMV for upper heads.pdf http://www.epri.com/abstracts/Pages/Prod uctAbst ract.aspx?Productld =000000000001007842 . I think there is a Rev 3 for th is now ....

EPf21 Visual Examination for Leakage of PWR Reactor Head Penetrations Revision 2 of 1006296, Includes 2002 Inspection Results and MRP Inspection Guidance WARNING: Please read the License Agreement on t he back cover before removing the Wrapping Material. Technical Report

From: Collins, Jay Sent: 17 Oct 2016 15:23:59 +0000 To: Tsao, John;Singal, Balwant Cc: Alley, David;Lingam, Siva;Pascarelli, Robert

Subject:

RE: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment Attachments: Wolf Creek verbal auth 14R-03 10-17-2016.docx Greetings, 1st Draft of the script for 14R-03. Please provide me comments when you can. Jay From : Tsao, John Sent: Thursday, October 13, 2016 2:09 PM To: Singal, Balwant; Collins, Jay Cc: Alley, David; Lingam, Siva ; Pascarelli, Robert

Subject:

Wolf Creek: Verbal Authorization script for Relief Request 14R-04 Alternate CRDM nozzle examinations Attached is the draft verbal authorization script for Relief Request 14R-04 regarding alternate examinations of CROM nozzles at Wolf Creek. I may change the script slightly after the licensee submits the final version of the relief request. Please review and provide changes. The attached verbal authorization does not include Relief Request 14R-03 which Jay is working on. Jay, if you want I can prepare the script for your relief request, 14R-03.

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELI EF REQUEST 14R-03 ALTERNATIVE TO USE VOLUMETRIC LEAK PATH FOR SUPPLEMENTAL EXAMS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated October 14, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-03 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration weld numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 at the Wolf Creek Generating Station. The licensee proposed (a) to perform a volumetric leak path assessment of each penetration nozzle in lieu of the surface leak path assessment required by Paragraph - 3200(b) of ASME Code Case N-729-1, and (b) if an unacceptable indication by the leak path assessment or volumetric exam is identified , the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The NRC staff finds that while the demonstrated volumetric leak path is not equivalent to a fully qualified surface leak path assessment, the licensee identified sufficient operational experience, technical basis and radiological dose hardship to show that regulatory compliance would result in hardship without a compensating increase in the level of quality and safety. For operating experience, the licensee showed that there has been no previous identified cracking or leakage identified from the CRDM nozzle penetrations or welds of the upper head at Wolf Creek. The NRC staff noted that while this fact does not preclude the possibility of cracking to be found as the plant continues to age, plants which have previously identified cracking are more likely to see subsequent and more significant cracking in the future. Given the lack of the initial cracking being identified in the nozzle heats of mate rial, at the operating temperatures of Wolf Creek, the NRC found that the potential for significant cracking this outage was less likely. For technical basis, the licensee identified that their inspection would be in compliance with the Wesdye Technical Justification Document showing an effective demonstration of the volumetric leak path technique. The NRC has accepted the use of a demonstrated volumetric leak path as part of the upper head inspection program under 10 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced NUREG/CR-7142, Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation, which found, in part, the use of a properly focused 0 degree probe could detect a leakage path under low leakage rates during operation that led to minimal wastage of the upper head low alloy steel. While the NRC staff did not find that the volumetric leak path assessment was equivalent to a qualified surface leak path assessment, the information does demonstrate the effectiveness of the volumetric leak path examination to detect low leakage rates, as performed in accordance with the licensee's proposed alternative.

For hardship, the licensee noted that a qualified surface leak path assessment could be performed in two manners that would require both additional radiological dose and time versus the performance of a volumetric leak path assessment. The licensee estimated 3.4 Rem and 1O days for an eddy current surface examination and 18 REM to perform a liquid dye penetrant examination of all of the 12 penetration welds. The NRC staff found both of these conditions to be of sufficient hardship given the operational experience and technical adequacy of the licensee's proposed alternative versus the regulatory requirement. Therefore, the NRC staff finds that the licensee's proposed alternative provides reasonable assurance of structural integrity until the next scheduled examination, and that compliance with the surface examination requirements of Paragraph -3200(b) of ASME Code Case N-729-1, for the subject welds, would result in hardship without a compensating increase in the level of quality and safety. Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation, I concur with the Component Performance, Non-Destructive Examination, and Testing Branch's determinations. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CROM penetration nozzles numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 such that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(ii)(D). Therefore, the NRG staff authorizes the use of relief request 14R-03 at the Wolf Creek Generating Station during the current refueling outage subject to the licensee's proposed alternative that if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee will revert to the requirements of Code Case N-729-1and10 CFR 50.55a(g)(6)(ii)(D). All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector. This verbal authorization does not preclude the NRG staff from asking additional clarification questions regarding Relief Request 14R-03, while preparing the subsequent written safety evaluation.

From: Collins, Jay Sent: 17 Oct 2016 07:54:18 -0400 To: Lingam, Siva Cc: Singal, Balwant

Subject:

RE: Wolf Creek Relief Requests Attachments: wolfcreekHead2016.pdf Greetings, Slides for the internal meeting. Jay Collins

From: Collins, Jay Sent: 17 Oct 2016 19:00:26 +0000 To: Barillas, Martha

Subject:

List of Plants with Cold Leg Temperature Upper Heads with indications of PWSCC in penetration nozzles and/or welds Attachments: Cold Head Cracking.jpg

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From: Lingam, Siva Sent: 17 Oct 201612:34:35 +0000 To: Cumblidge, Stephen

Subject:

FW: Wolf Creek Revised RRs - Internal Discussion Attachments: pr222_.pdf, wolfcreekHead2016.pdf

---

Original Appointmcnt-----

From: Lingam, Siva Sent: Monday, October 17, 2016 8:20 AM To: Lingam, Siva; Dodson, Douglas; Thomas, Fabian; Proulx, David; Drake, James; Werner, Greg; Anchondo, Isaac; Kopriva, Ron; Collins, Jay; Tsao, John; Alley, David; Pascarelli, Robert

Subject:

Wolf Creek Revised RRs - Internal Discussion When: Monday, October 17, 2016 9:00 AM- 10:00 AM (UTC-05:00) Eastern Time (US & Canada). Where: HQ-OWFN-1OB06- I2p Please note the following to discuss the subject RRs based on the attached revised RRs and the slides. The licensee wants NRC approval today to support the outage. Bridge No.: 877-935-1422 Passcode: mJ followed by # Date: October 17, 2016 (Monday) Time: 9:00 AM (Eastern Time)

From: lingam, Siva Sent: 17 Oct 2016 19:14:00 +0000 To: Karwoski, Kenneth Cc: Pascarelli, Robert; Peralta, Juan;Singal, Balwant

Subject:

RE: Wolf Creek SG Follow-up FYI/review. Thank you. From : Singal, Balwant Sent: Monday, October 17, 2016 3:00 PM To: lingam, Siva <Siva.Lingam@nrc.gov>

Subject:

Fwd: Wolf Creek SG Follow-up Please forward it ken karwoski. F rom: "Knust Jason B" < jaknust@WCNOC.com>

Subject:

[External'-Sender] Wolf Creek SG Follow-up Date: 16 October 2016 15:28 To: "Singal, Balwant" < Balwant. Singal@nrc.gov> Cc: "Wagner Pat G" <pawagne@WCNOC.com>, "Muilenburg William T"

<w imuile@WCNOC.com>

Balwant, We wanted to follow up with you regarding the 2 circumferential crack indications we notified you about last week. In addition, we have been working closely with the NRC Region IV ISi inspector, Ron Kopriva.

One of the indications was in SG D (Row 19, Column 83). This indication was characterized as PWSCC, and confirmed with the rotating Ghent probe. It was located within a geometric anomaly, 0.77" below the top of the hot leg tubesheet and was sized at 34% TW max depth and 32 degrees circumferential extent. Per the guidance in the EPRI Steam Generator In Situ Pressure Test Guidelines, no further testing is required. The other indication was in SG C (Row 41, Column 70). This indication was characterized as ODSCC, and was confirmed with the rotating Ghent probe. The indication is located within the top of the hot leg tubesheet expansion transition region, 0.23" below the top of the hot leg tubesheet. The maximum depth of this flaw was sized at 86% TW with a circumferential extent of 112 degrees. We performed in situ pressure testing on this indication today. No leakage was detected up to Steam Line Break Pressure and Burst Testing is not required. Condition Monitoring is met for this indication. Eddy current inspection following the in situ test determined no change to the flaw characteristics. Ron Kopriva witnessed the in situ test. We will stabilize and plug both of these tubes. As previously communicated, we have expanded our hot leg top of tubesheet program to perform 100% inspections in steam generators B/C/D. With the 100% inspection originally planned for SG A during

RF21, we have now performed 100% of the hot leg TIS in all 4 SGs. No additional crack-like indications have been identified. Thank you, Jason Knust Licensing Engineer Wolf Creek Nuclear Operating Corporation (620) 364-8831 x4424

From: lingam, Siva Sent: 17 Oct 2016 07:59:03 -0400 To: Drake, James;Werner, Greg;Anchondo, Isaac; Kopriva, Ron;Dodson, Douglas;Thomas, Fabian;Proulx, David Cc: Pascarelli, Robert;Alley, David;Collins, Jay;Tsao, John;Singal, Balwant

Subject:

FW: Wolf Creek Relief Requests Attachme nts: pr222_.PDF, wolfcreekHead2016.pdf See listing of Records Already Ava ilable to the Pub lic for these attachments. Attached please find the slides for the internal meeting to be setup in addition to the revised RRs from the licensee. From: lingam, Siva Se nt: Monday, October 17, 2016 7:50 AM To: Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <David.Proulx@nrc.gov> Cc: Taylor, Nick <Nick.Taylor@nrc.gov>; Burkhardt, Janet <Janet.Burkhardt@nrc.gov>

Subject:

FW: Wolf Creek Relief Requests We will have an internal call based on the attached revised RRs submittal from the licensee and the pictures placed in the Certrec. I will setup the call and the details will follow. From: lingam, Siva Sent: Friday, October 14, 2016 9:00 PM To: Tsao, John <John.Tsao@nrc.gov>; Collins, Jay <Jay.Collins@ nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov> Subje ct: FW: Wolf Creek Relief Requests Attached please find the revised RRs from the licensee for your review/comment s/evaluation. Thank you. From: Muilenburg William T [6] Se nt: Friday, October 14, 2016 6:42 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>; 'nick.taylolr@nrc.gov' <nick.taylolr@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian .Thomas@nrc.gov>; Li ngam, Siva

<Siva.Li nga m@nre.gov>

Subject:

[External_Send er) Wolf Creek Relief Requests

Everyone, Attached are Wolf Creek Relief Requests 14R-03 and 04.

Bill Muilenburg, 620-364-4186

From: Singal, Balwant Sent: 17 Oct 201615:00:25 -0400 To: Lingam, Siva

Subject:

Fwd : Wolf Creek SG Follow-up Please forward it ken kewasji. From: "Knust Jason B" <jaknust@WCNOC.com>

Subject:

[External_Sender] Wolf Creek SG Follow-up Date: 16 October 201 6 15:28 To: "Singal, Balwant" <Balwant.Singal@nrc.gov> Cc: "Wagner Pat G" <pawagne@WCNOC.com>, "Muilenburg William T"

<wimuile@WCNOC.com>

Bal want, We wanted to follow up with you regarding the 2 circumferential crack indications we notified you about last week. In addition, we have been working closely with the NRC Region IV ISi inspector, Ron Kopriva . One of the indications was in SG D (Row 19, Column 83). This indication was characterized as PWSCC, and confirmed with the rotating Ghent probe. It w as located within a geometric anomaly, 0.77" below the top of the hot leg tubesheet and was sized at 34% TW max depth and 32 degrees ci rcumferential extent. Per the guidance in the EPRI Steam Generator In Situ Pressure Test Guidelines, no further testing is required. The other indication was in SG C (Row 41, Column 70). Th is indication was charact erized as ODSCC, and was confirmed with the rotating Ghent prob*e. The indication is located within the top of the hot leg tubesheet expansion transition region, 0.23" below the top of the hot leg tubesheet. The maximum depth of this flaw was sized at 86% TW with a circumferential extent of 112 degrees. We performed in situ pressure testing on this indication today. No leakage was detected up to Steam Line Break Pressure and Burst Testing is not required. Condition Monitoring is met for this indication. Eddy current inspection following the in situ test determined no change to the flaw characteristics. Ron Kopriva witnessed the in situ test. We will stabilize and plug both of these tubes. As previously communicated, we have expanded our hot leg top of tubesheet program to perform 100% inspections in st eam generators B/C/D. W ith the 100% inspection originally planned for SG A during RF21, we have now performed 100% of the hot leg TIS in all 4 SGs. No additional crack-like indications have been identified. Thank you, Jason Knust Licensing Engineer Wolf Creek Nuclear Operating Corporation

(620) 364-8831 x4424 From: Taylor, Nick Sent: 17 Oct 2016 15:30:32 -0500 To: Lingam, Siva; Dodson, Douglas;Thomas, Fabian; Proulx, David Cc: Burkhardt, Janet

Subject:

RE: Wolf Creek Relief Requests

Siva, Did this internal call occur today? I have a few questions I observations about the relief request, will hold them for the internal call whenever that occurs.

Thanks, Nick Taylor 972-921-6398 From: Lingam, Siva Se nt: Monday, October 17, 2016 6:50 AM To: Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>

Cc: Taylor, Nick <Nick.Taylor@nrc.gov>; Burkhardt, Janet <Janet.Burkhardt@nrc.gov> Subje ct: FW : Wolf Creek Relief Requests We will have an internal call based on the attached revised RRs submittal from the licensee and the pictures placed in the Certrec. I will setup the call and the details will follow. From: Lingam, Siva Se nt: Friday, October 14, 2016 9:00 PM To: Tsao, John <John .Tsao@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

FW: Wolf Creek Relief Requests Attached please f ind the revised RRs from the licensee for your review/comments/evaluation. Thank you. From: Muilenburg Will iam T [mailto:wimuile@WCNOC.com) Se nt: Friday, October 14, 2016 6:42 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>; 'nick.taylolr@nrc.gov' <nick.taylolr@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Li ngam, Siva

<Siva. Li nga m@nre.gov>

Subject:

[External_Send er] Wolf Creek Relief Requests

Everyone, Attached are Wolf Creek Relief Requests 14R-03 and 04.

Bill Mu ilenburg, 620-364-4186

From: Tsao, John Sent: 17 Oct 2016 12:57:13 -0400 To: Collins, Jay;Alley, David Cc: Lingam, Siva;Pascarelli, Robert;Singal, Balwant

Subject:

RE: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment Attachme nts: Wolf Creek verbal auth 14R-04 10-17-2016.docx Per Dave's suggestion, I have revised the verbal authorization script for Relief Request 14R-04 by adding the following sentence to the second paragraph in the attached script. "The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety ... " From: Collins, Jay Sent: Monday, October 17, 2016 12:50 PM To: Alley, David <David.Alley@nrc.gov> Cc: Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>

Subject:

RE: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment Well, since it is a verbal and there are no actual requirements for what we put in the script. ... But to provide you "the quote", so you can say it one more time. as you go through my long script, I have include the attached a new version of the script. (Just the one line is added, you guys don't have to look at it again, thanks for the feedback btw) From: Alley, David Se nt: M onday, October 17, 2016 11:58 AM To: Collins, Jay <Jay.Collins@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Singal, Balwant

<Balwant .Singal@nrc.gov>

Cc: Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov>

Subject:

RE : Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumet ric Lea k Path Assessment John, Jay, Both verbals appear well written . I do have one question . In a normal SE in the first or second paragraph we say that the licensee is proposing its alternative in accordance with 50.55a(z)(1) (or in this case (z)(2)) or words to that effect. In the scripts I need to get all the way to the end to officially find out that these are (z)(2) requests. Do we normally put the authority for the proposed alternative up front? Dave

From : Collins, Jay Sent: Monday, October 17, 2016 11:24 AM To: Tsao, John <John .Tsao@nrc.gov>; Singal, Balwant <Balwant.Singa l@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert

<Robert.Pascarelli@nrc.gov>

Subject:

RE: Wolf Creek: Verbal Authorization script for Relief Request 14R-03 Volumetric Leak Path Assessment Greetings, 1st Draft of the script for 14R-03. Please provide me comments when you can. Jay From : Tsao, John Se nt: Thursday, October 13, 2016 2:09 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>; Pascarelli, Robert

<Robert.Pascarelli@nrc.gov>

Subject:

Wolf Creek: Verbal Authorization scri pt for Relief Request 14R-04 Alternate CRDM nozzle examinations Attached is the draft verbal authorization script for Relief Request 14R-04 regarding alternate examinations of CROM nozzles at Wolf Creek. I may change the script slightly after the licensee submits the final version of the relief request. Please review and provide changes. The attached verbal authorization does not include Relief Request 14R-03 which Jay is working on. Jay, if you want I can prepare the script for your relief request, 14R-03.

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELI EF REQUEST 14R-04 ALTERNATE EXAMINATION OF CONTROL ROD DRIVE MECHAN ISM NOZZLE PENETRATIO NS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated October 11, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-04 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration numbers 77 and 78 at the Wolf Creek Generating Station. The licensee proposed (a) an alternate examination distance for CROM nozzle numbers 77 and 78 in lieu of the required examination distance per ASME Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) and (b) not to perform the surface examination of the portion of the CROM nozzle below the J-groove weld as required by 10 CFR 50.55a(g)(6)(ii)(D)(3). The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that the proposed examination distance are acceptable for CROM nozzle numbers 77 and 78. This is based on the validity of the licensee's stress analysis and fracture mechanics calculation, demonstrating that within four refueling cycles, a potential flaw that initiates in the unexamined zone (below the J-groove weld) of the CROM nozzle numbers 77 and 78 will not propagate into the J-groove weld. At the end of every fourth refueling cycle, the licensee will perform an examination to confirm the structural integrity of CROM nozzles 77 and 78. The NRC staff finds the licensee's hardship justifiication is acceptable because of the considerable radiation dose and the nozzle configuration that are not conducive for the required examination. The NRC staff finds that the licensee's proposed alternative examination distances for CROM penetration nozzle numbers 77 and 78 provides reasonable assurance of structural integrity and leak tightness until the next scheduled examination, and that compliance with the surface examination requirements of 10 CFR 50.55a(g)(6)(ii)(D)(3) would result in hardship without a compensating increase in the level of quality and safety. Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation

As Chief of the Plant Licensing Branch IV-1, Office of Nuclear Reactor Regulation , I concur with the Component Performance, Non-Destructive Examination, and Testing Branch's determinations. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CROM penetration nozzles numbers 77 and 78 and that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(ii)(D). Therefore, the NRC staff authorizes the use of relief request 14R-04 at the Wolf Creek Generating Station for the remainder of the fourth 10-year ISi interval, which ends on September 2, 2025. All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third p*arty review by the Authorized Nuclear In-service Inspector. This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding Relief Request 14R-04, while preparing the subsequent written safety evaluation.

From: Tsao, John Sent: 17 Oct 2016 12:29:30 +0000 To: Lingam, Siva

Subject:

Accepted: Wolf Creek Revised RRs - Internal Discussion

From: Alley, David Sent: 19 Oct 2016 01:11:44 +0000 To: Rezai, Ali

Subject:

RE: RAI for WolfCreek-1 13R-13 MF8308 Coverage Attachments: WolfCreek-1 13R-13 RAI MF8308 Coverage PP DA.docx, WolfCreek-113R-13 SE MF8308 Coverage Draft with HOLES DA.docx Ali A few thoughts on both the SE and the RAls Dave From: Rezai, Ali Sent: M onday, October 17, 2016 10:48 AM To: Alley, David <David.Alley@nrc.gov>

Subject:

RAI for WolfCreek-113R-13 MF8308 Coverage Hi Dave, Attached is RA I for your review/concurrence. Due RAI: 11 -27 Have a great day.

Regards, ali

REQUEST FOR ADDITIONAL INFORMATION Comment IADI: I normally view MRP-RELIEF REQUEST 13R-13 REGARDING WELD EXAMINATION COVERAGE  : 146 as an augmented inspection WOLF CREEK NUCLEAR OPERATING CORPORATION f program. In that case it appears that WOLF CREEK GENERATING STATION  ! the RAJ is unnecessary. Given the DOCKET NUMBER 50-482  ! potential that the plant may not

                                                                                                                               / consider MRP-146 to be an i augmented inspection program, I

suggest that you get with the PM when By letter dated August 23, 2016 (Accession Number ML16243A039), Wolf Creek Nuclear Operating Corporation (the licensee) requested relief from the requirements of the American f he gets back (vacation right now I

                                                                                                                             ! think) and ask the PM to confirm that Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) specifically                                     i the plant considers MRP-146 to be an related to ASME Code Case N-460 "Alternative Examination Coverage for Class 1 and Class 2                                   /       augmented inspection program, I.e.,

Welds, Section XI, Division 1." Relief request 13R-13 pertains to the examination coverage of f these welds are not MRP 146 welds. the Class 1 and 2 piping welds at the Wolf Creek Generating Station (Wolf Creek).  ! Given that the request clearly states f that they are not in augmented To complete its review, the NRC staff requests the fol lowing additional information. f programs and MRP-146 should be

considered an augmented program, I

1. lrt§.1}9_~g_Q _~_9_t~!t~9-~~~~!_~_!g_!!1~-~l}!L~!!~9.~~-~!._ !~-~_IJ9_~~-~l}~-~!~!~_g_!~~JQIJg_~J!l.9" ________j ~~~*~:~~ko~~i~~~~:~~~~P.r~~~~t should be obvious.

         "Often, a pipe-to-valve weld is subjected to the highest temperature difference, and is more susceptible to thermal fatigue than an element further down the piping line."
         "The subject welds are not part of an augmented inspection program. "

The NRC staff notes that the welds in this relief request are categorized as "elements subject to thermal fatigue." As part of an augmented inspection program for managing thermal fatigue, industry has issued Materials Reliability Program (MRP)-146 "Management of Thermal Fatigue in Normally Stagnant Non-Jsolable Reactor Coolant System Branch Lines" and/or the Electric Power Research Institute (EPRI) interim guidance MRP 2015-025 "EPRl-MRP Interim Guidance for Management of Thermal Fatigue" (Accession Number ML15189A100).

a. Discuss why the welds in this relief request are not part of the augmented inspection program in MRP-146 and/or EPRI interim guidance MRP 2015-025.

b. Given the susceptibility to thermal fatigue and the reduced coverage obtained, and for assurance of structural integrity of unexamined volume of the weld, provide cumulative fatigue usage (CFU) factor for each weld.

2. The NRC staff notes that the refracted longitudinal (L) waves have shown to have better penetration capability in the cast austenilic stainless steel and austenitic stainless steel materials, and they could be used as an extra effort to scan the far-side of examination volume ("Best Effort" examination). The NRC staff also notes that the "Best Effort" examination is not a requirement. Given the reduced inspection coverage of the weld under consideration:

a. Discuss whether the license performed the "Best Effort" examination as an extra effort to interrogate the required downstream examination volume (far-side),

particularly the root of the weld and the heat affected zone (HAZ) of the base materials typically susceptible to high stresses and potential degradation, If not, explain;

b. Provide percentage of coverage obtained from the "Best Effort" examination if this examination was performed.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 13R-13 REGARDING WELD EXAMINATION COVERAGE WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NUMBER 50-482

1.0 INTRODUCTION

By letter dated August 23, 2016 (Accession Number ML16243A039), as supplemented by letter dated xxxxxxxxxx, 2016 (Accession Number MLxxxxxx), Wolf Creek Nuclear Operating Corporation (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) specifically related to ASME Code Case N-460 "Alternative Examination Coverage for Class 1 and Class 2 Welds, Section XI , Division 1." This relief request, 13R-13, pertains to the examination coverage of the Class 1 and 2 piping welds at the Wolf Creek Generating Station (Wolf Creek). Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 O CFR) 50.55a(g)(6)(i), the licensee requested relief from the required examination coverage and to use alternative requirements (if necessary), for inservice inspection (ISi) of the piping welds on the basis that the ASME Code requirement is impractical. 2.0 REGULATORY REQUIREMENTS Pursuant to 10 CFR 50.55a(g)(4), lnservice Inspection Standards Requirement for Operating Plants, throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME Code that become effective subsequent to editions specified in paragraphs (g)(2) and (3) of 50.55a and that are incorporated by reference in paragraph (a)(1 )(ii) of 50.55a, to the extent practical within the limitations of design, geometry, and materials of construction of the components. Pursuant to 10 CFR 50.55a(g)(4)(ii), Applicable ISi Code: Successive 120-month Intervals, inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of 50.55a 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using Section XI, that are incorporated by reference in paragraphs (a)(3)(ii) of 50.55a), subject to the conditions listed in paragraph (b) of 50.55a. However, a licensee whose inservice inspection interval commences during the 12 through 18-month period after July 21, 2011, may delay the update of their Appendix VIII program by up to 18 months after July 21 , 2011. Pursuant to 10 CFR 50.55a(g)(5)(iii), IS/ Program Update: Notification of Impractical /SI Code Requirements, if the licensee has determined that conformance with the ASME Code requirement is impractical for its facility, the licensee must notify the NRC and submit, as specified in § 50.4, information to support the determinations. Determinations of impracticality in accordance with 50.55a must be based on the demonstrated limitations experienced when attempting to comply with the Code requirements during the inservice inspection interval for which the request is being submitted. Requests for relief made in accordance with 50.55a must

be submitted to the NRC no later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought. Pursuant to 10 CFR 50.55a(g)(6)(i), Impractical ISi Requirements: Granting of Relief, the Commission will evaluate determinations under paragraph (g)(5) of 10 CFR 50.55a that ASME Code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines are authorized by law, and will not endanger life or property or the common defense and security, and are otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Background By letter dated February 21, 2007 (Accession Number ML070260538), the NRC approved implementation of the risk informed inservice inspection (RI-ISi) program for the Class 1 piping welds (Examination Category B-F and B-J) and the Class 2 piping welds (Examination Category C-F-1 and C-F-2) in the third 10-year ISi interval of Wolf Creek. The licensee developed the RI-ISi program in accordance with the NRC approved methodology of the Electric Power Research Institute (EPRI) Topical Report (TR)-112657, Revision B-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure" (Accession Number ML013470102). 3.2 Component Affected The affected components are ASME Code Class 1 and 2 piping welds (as identified in Tables 1 and 2 of Attachment 1 to this relief request). The licensee stated that,

• The four Class 1 welds consist of three pipe to valve welds and one fitting to flange weld in the reactor coolant system.
• The three Class 2 welds consist of pipe to valve welds in the residual heat removal system.

The licensee classified the above welds as Examination Category R-A, Item Number R 1.11 (elements subject to thermal fatigue) in accordance with EPRI TR-112657, Revision B-A, (Table 1 of ASME Code Case N-578-1 ). The licensee provided operating pressure and temperature, nominal pipe size (NPS), wall thickness, and materials of construction for each of the above welds. The licensee stated that,

• The pipes and the welds are made of austenitic stainless steel.
• The fitting and valve bodies are made of either forged or cast austenitic stainless steel.

3.3 Applicable Code Edition and Addenda

The code of record for the third 10-year ISi interval is the 1998 Edition through 2000 Addenda of the ASME Code. 3.4 Duration of Relief Request The licensee submitted this relief request for the third 10-year ISi interval which started on September 3, 2005, and ended on September 2, 2015. 3.5 ASME Code Requirement The ASME Code requirements applicable to this request originate in ASME Code, Section XI, Table IWB-2500-1. Alternative to these requirements is the RI-ISi program for Wolf Creek, that was developed by the licensee in accordance with the NRC approved methodology in EPRI TR-112657, Revision B-A (Accession Number ML013470102), and that was authorized by the NRC staff in a safety evaluation dated February 21, 2007 (Accession Number ML070260538). In both the ASME Code requirements and the NRC safety evaluation, the welds under this request are required to be volumetrically examined during each 10-year ISi interval, and 100 percent coverage of the required examination volume must be achieved. The extent of required examination coverage is reduced to essentially 100 percent by ASME Code Case N-460. This code case has been incorporated by reference into 10 CFR 50.55a by inclusion in Regulatory Guide 1.147, Revision 17. 3.6 Impracticality of Compliance The licensee stated that it was not possible to obtain greater than 90 percent of the ASME Code required examination volume due to limitations, which include configuration and geometry of the welds and/or the associated components and metallurgical constraints. In Table 3 and the diagrams in Attachment 1 to the relief request, the licensee described and illustrated the limitations that prevented ultrasonic scanning of the weld. Examples include a valve body that limits access to valve side of the weld, and a fitting that limits access to flange side of the weld, and that restricts the ultrasonic scanning. The licensee stated that the burden caused by compliance includes major modification of plant components which include redesign and replacement of the welds and associated components.

3. 7 Bases for Relief The licensee stated that it performed the ultrasonic testing (UT) to the maximum extent possible utilizing personnel qualified and procedures demonstrated in accordance with Appendix VIII of Section XI.

HOLE In the xxxxxxxxx letter, the licensee provided the percentage of coverage for the "Best Effort" examinations. The licensee stated that for the welds in this relief request with single sided access, it extended the beam path into the far side of the weld centerline to examine to the extent practical the other side of weld as a "Best Effort" examination. However, no credit was claimed for the "Best Effort" examination because a UT procedure must be qualified with flaws on the inaccessible side of the weld. Currently, there are no qualified single-side examination procedures and the existing UT technology is not capable of reliably detecting o sizing flaws on the far side of an austenitic weld. No unacceptable indications were identified.

1 Comment (AD(: This paragraph is (rhe licensee stated that there were 15 other austenitic stainless steel pipe welds with the / potentially the most important part of degradation mechanism of thermal fatigue that were examined during the third 10-year ISi / their submission. Please include this inte rval. The operating conditions and environment would be similar to the welds in this relief / conceo1 in the NRC evaluation. request. No indication of degradation due to thermal fatigue, or any other mechanism, has been / identified during the exa minations l_ __________________________________ -------________________________________ _/ HOLE The licensee stated that welds included in this request are not part of any augmented inspection programs including Materials Reliability Program (MRP)-146 "Management of [Thermal Fatigue in Normally Stagnant Non-lsolable Reactor Coolant System Branch Lines," a nd/or the Electric Power Research Institute (EPRI) interim guidance MRP 2015-025 "EPRl-MRP Interim Guidance for Management of Thermal Fatigue" (Accession Number ML15189A100). HOLE In the xxxx, letter, the licensee provided the cumulative fatigue usage (CFU) factQ!J calculated for each welded location in this relief request. The licensee stated that the welds in this relief request have been subjected to system leakage testing and no sign of leakage has been identified. 3.8 Proposed Alternative In Table 3 of Attachment 1 to this relief request, the licensee reported the percent coverage achieved for each weld examined. This is summarized below. Class 1 welds: BB-02-F019 50 percent BB-02-FW301 50 percent BG-21-F013B 50 percent EJ-04-F048A 50 percent Class 2 welds: EP-01-MW7152 50 percent EP-02-MW7162 50 percent EP-01-MW7165 50 percent The licensee proposed the above alternative coverage in lieu of the ASME Code required essentially 100 percent coverage. 3.9 NRC Staff Evaluation The NRC staff has evaluated relief request 13R-13 pursuant to 10 CFR 50.55a(g)(6)(i). The NRC staff's evaluation focused on: (1) whether a technical justification exists to support the determination that the ASME Code requirement is impractical; (2) that imposition of the Code required inspections would result in a burden to the licensee; and (3) that the licensee's proposed alternative (accepting the reduced inspection coverage in this case) provides reasonable assurance of structural integrity and leak tightness of the subject weld. The NRC

staff finds that if these three criteria are met that the requirements of 10 CFR 50.55a(g)(6)(i), (i.e., granting the requested relief will not "endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility") will also be met. Impracticality of Compliance* As described and demonstrated in the submittal, Table 3, and the sketches in Attachment 1 to 13R-13, the predominant limitations that prevented the licensee's UT to achieve essentially 100 percent coverage of the ASME Code required volume were the pipe to valve and the fitting to flange configurations and/or the metallurgical constraints. The licensee performed the UT from one side of the welds because scanning from the other side of the welds was not possible (single-sided scan). The NRC staff confirms that each weld's particular design configuration prevented the licensee to scan the welds from both sides. Therefore, the NRC staff finds that a technical justification exists to support the determination that achieving essentially 100 percent coverage is impractical. Burden of compliance The licensee proposed that making the welds accessible for inspection from both sides would require replacement or significant design modification to the welds and their associated components. The NRC staff finds that replacing or reconfiguring the components of the subject welds is the only reasonable means to achieve dual sided coverage of these welds and that replacement or reconfiguration of the pipe, valve, fitting, and flange constitutes a burden on the licensee. Structural integrity and leak tightness The NRC staff considered whether the licensee's proposed alternative provided reasonable assurance of structural integrity and leak tightness of the subject weld based on: (1) the examination coverage achieved and (2) safety significance of unexamined volumes - unachievable coverage (e.g., the presence or absence of known active degradation mechanisms and essentially 100 coverage achieved for similar welds in similar environments subject to similar degradation mechanisms). Examination Coverage Achieved In evaluating the licensee's proposed alternative, the NRC staff assessed whether it appeared that the licensee obtained as much coverage as reasonably possible and the manner in which the licensee reported the coverage achieved. From review of submittal and the sketches in to 13R-13, the NRC staff confirms that:

• The welds were examined using the appropriate equipment, ultrasonic modes of propagation, probe angles, frequencies, and scanning directions to obtain maximum coverage;
• The coverage was calculated in a reasonable manner;
• The UT procedures used were qualified as required by the regulation ;
• The coverage was limited by physical access (i.e., the configuration of one side of the weld did not permit access for scanning);
• No unacceptable indications were identified .

Therefore, the NRG staff found that the licensee made every efforts to obtain as much coverage as reasonably possible with the ASME Code required UT. Safety Significance of Unexamined Volumes - Unachievable Coverage In addition to the coverage analysis described above the NRG staff evaluated the safety significance of the unexamined volumes of welds - unachievable coverage. From review of submittal and the sketches in Attachment 1 to 13R-13, the NRG staff verified that:

• The licensee's UT has covered, to the extent possible, the regions (i.e., the weld root and the heat affected zone (HAZ) of the base material near the ID surface of the joint) that are typically susceptible to higher stresses and, therefore, potential degradation.
• HOLE For the stainless steel welds, the NRG staff notes that the coverage obtained was limited to the volume up to the weld centerline (near-side), because claiming coverage for the volume on the opposite side of the weld centerline (far-side) requires meeting the 10 CFR 50.55a(b)(2)(xv)(A)(2) far-side UT qualifications, which has not been demonstrated in any qualification attempts to date. The far-side volume was inspected by the "Best Effort" examination, no indications were identified, and no credit was taken for the coverage achieved from the "Best Effort" examination.
• HOLE At each of the welded location in this relief request, the licensee's calculated cumulative fatigue usage (CFU) factor does not exceed the limit of Section Ill of the SME Code. The CFU factor was determined based on the actual plant operating cycles. Therefore , this provide reasonable assurance that the potential for initiation and growth of fatigue cracks is low and of least concern at these welds and their associated HAZ of base materials.

Therefore, the NRG staff determined that based on the coverage achieved by the qualified UT, the supplemental "Best Effort" examinations, the examination of the weld root and its HAZ to the extent possible, and bounding CFU, it is reasonable to conclude that if significant service induced degradation had occurred, evidence of it would have been detected by the examinations that the licensee performed. In this analysis, the NRG staff also found that, in addition to the required volumetric examinations, these welds have received the required system leakage test according to the ASME Code, Section XI, IWB-2500 (Table IWB-2500-1, Examination Category B-P) during each refueling outage and IWC-2500 (IWC-2500-1, Examination Category C-H) each inspection period. Despite reduced coverage of the required examination volume, the NRC staff finds that this inspection will provide additional assurance that any pattern of degradation, if it were to occur, would be detected and the licensee will take appropriate correction actions. Therefore, the NRG staff finds that the volumetric examinations performed to the extent possible provide a reasonable assurance of structural integrity and leak tightness of the subject welds. Compliance with the ASME Code requirements for these welds would be a burden on the licensee.

4.0 CONCLUSION

As set forth above, the NRC staff determines that it is impractical for the licensee to comply with the ASME Code, Section XI requirement; that the proposed inspection provides reasonable assurance of structural integrity or leak tightness of the subject welds; and that granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i). Therefore, the NRC staff grants this relief request at Wolf Creek for the third 10-year ISi interval which commenced on September 3, 2005, and ended on September 2, 2015. All other ASME Code, Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In service Inspector.

From: Collins, Jay Sent: 18 Oct 2016 07:17:35 -0400 To: Taylor, Nick;Lingam, Siva Cc: Tsao, John;Alley, David;Cumblidge, Stephen

Subject:

RE: Call with Wolf Creek regarding head inspection Greetings, I am doing the 14R-03 relief and John Tsao is doing the 14R-04 relief. If you would like to have a call on the relief requests, we should be available after our branch meeting this morning ends at 10am our time, 9am Central. I am getting an automatic reply for you, so if you would like to do them by email, we could do that, as well. Stephen Cumblidge is making up a nice presentation about the volumetric leak path assessment, if you have questions on that item.

Thanks, Jay Collins NRR/DE/EPNB (301 )415-4038 Siva, we will be in 0-886 for our branch meeting from 9 to 10am.

From : Taylor, Nick Se nt: M onday, October 17, 2016 5:55 PM To: Drake, James <James.Drake@nrc.gov>; Pascarelli, Robert <Robert .Pascarelli@nrc.gov>; Alley, David

<David.Alley@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>; Pruett, Troy <Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>

Cc: Tsao, John <John.Tsao@nrc.gov>; Collins, Jay <Jay.Collins@ nrc.gov>; Dodson, Douglas

<Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Anchondo, Isaac
<lsaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Werner, Greg
<Greg.Werner@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>

Subject:

RE : Call with Wolf Creek regarding head inspection Hello Jim. My understanding was that there was going to be an NRC-only call to discuss the relief request. I have a number of questions based on my review of the relief request this morning. Has that call already occurred? I left a message with Siva Lingam (who is standing in for Balwant) to that affect as welt. ..

Thanks, Nick Taylor Chief, Projects Branch B 972-921-6398 From : Drake, James Se nt: Monday, October 17, 2016 4:51 PM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Vegel, Anton

<Anton.Vegel@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>; Pruett, Troy <Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>

Cc: Tsao, John <John.Tsao@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov>; Taylor, Nick

<Nick.Taylor@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian

<Fabian.Thomas@nrc.gov>; A11chondo, Isaac <lsaac.Anchondo@nrc.gov>; Kopriva, Ron
<Ron.Kopriva@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>

Subject:

Call with Wolf Creek regarding head inspection We had a conference call with the licensee and discussed our concern with other potentially relevant indications seen in the pictures provided that were not discussed in the relief request. The licensee stated that they have dispositioned all of the relevant indications on the vessel head and intend to address each of them. The pictures provided were not necessarily post inspection. They are going to draft up a shortened version of the quality control report with how they dispositioned and path forward for any relevant indications they had that are not addressed in the relief request. The specific penetrations were:64,53,75,56,32, 15,6,43,67,66, and 54. I let them know that these numbers were our best determinations from the pictures provided, but may not be completely accurate if we were off on the reference positions in the pictures. The licensee is working on the paper and will call me when they are ready to provide it. If you have any questions, feel free to contact me on my cell phone tonight or office phone tomorrow. Jim ~m1eJ f:. :Dra{e James F. Drake Office phone: 817-200-1558 Cell Phone: l<b)(6) I

From: Holston, William Sent: 18 Oct 2016 12:52:51 -0400 To: byka@asme.org;Gary Park;powersl@asme.org Cc: Manoly, Kamal;Benson, Michael;Holston, William;ASME Code Day Attendees;ASME Code Part icipants

Subject:

NRC ASME August 2016 Report Attachme nts: 2016-11 -- NRC Report to ASME - St Louis MO.docx I have attached the November staff report to ASME. Bill Holston

NRG Report to ASME November 2016 NRC Report for ASME Code Meetings November 2016- St Louis, MO Table of Contents 1 Amendment to 10 CFR 50.55a - ASME Code Edition/Addenda 2 2 ASME Code Case Rulemaking/Regulatory Guides 2 3 Operating Plant Issues and Material Degradation 4 4 NRO DCIP Quality Assurance and Vendor Inspection Branch Activities 8 5 New Reactor Licensing Activities 10 6 Multinational Design Evaluation Program (MDEP) Activities 13 7 10 CFR Part 21 Rulemaking 15 8 Commercial Calibration Services Status 15 9 NRC Staff Review of EPRI 1025243 Guideline for Commercial-Grade Design 16 and Analysis Computer Programs 10 NRC Staff Review of EPRI Guideline for Dedication of Commercial-Grade Items 16 for Use in Nuclear Safety-Related Applications 11 NRC Staff Interface with Nuclear Utilities Procurement Issues Committee 17 (NUPIC) 12 Reverse Engineering Information Notice 2016-01 17 13 License Renewal Activities 17 14 New Generic Letters 20 15 New Information Notices 20 16 New Regulatory Issue Summaries 20 17 NRC Publications of Potential Interest to ASME 20 18 Upcoming Public Meetings of Potential Interest to ASME 21 NRG Report to ASME November 2016

1. Amendment to 10 CFR 50.55a - ASME Code Edition/Addenda Current ASME Edition/Addenda The NRG has approved:

• Section Ill, Division 1 and Section XI, Division 1 of the Boiler and Pressure Vessel Code through the 2008 Addenda (76 FR 36232; June 21, 2011 ).
• The Operation and Maintenance of Nuclear Power Plants (OM Code) through the 2006 Addenda (76 FR 36232; June 21 , 2011 ).

Next ASME Edition/Addenda The next proposed amendment to 10 CFR 50.55a includes:

• The 2009 Addenda, the 2010 Edition, 2011 Addenda, and the 2013 Edition of the Boiler and Pressure Vessel Code.
• The 2009 Edition, 2011 Addenda and 2012 Edition of the Operation and Maintenance of Nuclear Power Plants (OM Code).
• Section XI Code Case N-824 will be directly listed in 50.55a as conditionally approved for use.
• Section XI Code Case N-729-4 will be directly listed in 50.55a as required with conditions.
• Section XI Code Case N-770-2 will be directly listed in 50.55a as required with conditions.

NRG Staff has addressed the public comments received and is in the process of preparing the final rule for publication. The final rule is expected to be published by December 2016. In preparation for the next rulemaking action, a review of the 2015 Edition of the ASME BPV Code and OM Code for incorporation by reference in 10CFR50.55a has been completed. The scope of the rulemaking action along with draft conditions are currently under development with a proposed rule publication following shortly after the 2009-2013 rule becomes final. The NRC discussed its preliminary positions on the 2015 Editions during a public meeting held on August 22, 2016.

2. ASME Code Case Rulemaking/Regulatory Guides Current RG Publications On November 5, 2014 a final rule was published in the Federal Register (79 FR 65776) that incorporates by reference the Regulatory Guides (RGs) listed below:

Supplements Addressed: Supplements 1 through 10 to the 2007 Edition Effective date for the RGs: December 5, 2014

• RG 1.84, Revision 36, "Design, Fabrication, and Materials Code Case Acceptability, ASME Section Ill" (ADAMS Accession No. ML13339A515).
• RG 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1" (ADAMS Accession No. ML13339A689).

NRG Report to ASME November 2016

• RG 1.192, Revision 1, "Operation and Maintenance Code Case Acceptability, ASME OM Code" (ADAMS Accession No. ML13340A034).

In addition, on November 5, 2014, a final guide was published in the Federal Register (79 FR 65776):

• RG 1.193, Revision 4, "ASME Code Cases Not Approved for Use" (ADAMS Accession No. ML 13350AOO 1).

Next RG Publications The proposed Code Case Rulemaking package was published in the Federal Register on March 2, 2016 The75 day public comment period closed on May 16, 2016. NRC Staff are currently addressing public comments received on the draft rule. The final rule is expected to be published in the spring of 2017. Scope of the Current ASME Code Case Rulemaking Code Case Supplements Addressed: Supplement 11 to the 2007 Edition through Supplement 10 to the 2010 Edition

• Draft Revision 37 of RG 1.84
• Draft Revision 18 of RG 1.147
• Draft Revision 5 of RG 1.193 Additional Code Cases considered for this rulemaking package at the request of ASME that are not listed in aforementioned Supplements:
• N-694-2 (Supp. 1 to the 2013 Edition) "Evaluation Procedure and Acceptance Criteria for PWR Reactor Vessel Head Penetration Nozzles" Section XI
• N-825 (Supp. 3 to the 2013 Edition) "Alternative Requirements for Examination of Control Rod Drive Housing Welds" Section XI
• N-845 (Supp. 6 to the 2013 Edition) "Qualification Requirements for Bolts and Studs" Section XI OM Code_Code Cases Addressed: 2009 Edition through 2012 Edition
• Draft Revision 2 of RG 1.192 Status of Next RG Publication The NRC staff also initiated the review of the next draft RGs that will address the Code Cases published in Supplement 11 to the 2010 Edition through Supplement 0 of the 2015 Edition of the ASME Code.

Standards Used in RGs and Other Guidance The NRC has placed on its website a series of lists of consensus standards, including those published by ASME, that are referenced in Regulatory Guides, in Inspection Manuals and Procedures, and in the LWR Standard Review Plan. The lists may be found at this website: http://www.nrc.gov/about-nrc/regulatory/standards-dev/consensus.html.OM Code NRG Report to ASME November 2016

3. Operating Plant Issues and Material Degradation Dissimilar Metal Butt Welds CALVERT CLIFFS UNIT 1 (LER No. 3172016002, ADAMS Accession No. ML16106A304)

PRESSURIZER SAFETY RELIEF NOZZLE TO SAFE-END WELD Ultrasonic (UT) examinations performed at Calvert Cliffs Nuclear Power Plant, Unit 1 identified a change from previous examinations in an axial flaw in a pressurizer safety relief nozzle to safe-end weld that was mitigated by the Mechanical Stress Improvement Process (MSIP) in 2006. Evaluation of the data identified one axially oriented flaw contained within the weld material with a depth measured as 81 .6% through-wall including the clad thickness. UT examinations prior to the application of MSIP identified an axial flaw in the same location as the 2016 flaw but a depth of 8% through-wall. UT following MSIP confirmed the flaw was still present at a depth of 8% through-wall. The ISi examinations in 2010 reported essentially no change in the through-wall depth of the indication. Given this information, the NRC staff is evaluating the residual stresses present post MSIPto assess if the current examination regimen for welds mitigated by MSIPneeds to be modified. Reactor Vessel Head Penetrations BRAIDWOOD 1 (EN 52275) - LIQUID PENETRATION EXAMINATION RESULTS IN INDICATIONS ON REACTOR VESSEL HEAD PENETRATION During the Braidwood Station Unit 1 Refueling outage (A 1R19), an in-service Liquid Penetration examination was performed on the previously repaired control rod drive mechanism (CROM) penetration 69. During the examination on the weld build up for CROM penetration 69, two indications were discovered. A 7/32 inch rounded indication was discovered located at 359 degrees on the reactor head portion of the weld buildup, and it is 4 inches from the transition of the head to penetration. A 1/4 inch rounded indication was also discovered located at 200 degrees at the transition of the head to penetration. The transition is the point where the vertical portion of the penetration meets the horizontal area of the reactor head. Rounded indications that exceed 3/16 inch are unacceptable. Peening The NRC staff has reviewed the MRP-335, Revision 3, Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement, which discusses peening as a mitigation technique to permit reduction in inspection frequency for the dissimilar metal butt welds in reactor coolant system piping and reactor vessel head penetration nozzles and associated J-groove welds that are fabricated from nickel-based Alloy 600/82/182 material in PWRs. The Safety Evaluation (SE) was issued to EPRI August 24, 2016 the transmittal letter requested EPRI publish an approved version of MRP-335, Revision 3 within three months. The approved version will incorporate the transmittal letter and the final SE after the title page. The NRC staff determined that, given the input variables proposed in MRP-335R3, the analyses provided did not fully support the inspection intervals proposed in MRP-335R3. Therefore, the NRC staff imposed conditions to ensure that the proposed inspection requirements in MRP-335R3 will provide adequate monitoring of the peened DMWs and RPVHPNs between required inspections. Licensees NRG Report to ASME November 2016 desiring to implement peening to obtain relaxation of examination requirements will still need to submit a plant specific alterative in accordance with 50.55a. Operational Leakage SUSQUEHANNA UNIT 1 (LER 3872016020ROO, ADAMS Accession No. ML16216A378) - REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE AT LOCAL POWER RANGE MONITOR (LPRM) HOUSING AS A RESULT OF IGSCC While performing under vessel inspections, a Reactor Coolant Pressure Boundary (RCPB) leak was found at the Local Power Range Monitor (LPRM) 24-09 housing above the housing flange, on the LPRM housing tube. The leakage was determined to be non-isolable from the reactor vessel. At the time of discovery, Unit 1 was in Mode 4. The cause of the RCPB leakage was determined to be lntergranular Stress Corrosion Cracking (IGSCC). Corrective actions to repair the leak have been completed. In addition, a visual inspection was performed on all the Unit 1 LPRM, Intermediate Range Monitor (IRM), and Source Range Monitor (SRM) In-core monitor housings and no further issues were identified. There was no operational impact as a result of this event due to the plant being in Mode 4 at the time of discovery. This event resulted in an eight (8) hour Emergency Notification System (ENS) communication pursuant to 10 CFR 50.72(b)(3)(ii)(A). ARKANSAS NUCLEAR ONE UNIT 1 (EN 52271) UNISOLABLE LEAK ON DECAY HEAT REMOVAL PIPING DUE TO WELD FAILURE ON A 1" COMMON PIPE While in Mode 6, both trains of Decay Heat (Residual Heat Removal) were declared inoperable due to a cracked weld on a 1" common pipe. The leak developed in a USAS 831 .7, Class1 pipe at a weld upstream of pressure indication isolation valve DH-1037. The leak was not isolable from the common 8-inch Decay Heat piping and encompassed approximately 1/3 [one third] of the pipe circumference. At the time of discovery, the unit was in Lowered Inventory with both Loops of Decay Heat in service. Subsequently, one train of Decay Heat was secured to reduce the likelihood of crack propagation. One Train of Decay Heat remained in service providing the function of removing Decay Heat and the other train remained readily available. The leak was approximately 0.25 gallons per minute with a pipe pressure of 140 psi. Compensatory measures are in place and include an individual posted to watch the pipe in case plugging is necessary. Repairs to the pipe were to be completed once pipe could be drained. WOLF CREEK (EN 52218) TECHNICAL SPECIFICATION REQUIRED SHUTDOWN While operating in MODE 1 at 100 percent rated thermal power and placing Excess Letdown in service for Reactor Coolant System (RCS) leak detection, RCS operational leakage exceeded 1 gpm [gallon per minute] unidentified leakage as identified by performing RCS Water Inventory Balance using the Nuclear Plant Information System Computer. This NRG Report to ASME November 2016 required the Unit to be placed into Mode 3 in 6 hours. Trending of containment sump level indicated the leakage was inside containment with the exact location within containment unknown. The licensee made a containment entry and eventually found the source of the unidentified leakage. While looking down on the vessel head the licensee identified signs of a boric acid leak over a mirrored insulation panel. After removing the panel and using a camera the licensee saw a plume in the area of several penetrations. The licensee was able to determine that the leak was on a core exit thermocouple nozzle threaded connection. The licensee also determined that this was not pressure boundary leakage. In addition, the licensee identified that excess letdown made the leak rate seem worse than the actual value. The leak rate was eventually quantified at around 0.6 gpm. Without being pressure boundary leakage and since the leak rate was less than 1 gpm, the licensee was able to exit the LCO. The licensee has decided to go into their planned refueling outage and will perform some pre-outage surveillances before cooling down to MODE 5. The leak will be repaired during the refueling outage while the head is on the stand. The boric acid deposits on the top of the RPV head by this non-pressure boundary leak have presented significant difficulties in performing the examinations of the RPV head penetrations to demonstrate leakage is not present from the penetrations. CLINTON 1 (EN51939, LER4612016007ROO, ADAMS Accession No. ML16201A232) MAIN STEAM LINE FLEXIBLE HOSE INTERGRANULAR STRESS CORROSION CRACKING IDENTIFIED DURING REFUELING OUTAGE While the plant was in Mode 4 (Cold Shutdown) during refueling outage C1 R16, it was discovered that water was leaking from two separate flexible hoses connecting the main steam line (MSL) to flow instrumentation. Steam flow during power operations is measured in each MSL using instrument taps off the inside and outside of the respective piping elbow. Pressure sensed in each of the lines is used to derive the steam flow. Flexible hose 1B21-D372C - located at the inner elbow on MSL 'B' had water leaking slowly in a thin, steady stream. The leak originated from the collar on the end of the hose closest to MSL 'B'. No mechanical damage was noted on the flexible hose or attached insulation. The vacuum port protective jacket was in place. Flexible hose 1 B21-D372E - located at the inner elbow on MSL 'C' had water dripping out slowly, less than 5 drops/minute. The leak was coming from the area of the vacuum port near the top of the hose, going down the side, and dripping off the bottom. A failure analysis of the flexible hose failures identified the failure mechanism as lntergranular Stress Corrosion Cracking (IGSCC). Both leaking flexible hoses 1 B21-D372C and 1 B21-ID372E were replaced during the refueling outage and their respective high side flexible connections were also replaced. No additional leaks were found during an inspection of other flexible hoses connected to MSLs and the reactor recirculation system. An examination of monitored drywell points prior to plant shutdown for C1 R116 showed no change in temperature, pressure or airborne radiation levels. IGSCC resulted in the failed flexible hose discovered during the C1 R16 walkdown. The root cause evaluation for this event determined that the corrective actions to prevent recurrence of the condition identified June 18, 2007 (LER 2007-003) failed to eliminate or significantly NRG Report to ASME November 2016 reduce below threshold any of the three factors required for IGSCC to exist (susceptible material, tensile stresses, and aggressive environment). The leaking flexible main steam line hoses and the remaining flexible hoses on the MSLs Band C were replaced during C1 R16. The remaining inner elbow flexible hoses on MSLs A and D have been scheduled tor replacement during the next refueling outage C1 R17. A design modification is planned to eliminate or significantly reduce at least one of the three factors required for IGSCC (susceptible material, tensile stress, or corrosive environment) to below the threshold where IGSSC can be initiated. DRESDEN Unit 2 (EN 51934, LER 2372016002ROO, ADAMS Accession No. ML16278A007) HPCI INLET STEAM DRAIN POT PIPING LEAK RESULTING IN HPCI INOPERABILITY A through-wall steam leak was observed in the Unit 2 High Pressure Coolant Injection (HPCI) inlet drain pot drain piping during planned maintenance on Division II of the* Low Pressure Coolant Injection (LPCI) system. The leak was identified to be on the Inlet Drain Pot line upstream of the HPCI Inlet Drain Pot 2A Inboard Drain Valve, Air Operated Valve (AOV) 2-2301 -29, which is ASME Code Class 2 piping. At 1157 CDT, the station entered the Action Statement in the Technical Requirements Manual (TRM) 3.4.a to isolate the adversely affected ASME Code Class 2 component. At 1457 CDT, the flow path containing the leaking pipe was isolated and HPCI was declared inoperable. At this time, the station entered Technical Specification (TS) 3.5.1 Condition K due to the inoperability of HPCI and a division of LPCI. Condition K directed entry into TS Limiting Condition for Operation (LCO) 3.0.3. At 1710 CDT, the LPCI system was restored and TS LCO 3.0.3 was exited but the unit remained in TS 3.5.1 Condition G. At 0042 CDT on 5/18/2016, the adversely affected piping was replaced with stainless steel and HPCI was declared operable which allowed for TS 3.5.1 Condition G to be exited. This event is reportable under 10 CFR 50.73(a)(2)(v)(D), "Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident." The cause of the failure was local internal thinning from a mechanical erosion mechanism. Based on the cratered appearance of the eroded surface, the thinning was due to liquid droplet impingement erosion. The through-wall leak occurred toward the downstream side of the elbow where the droplet impact angle was high (close to 90 degrees). Additional causes may be determined during the investigation. NORTH ANNA UNIT 2 (EN 52137, LER 3392016001ROO, ADAMS Accession No. ML16271A408) - TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO REACTOR COOLANT SYSTEM LEAK Following a containment walkdown to investigate an increase in RCS unidentified leakage to 0.15 gpm, a through wall leak was identified in the controlled bleed-off piping associated with the Reactor Coolant Pump seal for 2-RC-P-1 C. The source of the leakage could be NRG Report to ASME November 2016 isolated and was considered RCS pressure boundary leakage. This was evaluated as RCS pressure boundary leakage and North Anna Unit 2 entered TS 3.4.13.B (RCS Operational Leakage) and commenced a shutdown to mode 5. While in Mode 5, the controlled bleed-off piping associated with the RCP seal for 2-RC-P-1 C was replaced . The direct cause of the RCS unidentified leakage was determined to be a large mean stress placed on the socket weld due to the controlled bleed-off line not being properly aligned in the downstream pipe support, and therefore not allowing for the thermal growth of the RCS. As a result of the large mean stress, a crack initiated at a small defect (lack of fusion) in the toe of the socket weld and propagated through the weld due to normal cyclic vibration from the Reactor Coolant Pump. 50.55a RULEMAKING NRC staff plans to incorporate by reference Nonmandatory Appendix U, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Piping and Class 2 and 3 Vessels and Tanks," of Section XI with conditions in the Final Rule to incorporate the 2009-2013 Editions and Addenda of ASM E B&PV Code. ASME sent the NRC a letter describing the conclusion of their activities for addressing operational leakage of pressure retaining components (ADAMS Accession No. ML15099A624 ). In view of ASME's Pressure Boundary Leakage project team's conclusions, the NR.C sent a letter back to ASME on July 14, 2015. (ADAMS Accession No. ML15188A057).

4. NRO DCIP Quality Assurance and Vendor Inspection Branch Activities NRO Vendor Inspection The NRO vendor inspection program is described in Inspection Manual Chapter (IMC) 2507, "Vendor Inspections." This IMC was last updated on October 3, 2013. This IMC is implemented by various Inspection Procedures (IPs) including:

IP 43002: Routine Inspections of Nuclear Vendors; IP 43003: Reactive Inspections of Nuclear Vendors; IP 43004: Inspection of Commercial-Grade Dedication Programs; IP 43005: NRC Oversight of Third Party Organizations Implementing Quality Assurance Requirements; IP 36100: Inspection of 10 CFR Part 21, Progra ms for Reporting Defects and Noncompliance; IP 37805: Engineering Design Verification Inspections; IMC 0617: Vendor and Quality Assurance Implementation Inspection Reports; and IMC 2507: Vendor Inspections FY 16 Vendor Inspection Plans

• AP1000 modular construction (structural and mechanical)
• AP1000 mechanical and electrical qualification test programs
• Digital Instrumentation and Control for AP1000
• Valve and pump manufacturing NRG Report to ASME November 2016
• Commercial-gra*de dedication organizations
• Reverse engineering activities Vendor Inspection Reports Issued, Completed, and Planned Inspections
• Equipos Nucleares, S.A (ENSA) Santander, Spain - issued
• Electroswitch Corporation, Weymouth, MA- issued
• Lisega, Inc., Kodak, TN - issued
• General Electric (GE) Oil & Gas, Pineville, LA - issued
• SPX, Copes-Vulcan, McKean, PA - issued
• Namco Controls Corporation, Elizabethtown, NC - issued
• Aecon Industrial, Cambridge, Canada - issued
• Westinghouse Electric Company, Warrendale, PA - completed
• Paxton & Vierling Steel Company, Carter Lake, IA, completed
• Mangiarotti S.p.A, Monfalcone, Italy - completed
• Westinghouse Electric Company, Cranberry Township, PA - completed
• Target Rock, Huntsville, AL - planned
• Enercon, Oklahoma City, OK - planned
• Curtiss-Wright EMO, Cheswick, PA - planned
• Creusot Forge, Le-Creusot, France - planned
• PElco, Memphis, TN - planned
• GE Hitachi, Wilmington, NC - planned
• WECTEC, Charlotte, NC - planned Previously issued NRC inspection and trip reports are located at:

http://www.nrc.gov/reactors/new-reactors/oversight/quality-assurance/vendor-insp/insp-reports.html New Vendor Inspection Quality Assurance Website Links The NRC has implemented website pages to make it easier to become familiar with and follow vendor inspection and QA related activities: http://nrcweb.nrc.gov:400/reactors/new-reactors/oversiqht/quality-assurance/vendor-insp.html. As part of the Vendor Outreach and Communications Strategy, the NRC held its 2016 Biannual Vendor Workshop in coordination with NUPIC vendor meeting in St. Louis, MO on June 23, 2016. A total of 422 representatives from 111 countries including United States attended the vendor workshop. Presentations from tlhe 2016 Vendor Workshop and past workshops are available on our public website at the below link. NRG Report to ASME November 2016 http://www.nre.gov/reactors/new-reactors/oversight/qua Iity-ass uran ce/vendor-oversig ht. htm I The Frequently Asked Questions (FAQ) page addresses Quality Assurance for New Reactors and currently has three main categories: 10 CFR Part 21 FAQs, Commercial Grade Dedication FAQs, and Enforcement FAQs. The page provides quick links to questions we have received in the past about the mentioned topics: http://www. nrc.gov/reactors/new-reactors/oversight/quality-assurance/gual-assure-fags.html The web page link below serves as a categorization tool and provides a list of all applicable QA Inspections for New Reactor Licensing and Vendor QA Inspection reports that have either a Notice of Nonconformance (NON) or Notice of Violation (NOV) within a specific criterion of 10 CFR 50 Appendix B or 10 CFR Part 21 related issue. The page is routinely updated with every new inspection report that is released: http://www. nre.gov/reactors/new-reactors/oversight/qua Iity-ass uran ce/non conformances-violation s.htm I The web page link below describes the vendor inspection program (VIP). The VIP verifies that reactor applicants and licensees are fulfilling their regulatory obligations with respect to providing effective oversight of the supply chain. It is accomplished through a number of activities, including: performing vendor inspections that will verify the effective implementation of the vendor's quality assurance program, establishing a strategy for vendor identification and selection criteria, and; ensuring vendor inspectors obtain necessary knowledge and skills to perform inspections. In addition, the VIP addresses interactions with nuclear consensus standards organizations, industry and external stakeholders, and international constituents: http://www.nrc.gov/docs/ML 1607/ML16075A461.pdf

5. New Reactor Licensing Activities As of January 28, 2016, the status of new reactor licensing under 10 CFR Part 52 is as follows:

Design Certification NRC has issued five design certifications to date (ABWR, System 80+, AP600, AP1000 and ESBWR). These are certified in 10 CFR Part 52, Appendices A, B, C, D, and E respectively. The NRC staffs review of the AREVA's EPR (evolutionary pressurized-water reactor design from France) is suspended at the request of the applicant in its letter dated February 25, 2015, until further notice. The NRC staffs review of the Mitsubishi Heavy Industries' US-APWR design certification application (for an advanced pressurized-water reactor design from Japan) is currently on hold at the request of the applicant except for a few key areas. The NRC staff completed its review of General Electric-Hitachi's ESBWR (first passive BWR) and issued its final safety evaluation report (FSER) in March 2011. On March 24, 2011, the NRC issued in the Federal Register a proposed rule (76 FR 16549) for public NRG Report to ASME November 2016 comment on the ESBWR design certification. The NRC final rule adding Appendix E to 10 CFR Part 52 to certify the ESBWR standard design was published on October 15, 201 4 in the Federal Register (79 FR61983) and became effective on November 14, 2014. The Korea Hydro and Nuclear Power (KHNP) submitted a standard design certification application for its APR-1400 standard plant design to the NRC on September 30, 2013. The NRG staff conducted an acceptance review of the application for completeness, technical adequacy, and acceptability for docketing. In a letter to KHNP dated December 19, 2013, the NRC staff discussed the results of its acceptance review. The NRG noted that it decided not to accept the application for docketing at that time because the application was not ready in several key areas. The NRC staff continued pre-application interactions with KHNP to support preparation of a complete application by December 2014. On December 23, 2014, KHNP resubmitted the standard design certification application for its APR-1400 design. The NRC staff accepted the APR1400 design certification application for docketing in its letter dated March 4, 2015, based on its determination that the application is sufficiently complete and technically adequate to allow the NRC staff to conduct its detailed technical review. In addition, the NRC staff is reviewing two applications for design certification renewal:

• ABWR GE-Hitachi (application submitted on December 7, 2010)
• ABWR GE-Toshiba (Revision 1 to application submitted on June 22, 2012)

Earlv Site Permits (ESPs) NRC has issued four ESPs (Clinton, Grand Gulf, North Anna, and Vogtle). The NRC's issuance of the Vogtle ESP on August 26, 2009, was the first based on a specific technology (AP-1000) and the first to include a limited-work authorization (LWA). The NRC received an application for an ESP for the Victoria County Station submitted by Exelon on March 25, 2010. The site is located in Victoria County, Texas, with no specific technology selected. On August 28, 2012, Exelon requested withdrawal of the Victoria County Station ESP application from the docket. By letter dated October 3, 2012, NRC accepted the applicant's request, and the application was withdrawn. The NRG received an ESP application for the PSEG site in New Jersey (same site as Hope Creek and Salem 1&2). The ESP application was tendered on May 25, 2010, and was docketed on August 4, 2010. This application uses the Plant Parameter Envelope (PPE) approach which means no specific reactor design has been selected. The NRC recently issued the final Environmental Impact Statement (EIS) for the ESP. The NRG is currently preparing for an Atomic Safety Licensing Board mandatory hearing on the permit application. The hearing will determine whether the staffs environmental review, documented in the final EIS, and the safety review, documented in the Final Safety Evaluation Report, support the findings necessary to issue the permit. Combined License (COL) Applications NRC is currently reviewing 9 COL applications (14 new reactor units): 3 AP-1000: William S. Lee Station 1&2, Shearon Harris 2&3*, Levy County 1&2, Be llefonte 3&4*, and Turkey Point 6&7 NRG Report to ASME November 2016 2 ESBWR: Fermi 3, North Anna 3, Grand Gulf 3*, River Bend 3*, Victoria County 1 and 2** 2 EPR: Calvert Cliffs 3**, Bell Bend*, Nine Mile Point 3**, Callaway 2* 1 US-APWR: Comanche Peak Units 3 and 4

• NRC staff review suspended at request of applicant.
  • Application withdrawn.

On June 8, 2015, Unistar requested to withdraw the Calvert Cliffs, Unit 3 combined license application. On April 25, 2013, Dominion Virginia Power revised its technology selection from the US-APWR nuclear technology and selected the GEH ESBWR nuclear technology for the North Anna Unit 3 project. The initial phase of the North Anna Unit 3 combined license application was submitted to the NRC in July 2013, and the final portion of the application was submitted in December 2013. The NRC issued the combined license and limited work authorization for Vogtle Electric Generating Plant, Units 3&4 on February 10, 2012. The Vogtle plants reference the AP1000 design certification amendment. It was the first combined license issued by the NRC to construct and operate a nuclear power plant under the alternative licensing process in 10 CFR Part 52. It is the first time since 1978 that the NRC issued a license to construct a nuclear power plant in the United States. The NRC staff issued the combined license for V.C. Summer 2&3 on March 30, 2012. The V.C. Summer 2&3 plants reference the AP1000 design certification amendment. On February 4, 2015, the NRC Commissioners held a mandatory hearing on the combined operating license (COL) for Fermi, Unit 3. On May 1, 2015, the NRC issued the combined license for Fermi, Unit 3. This is the first combined liicense for an application referencing the ESBWR design. On November 19, 2015, the NRC Commissioners held a mandatory hearing on the combined licenses for South Texas Project, Units 3 and 4 referencing the GE-Toshiba ABWR Design. On February 12, 2016, the NRC issued the combined license for South Texas Project, Units 3 and 4. This is the first combined license for an application referencing the GE Toshiba ABWR design. Advanced Reactors Program NRC established an advanced reactors program in the Office of New Reactors. Currently, there are no applications under review, but several applications are expected in the next three years including:

• Integral PWRs (iPWRs):
• NuScale (iPWR) - NuScale Power is developing a modular, scalable 50 MWe iPWR. Pre-application reviews are currently under discussion. The design certification is expected to be submitted to the NRC in November or December of 2016.

NRG Report to ASME November 2016

• B&W mPower (iPWR) - B&W is developing a modular, scalable 180 MWe iPWR.

At this time, mPower has reduced its activities in the mPower development, and have not provided a submittal date for the application.

• TVA is planning to submit its early site permit in the second quarter of 2016 for its Clinch River site near Oakridge, Tennessee.
• Holtec is developing the Holtec Inherently Safe Modular Underground Reactor SMR 160 design that has a 160 MWe electrical power output. They plan to pursue a Part 50 licensing process that requires an applicant to apply for a construction permit and a subsequent operating license. They have not provided an application submittal date.
• XEnergy has indicated it plans to submit a design certification application to the NRC within the next few years for its pebble-bed high temperature gas-cooled reactor. The XEnergy reactor (Xe-100) is a helium-cooled reactor with a power rating of 125 MWt.
• Advanced Reactor (non-light water reactors) Guidance Development:
• NRC has received Idaho National Laboratory (INL) generated Department of Energy technical report "Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors." The INL report is the culmination of phase one of a two-phase initiative by the DOE and the NRC to develop advanced reactor safety design criteria from which the principal design criteria could be derived for advanced reactor concepts. The NRC will follow its normal process for developing and issuing regulatory guidance and anticipates completion of such guidance by the end of 2016.

6. Multinational Design Evaluation Program CMDEP) Activities MDEP is a multinational initiative to develop innovative approaches to leverage the resources and knowledge of mature, experienced national regulatory authorities who are tasked with the regulatory design review of new reactor plant designs. Some of the issue-specific working groups established under the MDEP organization that the NRC participates in are the Codes and Standards Working Group (CSWG), whose goal is to achieve harmonization of code requirements for pressure-boundary components, and the Vendor Inspection Cooperation Working Group (VICWG), whose goal is to maximize the use of the results of inspections obtained from other regulators' efforts in inspecting vendors.

Vendor Inspection Cooperation Working Group (VICWGJ The MDEP VICWG was formed because component manufacturing is currently subject to multiple inspections and audits similar in scope and in safety objectives, but conducted by different regulators to different criteria. The primary goal of the VICWG is to maximize the use of the results obtained from other regulators' efforts in inspecting vendors. The MDEP VICWG continues to achieve its short-term goals and is making progress towards achieving its long term goals. The VICWG continues to focus on maximizing information sharing, joint inspections (multiple regulators inspecting to the regulatory requirements of one country), and witnessing of other regulators' inspections. The NRC participated in 6 witnessed and joint inspections. NRG Report to ASME November 2016 The working group enhances the understanding of each regulator's inspection procedures and practices by coordinating witnessed inspections of safety related mechanical pressure retaining components (Class 1) such as pressure vessels, steam generators, piping, valves, pumps, etc., and quality assurance inspections. Witnessed inspections consist of one regulator performing an inspection to its criteria, observed by representatives of other MDEP countries. The benefits to the observing countries include additional information and added confidence in the inspection results . MDEP regulators are using the experience gained during conduct of VICWG witnessed inspections in their inspection planning. The MDEP VICWG held its 171h meeting during the week of April 4 in Tokyo, Japan. This meeting included members from France, Canada, Japan, the Republic of Korea, South Africa, Finland, the Russian Federation, the United Kingdom and the United States. Sweden, United Arab Emirates, India, Canada, Finland, and Turkey were not in attendance. Because the meeting was hosted in Dijon, France it allowed multiple members of ASN to participate in the meeting. The group discussed planned inspections and reviewed the inspection lists presented by the US, France and Korea. The group discussed areas of common interest (i.e., counterfeit, suspect, and fraudulent items (CSFI) and reverse engineering) and identified several inspection activities that could be conducted as MDEP activities. The NRC will be supporting a Multinational inspection at Creusot Forge in November. Canada also proposed inspections at Velan valves and KINS discussed Dresser and Siemens. The members are continuing to communicate by e-mail to plan and conduct inspections. Codes and Standards Working Group (CSWG) The MDEP group's goal is to harmonize and converge national codes, standards, and regulatory requirements and practices applicable to pressure boundary components while recognizing the sovereign rights and responsibilities of national regulators in carrying out their safety reviews of new reactor designs. The CSWG published several reports on codes and standards related to pressure boundary components, and it provides a regulatory forum for groups such as the World Nuclear Association's Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group to coordinate with concerning international harmonization efforts. In November 2015, the CSWG met with representatives from CORDEL and the SDOs. Representatives from CORDEL presented and discussed their harmonization efforts concerning NOE Personnel Qualifications; Non-Linear Analysis Design Rules; and Welding and Welding Qualifications with the CSWG. In addition the CORDEL representatives presented on the results of their meeting in China with Chinese counterparts. The representatives from the SDOs presented the status of their harmonization activities through their Code Convergence Board. CSWG will continue to follow closely the activities of the SDOs and CORDEL through 2016, at which time the CORDEL pilot program for convergence using the SDO Convergence Board process is expected to complete at least one code convergence. Also, the CSWG may discuss the possibility of turning over its regulatory interface with the SDOs and CORDEL's activities to another international regulatory organization (e.g., NEA's Committee on the Safety of Nuclear Installations), as many of the topics are growing beyond the MDEP mandate. NRG Report to ASME November 2016

7. 10 CFR Part 21 Rulemaking The NRC staff is currently reviewing the Revision 1 of NEI 14-09, "Guidelines for Implementation of 10 CIFR Part 21, Reporting of Defects and Noncompliance," dated February 2016, which the NRC plans to endorse in a Regulatory Guide. The NRC staff has completed draft guide DG-1291 , "Evaluating Deviations and Reporting Defects and Noncompliance."

Based on Project AIM recommendations and rebalancing of agency's work load, Part 21 rulemaking has been cancelled. The NRC will continue to work on issuing DG-1291 for public comment later this year and hopes to issue the final regulatory guide in 2017.

8. Commercial Calibration Services Status By letter dated April 29, 2014, the Nuclear Energy Institute (NEI) submitted Revision 0 of NEI 14-05, "Guidelines for the Use of Accreditation i11 Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services," to the U.S. Nuclear Regulatory Commission (NRC) for NRC staff review and endorsement. NEI 14-05 provides an approach for licensees and suppliers of basic components for using laboratory accreditation by Accreditation Bodies (ABs) that are signatories to the International Laboratory Accreditation Cooperation (ILAC) Mutual Recognition Arrangement (MRA)

(hereby after referred to as the ILAC accreditation process) in lieu of performing commercial-grade surveys for procurement of calibration and testing services performed by domestic and international laboratories accredited by ILAC signatories. By letter dated February 9, 2015 (ADAMS Accession No. ML14322A535), the NRC staff transmitted its safety evaluation (SE) identifying the guidelines contained in NEI 14-05, Revision 1 (ADAMS No. ML14245A391) as an acceptable approach for licensees and suppliers to meet the requirements of 10 CFR Part 50, Appendix B to use ILAC laboratory accreditation as part of the commercial-grade dedication process for procurement of calibration and testing services. NRC's endorsement of NEI 14-05, Revision 1, expands the NRC's acceptance of the ILAC accreditation process first documented in SE on an Arizona Public Service request (ADAMS Accession No. ML052710224). NRC's earliier acceptance was limited to calibration laboratory services accredited by specific U.S. Accrediting Bodies (ABs). The SE (1) confirms that NEI 14-05, Revision 1, reflects the ILAC accreditation process previously approved; (2) provides an evaluation of the unique aspects of NEI 14-05, Revision 1; (3) constitutes formal NRC endorsement. of the guidelines in NEI 14-05, Revision 1, for using the ILAC accreditation process in lieu of performing a commercial-grade survey; and (4) finds that the ILAC accreditation process continues to satisfy the requirements of Appendix B to 10 CFR Part 50 and, therefore, is acceptable. On March 16, 2016, the NRC issued Regulatory Issue Summary (RIS), 2016-01, "Nuclear Energy Institute Guidance for the Use of Accreditation in Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services." The NRC staff is issuing this RIS to notify addressees of one method found acceptable by the NRC staff for procurement of calibration and testing services performed by domestic and international laboratories for use in safety-related applications. Both domestic and international laboratories are required to be accredited by accreditation bodies (ABs) that are signatories to the International Laboratory Accreditation Cooperation (ILAC) Mutual Recognition NRG Report to ASME November 2016 Agreement (MRA) (hereafter referred to as the ILAC accreditation process) in order for licensees and suppliers of basic components to use these services in lieu of performing commercial-grade surveys. On April 1, 2016, the NRC staff approved the license amendment request submitted by Union Electric Company (dba, Ameren Missouri, the licensee) to change the operating quality assurance program revision 31, for Callaway Plant, Unit 1, to adopt NEI 14-05. Revision 1.

9. NRC Staff Review of EPRI 1025243 Guideline for Commercial-Grade Design and Analysis Computer Programs By letter dated July 18, 2012, the Nuclear Energy Institute (NEI) submitted Electric Power Research Institute (EPRI) 1025243, Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Related Applications for staff review and approval. EPRI 1025243 describes a dedication methodology for commercial-grade design and analysis computer programs for use in meeting regulatory requirements. EPRI 1025243 follows the method provided in EPRI NP-5652, which the NRC conditionally endorsed in Generic Letter 89-02.

On July 1, 2015, the NRC issued for public comment DG-1305, "Acceptance of Commercial-Grade Design and Analysis Computer Programs for Nuclear Power Plants." The DG provides new guidance that describes acceptance methods that the NRC staff considers acceptable in meeting regulatory requirements for acceptance and dedication of commercial-grade design and analysis computer programs for nuclear power plants. The DG-1305 public comment period is now closed. The NRC staff has received and evaluated 40 comments from NEI,, other stakeholders and the public. Regulatory Guide 1.231, "Acceptance of Commercial Grade Design and Analysis Computer Programs for Nuclear Power Plants," is in the final review process and should be issued in the fall. This will be the first software-specific Commercial-Grade Dedication (CGD) guidance with a limited scope for safety-related use of commercial Design & Analysis Computer Programs.

10. NRC Staff Review of EPRI Guideline for Dedication of Commercial-Grade Items for Use in Nuclear Safety-Related Applications In September, 2014, the EPRI issued the 2014 Technical Report 3002002982, "Plant Engineering: Guideline for the Acceptance of Commercial-Grade Items in Nuclear Safety-Related Applications" - Revision 1 to EPRI NP-5652 and TR-102260. The NRC staff participated during many of the EPRI technical advisory group (TAG) meetings held at EPRl 's offices in Charlotte, North Carolina.

This report describes a methodology that can be used to dedicate commercial-grade items for use in safety-related applications. The scope of applications for which commercial-grade item dedication is used has evolved significantly since the EPRI published its reports Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications (NCIG-07) (NP-5652) and Supplemental Guidance for the Application of EPRI Report NP-5652 on the Utilization of Commercial Grade Items (TR-102260) in 1988 and 1994, respectively. The guidance in this final report reflects lessons learned and addresses NRG Report to ASME November 2016 challenges that have been identified through expanded use of the original guidance. This report supersedes both original reports in their entirety. Draft Guide DG-1292, "Dedication of Commercial-Grade Items for Use in Nuclear Power Plants," was issued on June, 30, 2016, for a 60 day public comment period. This new RG approves for use, in part, Revision 1 to the EPRI NP-5652 and TR-102260, with respect to acceptance of commercial-grade dedication of items used as basic components for nuclear power plants.

11. NRC Staff Interface with Nuclear Utilities Procurement Issues Committee (NUPIC)

During the weeks of June 20 and October 17, 2016, the NRC staff participated and made presentations at the NUPIC General Membership Meeting and the Annual Vendo1r Meeting in St. Louis, MO, and the NUPIC General Membership meeting in Minneapolis, MN, respectively. The NRC addressed ongoing staff initiatives including an update on vendor inspection activities and key findings from those inspections. The NRC periodically accompanies a NUPIC Joint Utility Audit team to observe selected audits and ensure that the audit process remains an acceptable alternative to the NRC's vendor inspection/audit program. The NRC staff continues to rely on the effectiveness of the NUPIC joint utility audit process for evaluating the implementation of quality assurance programs of suppliers to the nuclear industry. The NRC issues trip reports to document NRC observation of audits performed by NUPIC that are available at the below web-site link: http://www.nrc.gov/reactors/new-reactors/oversightlquality-assurance/nupic-industrv.html

12. Reverse Enineering lnfortmation Notice 2016-01 On July 15, 2016, NRC Information Notice 2016-09: Recent Issues Identified when using Reverse Engineering Techniques in the Procurement of safety-related Components was issued. The NRC is issuing this information notice to inform addressees of issues that the NRC staff has identified concerning the supply of replacement safety-related components.

Specifically, this IN describes instances where reverse engineering techniques were used to manufactured replacement components, and where the components were supplied without first verifying the supplied components met all safety-related design requirements. The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. Reverse engineering was discussed at one on the breakout sessions at the June NRC Workshop on Vendor Oversight.

13. License Renewal Activities Following are on-going activities related to license renewal:

Current status of applications, staff reviews and approvals

• 83 units approved (81 operating plants with renewed licenses) o 1 (2 units) in hearings (Indian Point 2 & 3) - supplemental SER issued November 2014, hearings held November 2015 o 3 (4 units) completed ACRS Full Committee meeting (Fermi, Grand Gulf, LaSalle 1 & 2)

NRG Report to ASME November 2016 o 2 (3 units) awaiting follow-up ACRS Subcommittee (Seabrook [TBD], Diablo Canyon 1 & 2 [TBD]) o 2 (3 units) awaiting ACRS Subcommittee (Waterford 3 [7/2017), South Texas Project 1 & 2 [11/2016))

• 1 application (1 unit) with scheduled submittal in 2016:

o April to June 2017 - River Bend o October 2019- Perry o January to March 2021 - Clinton o April to June 2022 - Comanche Peak 1 & 2 Forty-five units have entered the operating period beyond 40 years: o Oyster Creek - April 9, 2009 o Indian Point 2 - September 28, 2013 o Nine Mile Point 1 - August 22, 2009 o Oconee 2 - October 6, 2013 o Ginna - September 18, 2009 o Browns Ferry 1 - December 20, o Dresden 2 - December 22, 2009 2013 o H.B. Robinson - July 31, 2010 o Cooper Nuclear Station - Jan. 18, o Monticello - September 8, 2010 2014 o Point Beach 1 - October 5, 201 O o Duane Arnold - February 21 , 2014 o Dresden 3 - January 12, 2011 o Three Mile Island 1 - April 19, 2014 o Palisades - March 24, 2011 o ANO 1 - May 20, 2014 o Vermont Yankee - March 21, 2012 o Browns Ferry 2 - June 28, 2014 o Surry 1 - May 25, 2012 o Peach Bottom 3 - July 2, 2014 o Pilgrim - June 8, 2012 o Oconee 3- July 19, 2014 o Turkey Point 3 - July 19, 2012 o Calvert Cliffs 1 - July 31 , 2014 o Quad Cities 1 - December 14, 2012 o Hatch 1 - August 6, 2014 o Quad Cities 2- December 14, 2012 o FitzPatrick - October 17, 201 4 o Surry 2 - January 29, 2013 o DC Cook 1 - October 25, 2014 o Oconee 1 - February 6, 2013 o Prairie Island 2 - October 29, 2014 o Point Beach 2- March 8, 2013 o Brunswick 2 - December 27, 2014 o Turkey Point 4 - April 10, 2013 o Millstone 2 - July 31, 2015 o Peach Bottom 2 - August 8, 2013 o Indian Point 3 - December 12, 2015 o Fort Calhoun 1 - August 9, 2013 o Beaver Valley - January 29, 2016 o Prairie Island 1 - August 9, 2013 o St. Lucie 2 - March 1, 2016 NRG Report to ASME November 2016 o Browns Ferry 3 - July 2, 2016 o Brunswick 1 - September 8, 2016 o Calvert Cliffs 2 - August 13, 2016 o Salem 1 - August 13, 2016 Technical Issues

• Steam Generator Divider Plates and Tube-to-Tubesheet Welds

           -   Steam Generator Task Force submitted a topical report to address necessity for these inspections
           -  LR-ISG-2016-01 comment period ended on July 7, 2016. Staff is evaluating comments and expects to issue final guidance this fall.

• PWR Vessel Internals (MRP-227-A)

           -   Staff evaluating revision 1
           -   Staff has developed an approach for aging management of PWR vessel internals for subsequent license renewal using the existing MRP-227-A program and a gap analysis to address expected aging differences between 60 and 80 years.

Subsequent License Renewal The NRC issued for public comment GALL and SRP reports to address subsequent license renewal (SLR), for plant operation to 80 years:

• Draft NUREG-2191, Volumes I and II (December 2015), "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Draft Report for Comment" (ADAMS Accession Nos. ML15348A111 and ML15348A153)
• Draft NUREG-2192 (December 2015), "Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, Draft Report for Comment" (SRP-SLR) (ADAMS Accession No. ML15348A265)

The staff received public comments on February 29, 2016 and is currently evaluating the input. The SRP-SLR Report, the GALL-SLR Report, and a NUREG report on "Disposition of Public Comments and Technical Bases" are schedulled to be published in July 2017. More information on subsequent license renewal, including detailed information on public meetings, can be found at: http://www.nrc.gov/reactors/operating/licensing/renewal/subseguent-license-renewal.html Applications for subsequent license renewal are expected in the third quarter of 2018 (Peach Bottom Atomic Power Station, Units 2 and 3) and by the end of the first quarter of 2019 (Surry Power Station, Units 1 and 2). Research Activities The NRC's Office of Nuclear Regulatory Research (RES) issued the following report related to license renewal and aging management:

• NUREG/CR-4513, Revision 2, "Estimation of Fracture Toughness of Cast Stainless Steels during Thermal Aging in LWR Systems." ML16145A082 (April 2016).

14. New Generic Letters Since the last Code Week, no Generic Letters (GL) were issued.

15. New Information Notices Since the last Code Week, the following Information Notice (IN) was issued:

i. No recently issued INs of interest.

16. New Regulatory Issue Summaries Since the last Code Week, the following Regulatory Issue Summaries (RIS) were issued:

• No recently issued RISs of interest.

17. NRC Publications of Potential Interest to ASME Since the last Code Week, the following publication that may be of interest to ASME was issued:

i. WASH-1400 (NUREG/KM-0010), The Introduction of Risk Assessment to the Regulation of Nuclear Reactors, August 2016

• This is an update to WASH-1 400 incorporating information from a November 2015 presentation, "WASH-1400 and the Origins of Probabilistic Risk Assessment in the Nuclear Industry,"

ii. NUREG/BR-0292, Revision 1, Safety of Spent Fuel Transportation, August 2016

• This publication explains the NRC's role in the safe packaging and transport of spent nuclear fuel from commercial nuclear power plants. The NRC oversees the design, manufacture, use, and maintenance of containers for these radioactive shipments.

iii. NUREG-2201, Probabilistic Risk Assessment and Regulatory Decision Making: Some Frequently Asked Questions, September 2016

• Probabilistic risk assessment (PRA) is an important decision-support tool at the U.S. Nuclear Regulatory Commission. The availability of experiential data for accidents, including those at the Fukushima Dai-ichi nuclear power plant, raises natural questions regarding the need for and utility of PRA, which is, at heart, a systems modeling-based analytical approach. This report addresses these questions using the format of frequently asked questions (FAQs).

iv. DG-1331, Service level I, II, Ill and In-Scope License Renewal Protective Coatings Applied to Nuclear Power Plants

• This regulatory guide is the proposed revision 3 to Regulatory Guide 1.54. The revision addresses updated references related to coatings and expands the guidance to include internal coatings on in-scope (for license renewal purposes) components. This revision also provides guidance for new reactor designs with the recognition that the licensee or applicant may need to adj ust some features based on the particular plant design.

v. NUREG/CR-7217, Application of Automated Analysis Software to Eddy Current Inspection Data from Steam Generator Tube Bundle Mock-up, September 2016

• This report documents the results of evaluations of computerized data screening software used for analyzing eddy current data obtained during the inspection of steam generator tubes.

vi. Revision to Regulatory Guide (RG) 1.28, "Quality Assurance Program Criteria (Design and Construction)" (Draft RG 1326)

• The NRC staff continues to endorse the previous guidance in the current RG 1.28, Quality Assurance Program Criteria (Design and Construction), Revision 4 , issued in June 2010, and is not aware of any issues that would preclude its use.

Revision 4, of RG 1.28 extended the scope of the NRC's endorsement to include NQA-1, Part II. Part II contains amplifying QA requirements for certain specific work activities that occur at various stages of a facility's life. The work activities include, but are not limited to, management, planning, site investigation, design, computer software use, commercial-grade dedication, procurement, fabrication, installation, inspection, and testing.

• In June 2015, the NRC staff completed a review and identified that differences exist between the previously NRC accepted guidance (NQA-1-2008 and NQA-1a-2009 addenda) and the most recently issued guidance from the ASME (NQA-1b-2011, NQA-1 -2012 and NQA- 1-2015). Therefore, the staff has developed draft RG-1326 with the intent to approve for use, with several regulatory positions, the guidance from ASME NQA-1b-2011 , NQA-1-2012 and NQA-1-2015.
• The NRC expects to issue draft RG-1326 for public comment by the end of the year and hopes to issue the final regulatory guide in 2017.

18. Upcoming Public Meetings of Potential Interest to ASME The following public meetings, either upcoming or recently transpired, may be of interest to ASME:

i. To discuss th,e outcome of the staffs review of the potential optimization of the subsequent license renewal application review process. 11/10/16, 9:00 AM - 12:00 PM, Teleconference Refer to the NRC Public Meeting Web Page at http://meetings.nrc.gov/pmns/mtg for a list of all currently-scheduled public meetings and further details.

From: lingam, Siva Sent: 18 Oct 2016 15:38:05 +0000 To: Pascarelli, Robert

Subject:

RE: Wolf Creek Revised RRs - Internal Discussion FYI only. From: Taylor, Nick Se nt: Tuesday, October 18, 2016 11:35 AM To: lingam, Siva <Siva.Lingam@nrc.gov>

Subject:

RE: Wolf Creek Revised RRs - Internal Discussion Thanks Siva. I didn't realize that, and neither of them appeared to know about the call when I asked them. Perhaps they just weren't watching their email carefully. From: Lingam, Siva Se nt: Tuesday, October 18, 2016 10:32 AM To: Taylor, Nick <Nick.Taylor@nrc.gov> Subje ct: FW: Wolf Creek Revised RRs - Internal Discussion For the internal call held yesterday at 9:00 AM (Eastern), I did include both the resident inspectors and David Proulx (see the attached scheduler).

From: lingam, Siva Sent: 18 Oct 2016 07:45:11 -0400 To: Tsao, John

Subject:

RE: Wolf Creek--Acceptance Review for Relief Request 14R-04 Alternate examination of CROM nozzles (MF8456) Thank you. From: Tsao, John Sent: Tuesday, October 18, 2016 7:43 AM To: Singal, Balwant <Balwant.Singal@nrc.gov>; Lingam, Siva <Siva.lingam@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Collins, Jay <Jay.Colliins@nrc.gov>

Subject:

Wolf Creek--Acceptance Review for Relief Request 14R-04 Alternate examination of CROM nozzles (MF8456) Balwant & Siva, Below is my input for the acceptance review of the subject relief request. By letter dated October 11 , 2016, Wolf Creek Nuclear Operating Corporation {the licensee) submitted Relief Request 14R-04 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration numbers 77 and 78 at the Wolf Creek Generating Station. In accordance with Nuclear Regulatory Commission's (NRC's) process as described in LIC-109, "ACCEPTANCE REVIEW PROCEDURES," the NRC staff has performed an acceptance review to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review was also intended to identify whether the request has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant. The NRC staff has concluded that the subject relief request does provide technical information in sufficient detail to enable the NRC staff to proceed with its detailed technical review and make an independent assessment regarding the acceptability of the proposed relief request in terms of regulatory requirements and the protection of public health and safety and the environment. If needed, the NRC staff may request for additional information to complete its technical review.

From: Cumblidge, Stephen Sent: 19 Oct 2016 15:32:01 -0400 To: Sengupta, Abhijit Cc: Collins, Jay

Subject:

RE: sharepoint link Attachments: Volumetric Leakage Path.pptx Here it is. Stephen Cumblidge Materials Engineer US Nuclear Regulatory Commission Mail Stop OWFN/9 H6 Washington, DC 20555-0001 Telephone: (301) 415-2823 (Office) From: Sengupta, Abhijit Sent: Wednesday, October 19, 2016 2:55 PM To: Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>

Subject:

sharepoint link Stephen How are you. Could you please send me the sharepoint link where presentations are stored from today's call.

Thanks, Abhijit

..c I ' ro 0.... Q) C) c I ' ro ID ~ E ID _J en en -~ h en

      ~

Q) <:( E

J 0

Ultrasonic inspections Interference fl~;,~ __:__ -.~ai.-. performed to see flaws in welds and penetration tubes need to scan above and below the weld, No Leak as the weld is not straight. This scanning, as an unintentional byproduct, produces images from the ultrasound reflecting from the interference fit region. It did not take long for people to figure out that leaking nozzles produced different patterns in Leak the interference fit than non-leaking nozzles.

So, what is going on? Some reflection and some transmission will occur at the interference fit. The amount of sound reflected is affected by the local tightness of the fit, the local smoothness of the metals, and the local presence of boric acid.

                       ?%
                       ~

(((((

                      )))))))}))I) ) ) ) )

Very Little reflection

                        ~  0% Weld
                      ))l ))))}})))))})))))))))))

Cladding (Stainless St __,,,,-- Butter (Alloy 82/182)

100%

• Air Total reflection

Ultrasound is sensitive to changes in the interference fit as the two metal surfaces are in tight contact. The surfaces were not made mirror-smooth prior to the interference fit, so some odd features will be present. Even so, notches, deep scratches, and a contractor scribing "PNNL" in an interference fit can be clearly detected.

           'l... t*-IJM*  1---.... ,.,. , ,.__. . . . '-'l . . PH*t ..... - .... ,.,,.... .~ 1                               ~ ' l r.:..- bt 0

0

                                                                                                                                             ~

00 0 3 3 _..._ l_ ...

                                                                                                                                             )>

x

           ....,,.. uu -
           ,.....,,......                                                                                                              l--t  Q.)

0 to 170 deg. Circu mfere nce

Interference fits without leaks can still have odd features, depending on the smoothness and how the data was collected. False positives are possible if there are gouges and false negatives are possible if thee is little boric acid present. Interference fits with no leakage present

Leaks can produce odd patterns in the ultrasonic examinations of the interference fit. The random-looking patterns imaged by the volumetric leak path assessments can be reproduced . The general pattern remains the same, although different frequencies or methods (Zero degree vs. TOFD) may result in some differences. vVestiughouse Data PNNL 2.25 MHz Data

   )~ *~--

w* r ==" -

                       'wetted Side                   Wetted Side

In this case PNNL used a 5 MHz zero-degree probe to inspect the interference fit. Their results closely match industry scans of the same nozzle, with higher resolution and greater sensitivity. u ~ ~

The patterns in the UT images are apparently caused by the presence and absence of boric acid deposits that couple ultrasound through the interference fit. 135 Degrees

The High resolution data closely matches the boric acid pattern in Nozzle 63 from North Anna. Reflections come from areas with little or no boric acid and areas with more boric acid are detectable as areas of greater transmission.

Cone Iusions Volumetric Leakage Path Assessments can be effectively used to detect boric acid in the interference fit Volumetric leak path Assessments can give ambiguous results, but has been largely reliable ASME has decided not to qualify Volumetric Leakage Path Assessments Further Reading:

• Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation
• NUREG/CR-6996 Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechanism Penetrations
• Materials Reliability Program : Volumetric Leak Path Assessment for Vessel Upper Head Penetrations (MRP

  -249}

From: Drake, James Sent: 19 Oct 2016 14:34:44 -0500 To: Collins, Jay

Subject:

RE: Fw: See Attached I think the response is adequate. Jim From: Collins, Jay Sent: Wednesday, October 19, 2016 2:34 PM To: Drake, James <James.Drake@nrc.gov>

Subject:

RE : Fw: See Attached I was going with the "If so, then that is not being done." And answering "it is not so. Do you have any other questions or comments?" Is that legitimate or not? I do get questions now and then from stakeholders. Jay From: Drake, James Sent: Wednesday, October 19, 2016 3:31 PM To: Collins, Jay <Jay.Collins@nrc.gov>

Subject:

RE: Fw: See Attached Sounds like it addresses his concern. However, This sounds like an allegation. Does headquarters have an allegation process? Jim From: Col lins, Jay Sent: Wednesday, October 19, 2016 10:52 AM To: Burnell, Scott <Scott.Burnell@nrc.gov> Cc: Drake, James <James.Drake@nrc.gov>

Subject:

FW: Fw: See Attached My draft response, if you guys have comments or suggestions .... Greetings, Thank you for your question. The initial concise answer to your question is, no. The size of the main piping loop or the hole in the piping system does not determine the weld categorization size for inspection requirements either under the American Society of Mechanical Engineer's Boiler and Pressure Vessel (ASME) Code or ASME Code Case N-770-1 as described in Regulatory Issue Summary 2015-010, "Applicability Of ASME Code Case N-770-1 As Conditioned In 10 CFR 50.55a, "Codes And Standards," To Branch Connection Butt Welds." (http://www.nrc.gov/docs/ML1506/ML15068A131 .pdf) The inspection size category for the branch connection weld is determined in accordance with 10 CFR 50.55a, the ASME Code, and Owner Requirements. Typically the Owner utilizes the allowances by the ASME Code to set the

size of the branch connection weld, for its inspection category, to the size of the process or branched pipe. In addition, ASME Code Case N-770-1 only addresses ASME Code Class 1 pressurized water reactor piping and vessel nozzle butt welds fabricated with Alloy 82/182 weld filler metal. There are some dissimilar metal nickel alloy branch connection butt welds in the ASME Class 1 piping of Babcock & Wilcox pressurized water reactor designs, but the process or branched pipe size for each of these locations is less than NPS 2. Therefore, each of these welds falls out of the volumetric inspection requirement of ASME Code Case N-770-1 , as mandated by 10 CFR 50.55a(g)(6)(ii)(F). However, these branch connection welds do receive visual inspections and are part of the system pressure tests each time the reactor restarts after a scheduled refueling outage. If you have additional information or concerns, please feel free to contact us, Jay Collins (301 )415-4038 From: Sam nuke [7] Sent: Wednesday, October 19, 2016 10:56 AM To: Drake, James <James.Drake@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov>

Subject:

[External_Sender] Fw: See Attached From: Sam nuke Sent: Wednesday, October 12, 2016 1:45:58 PM To: Jay.collins@nrc.gov

Subject:

Fw: See Attached Mr Collins, several B&W plants have outages and have not inspected their branch connections in which the hole in the pipe is greater than 2" with a volume metic method, if the process p iping is less than 2" and the branch weld is greater than 2" does that branch weld require a volumetric exam? If so that is not being done From: Sam nuke Sent: Sunday, September 18, 2016 6:08:45 PM To: James.Drake@nrc.gov

Subject:

Fw: See Attached From: Sam nuke Sent: Sunday, September 18, 2016 4:37 PM To: James.Drake@nrc.cov

Subject:

See Attached

From: lingam, Siva Sent: 19 Oct 2016 15:12:17 -0400 To: Collins, Jay;Tsao, John;Taylor, Nick; Drake, James

Subject:

RE: Relief Request Number 14R-03, Request for Relief from Paragraph-3200(b) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds and Relief Request 14R-04, Request for Relief from the Requirements of ASME Code Case N-729-1. FYI From : E-RIDS3 Resource Sent: Wednesday, October 19, 2016 3:07 PM To: WolfCreekEIS Resource <WolfCreekEIS.Resource@nrc.gov>; Watford, Margaret

<Margaret.Watford@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; RidsRgn4MailCenter Resource <RidsRgn4MailCenter.Resource@nrc.gov>; RidsResDE Resource
<RidsResDE.Resource@nrc.gov>; RidsNrrPMWolfCreek Resource
<RidsNrrPMWolfCreek.Resource@nrc.gov>; RidsNrrDorllpl4-2 Resource <RidsNrrDorlLpl4-2.Resource@nrc.gov>; RidsManager Resource <RidsManager.Resource@ nrc.gov>; Regner, Lisa
<Lisa.Regner@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli @nrc.gov>; Lyon, Fred
<Fred.Lyon@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov>; !Burkhardt, Janet
<Janet.Burkhardt@nrc.gov>

Subject:

Relief Request Number 14R-03, Request for Relief from Paragraph-3200(b) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds and Relief Request 14R-04, Request for Relief from the Requirements of ASME Code Case N-729-1. ADAMS Distribution Notification A047 - OR Submittal: Inservice/Testing/Relief from ASME Code; related correspondence Open ADAMS PS Document(Relief Request Number 14R-03, Request for Relief from Paragraph-3200(b) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds and Relief Request 14R-04, Request for Relief from the Requirements of ASME Code Case N-729-1.) View ADAMS PS Properties ML16293A5S1 Accession ML16293A581 Number Relief Request Number l4R-03, Request for Relief from Paragraph-3200(b) of ASME Code Case N-729- 1 for Reactor Vessel Head Penetration Nozzle Welds Title and Relief Request 14R-04, Request for Relief from the Requirements of ASME Code Case N-729-1. Docket 05000482 Number Document 10/14/2016 Date Author McCoy JH Name

Author Wolf Creek Nuclear Operating Corp Affiliation Addressee Name Addressee NRC/Document Control Desk Affiliation NRC/NRR Document Letter Type Availabi lity Publicly Avai lable Date to be 10/2712016 Released Document Non-Sensitive Sensitiv ity Comment Date Added 10/ 19/2016 DPCautoadd Keyword gps l stt

From: Pascarelli, Robert Sent: 19 Oct 2016 07:43:46 -0400 To: Lingam, Siva

Subject:

RE: Comments on Relief Request 14R-03 Thanks Siva. Please set up a call sometime this afternoon. Dave Alley is in training all day but we should be OK if Jay Collins and some of the other reviewers can make it. From: Lingam, Siva Sent: Wednesday, October 19, 2016 7:00 AM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov> Cc: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

RE: Comments on Relief Request 14R-03 From: Taylor, Nick Sent: Tuesday, October 18, 2016 11:50 PM To: Collins, Jay <Jay.Collins@nrc.gov>; Lingam, Siva <Siva .Lingam@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Alley, David <David.Alley@nrc.gov>

Cc: Drake, James <James.Drake@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>

Subject:

Comments on Relief !Request 14R-03 Good evening everyone, I'm sorry this has taken so long for me to send out a note with my thoughts on the relief request. I was a busy day today onsite. I am at Wolf Creek, and actually went and stood on the head today, as well as spending a significant chunk of time talking with our ISi and RP inspectors, and spent about 1.5 hours talking with their VP of Engineering this afternoon (who signed the relief request). I've have a few thoughts to share, and would like to provide them to help inform your decision. Instead of putting them all in an email and possibly creating a lot of email buzz, I think it would be best to get on the phone sometime Wednesday to share my thoughts. It may be that granting relief is the appropriate action - I just want to be sure you all understand some of the things in the request for relief a full view of the actual conditions at the plant. Please let me know if there is a good time for a short call to discuss on Wednesday. My only "bad" times are between 0830-1100 central time. Thanks! Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV 0: (817) 200-1141 C : l(b)(6) I E: rnck.tayior@nrc.gov

I From : Collins, Jay Sent: Tuesday, October 18, 2016 6:18 AM To: Taylor, Nick <Nick.Taylor@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov> Cc: Tsao, John <John.Tsao@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Cumblidge, Stephen

<Stephen.Cumblidge@nrc.gov>

Subject:

RE: Call with Wolf Creek regarding head inspection Greetings, I am doing the 14R-03 relief and John Tsao is doing the 14R-04 relief. If you would like to have a call on the relief requests, we should be available after our branch meeting this morning ends at 10am our time, 9am Central. I am getting an automatic reply for you, so if you would like to do them by email, we could do that, as well. Stephen Cumblidge is making up a nice presentation about the volumetric leak path assessment, if you have questions on that item.

Thanks, Jay Collins NRR/DE/EPNB (301 )415-4038 Siva, we will be in 0-886 for our branch meeting from 9 to 10am.

From: Taylor, Nick Sent: 19 Oct 2016 15:47:50 -0500 To: Collins, Jay;Lingam, Siva;Tsao, John;Drake, James;Dodson, Douglas;Thomas, Fabian;Proulx, David Cc: Pascarelli, Robert;Alley, David;Cumblidge, Stephen

Subject:

RE: Wolf Creek Relief Requests 14R-03 and 14R-04 (CAC No. MF8456)

All, 1 just got out of a sit-down meeting with the Plant Manager (Steve Smith). According to Steve (and validated by the cu1Tcnt OCC schedule), the licensee plans to have their fina l vessel head cleaning complete by Thursday, October 241h. It 's not quite clear yet what "clean" means or how they will achieve the end result, but we've expressed our expectation that the final cleaning would allow licensee and NRC to sec the bare metal condition of the vessel head. We will follow this through the resident inspectors next week.

Thanks, Nick Taylor From: Collins, Jay Sent: Wednesday, October 19, 2016 2:37 PM To: Lingam, Siva <Siva.Lingam@nrc.gov>; Taylor, Nick < Nick.Taylor@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Drake, James <James.Drake@nrc.gov> ; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <David.Proulx@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@ nrc.gov>; Alley, David <David.Alley@nrc.gov>; Cumb lidge, Stephen <Stephen.Cumblidge@ nrc.gov>

Subject:

RE: Wolf Creek ReliefRequests l4R-03 and l4R-04 (CAC No. MF8456) << File: Volumetric Leakage Path.pptx >> Attached is Stephen Cumblidgc's slides explaining the volumetric leak path assessment. If you have any questions please let us know. Jay

---

Original Appointment-----

From: Lingam, Siva Sent: Wednesday, October 19 , 20 16 7:44 AM To: Lingam, Siva; Taylor, Nick; Collins, Jay; Tsao, John; Drake, James; Dodson, Douglas; Thomas, Fabian; Proulx, David Cc: Pascarelli, Robert; Alley, David S ubject: Wolf Creek Relief Requests 14R-03 and 14R-04 (CAC No. MF8456) When: Wednesday, October 19, 20 16 12:00 PM-1:00 PM (UTC-05:00) Eastern Time (US & Canada). W here: HQ-OWFN-10B06-12p Please note the following to discuss the subject RRs at the request of Nick Taylor: Bridge No.: 877-935-1422 Passcode: ~ followed by # Date: October 19, 2016 (Wednesday) Time: 12:00 PM (Eastern Time)

We are still waiting for the licensee's repo1t providing justification for not inspecting the nozzle penetrations other than 12 nozzles mentioned in the subject RRs. This is what I gathered from my BC who participated in the conference call held on October 17, 2016, at 5 :00 PM (Eastern).

From: Tsao, John Sent: 19 Oct 2016 11:46:04 +0000 To: Lingam, Siva

Subject:

Accepted: Wolf Creek Relief Requests 14R-03 and 14R-04 (CAC No. MF8456)

From: Alley, David Sent: 20 Oct 2016 14:16:00 +0000 To: Ross-Lee, MaryJane

Subject:

RE: Wolf Creek, Shearon Harris, Palo Verde Attachments: RE: Harris Vessel Head Flaw - Update to EN 52297 See attached emails Dave From: Ross-Lee, MaryJane Sent: Thursday, October 20, 2016 8:58 AM To: Alley, David <David.Alley@nrc.gov>

Subject:

RE: Wolf Creek, Shearon Harris, Palo Verde Where is the Shearon Harris indication? What location? Mary Jone Ross-Lee (MJ) Deputy Director, Division of Engineering Office of Nuclear Reactor Regulation OWFN 9H1 US Nuclear Regulatory Commission ~ Office: 301-415-3298 ~ e-mail: maryjane.ross-lee@nrc.gov From: Alley, David Sent: Wednesday, October 19, 2016 7:54 PM To: Ross-Lee, MaryJane <MarvJane.Ross-Lee@nrc.gov>

Subject:

Wolf Creek, Shearon Harris, Palo Verde MJ A bit tonight and then I will update in the morning. Wolf Creek They are planning to complete head cleaning (method yet undefined) by the 24th. We still have nothing from the license concerning nozzles other than the original 12. We have been expecting some sort of documentation for the last 2 days. Supposedly it will come tomorrow (Thursday). Licensee appears to be betting on relief request being granted as region says they are not making plans to do the surface exam. Biggest issue appears to be a fear on the part of the licensee of false calls if the surface exam is conducted. Regional inspector will look at head tomorrow to check the status of nozzles other than the 12 and look for evidence of whether the corrosion on the head and nozzle annuli is more than surface corrosion.

If more than surface corrosion observed, it could indicate that corrosion has been occurring for longer than the canopy seal leak. This might cause us to believe that there could be a leak through a J groove weld. Under those circumstances, we may not want to authorize tile proposed alternative. Shearon Harris Right now this is sounding more like a rounded indication on the weld (some disagreement concerning location). May be at a location of an indication at the time the weld was made. If rounded and especially if at same location as previous indication, it is probably a fabrication flaw which may be similar to those which are commonly found on encapsulation repairs (the Westinghouse approach to repair of cracked J groove welds). If that is true, grinding out the flaw would likely be appropriate. Most of the above is subject to verification. Our opinion could change significantly depending on confirmation or correction of the above No grinding is supposed to be done tonight. Palo Verde Nothing more than we reported this afternoon (9 inch circ indication on a 14 inch diameter line associated with safety injection tank). Materials and exact location of the indication are unknown. Dave David Alley PhD. Chief, Component Performance NDE and Testing Branch US Nuclear Regulatory Commission 11555 Rockville Pike Rockville MD 20852 301-415-2178

From: Galvin, Dennis Sent: 20 Oct 2016 09:43:19 -0400 To: Collins, Jay;Tsao, John;Alley, David Cc: Barillas, Martha;Butcavage, Alexander

Subject:

RE: Ha1rris Vessel Head Flaw - Update to EN 52297 Attachments: Drawing of flaw in w eld of previous repair

All, I have attached a diagram provided by the resident of the location of the flaw. Note that I have removed Jeanne Dion from the email list. I wanted her to see the first emails but she doesn't need the other ones ..

Note that the RI indicated that someone from the licensee this morning said that dye penetrant test was done incorrectly and that there was no flaw. However, two level 3 weld inspectors confirmed the flaw. The RI said Al Butcavage made the same point and agrees there is an indication of a flaw. The RI is following up. For now, I will hold off scheduling a call with the licensee until we hear something new from the licensee, the RI or from Al unless there is something you want to pass along.

Thanks, Dennis Galvin Project Manager U.S Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Licensing Project Branch 2-2 301-415-6256 From: Collins, Jay Sent: Thursday, October 20, 2016 8:36 AM To: Tsao, John <John.Tsao@nrc.gov>; Galvin, Dennis <Dennis.Galvin@nrc.gov>; Alley, David

<David.Alley@nrc.gov>

Cc: Barillas, Martha <Martha.Barillas@nrc.gov>; Dion, Jeanne <Jeanne.Dion@nrc.gov>; Butcavage, Alexander <Alexa nder.Butcavage@nrc.gov>

Subject:

Re: Harris Vessel Head Flaw - Updat e to EN 52297 Greetings, The indications that have been found on the Westinghouse embedded fl aw repaired nozzle surfaces have been attributed due to a combination of difficulties associated with the manual grinding activities, manual PTs and then the pressurization/thermal cycling. l believe with the half nozzle repair, other than the lack of manual grinding, we could have similar issues here. That is, unless there is a linear aspect to the indication, that could be a concern. That is why l

would just like to get that information, when available. I don't know that we need to have a call with the licensee, but just the final results of their inspection prior to flaw removal, that would be appreciated. Thoughts? Jay From: Tsao, John Sent: Thursday, October 20, 2016 8:26:29 AM To: Galvin, Dennis; Collins, Jay; Alley, David Cc: Barillas, Martha; Dion, Jeanne; Butcavage, Alexander

Subject:

RE: Harris Vessel Head Flaw - Update to EN 52297 Jay, as you can see in Dennis' email below that the flaw grew from 0.177 inches to .331 inches. This growth rate is aggressive. I suppose that the driving force is due to weld residual stresses, not the primary water stress corrosion cracking From: Galvin, Dennis Sent: Thursday, October 20, 2016 8:22 AM To: Collins, Jay <Jay.Collins@n rc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>

Cc: Barillas, Martha <Martha.Barillas@nrc.gov>; Dion, Jeanne <Jeanne.Dion@nrc.gov>

Subject:

RE: Harris Vessel Head Flaw - Update to EN 52297 Here is the update from the RI this morning: Indications of cracking have been identified on a previously repaired nozzle penetration (Nozzle 23). The crack is in the weld material that provides a structural seal between the bottom of the half-nozzle and the reactor vessel head . This was an existing flaw that was identified at the time of the repair, but was considered acceptable due the dimensions of the crack. The crack has since grown from .177 inches to 0.331 inches. Flaw is in the weld material not the nozzle. Licensee submitted a supplement to EN 52297 as another example of a degraded condition on the RVH. This is the first time that a previously repaired nozzle will need to undergo repair. Licensee is still developing a strategy for path forward on Nozzle 23 . From: Galvin, Dennis Sent: Thursday, October 20, 2016 8:20 AM To: Collins, Jay <Jay.Collins@n rc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>

Cc: Barillas, Martha <Martha.Barillas@nrc.gov>; Dion, Jeanne <Jeanne.Dion@nrc.gov>

Subject:

Harris Vessel Head Flaw - Update to EN 52297 Importance: High

All,

I just received this. I also talked to the licensee. There are no immediate plans for repairs and they are aware of our interest in getting the most data possible. The licensee is working on getting me some information Martha requested. If I hear anything I will update you.

Thanks, Dennis Galvin Project Manager U.S Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Licensing Project Branch 2-2 301-415-6256 From : Caves, John R [8]

Sent: Thursday, October 20, 2016 8:17 AM To: Galvin, Dennis <Dennis.Galvin@nrc.gov>

Subject:

[External_Send er] FW: Update to EN 52297 From: Nordby, Ingrid M Sent: Wednesday, October 19, 2016 6:06 PM To: HNP Lie Reg Prog; Hickerson, Lonnie; Waldrep, Benjamin C; Hamilton, Tanya M; Griffith, Donald L; Miller, D Bryan; O'Connor, Sean Thomas; Grantham, Mark A; Womack, Frankie L Cc: Zaremba, Arthur H.; Treadway, Ryan I; Green, Mary Kathryn; Sipe, Rita B; Jones-Young, Allison D; Volk, Samuel Joseph; Miller, Kris I; Fletcher II, Cecil Alexander; Grzeck, Lee; Nolan, Chris; Robertson, Jeffrey N; Wasik, Christopher J; Zimmerman, Tony

Subject:

Update to EN 52297 NRC was updated at 1756 EDT about Nozzle 23. Ingrid Nordby, P.E. Sr. Licensing Engineer, Regulatory Affairs Harris Nuclear Plant 919-362-2326 ingrid.nordby@duke-energy.com

From: Riches., Mark Sent: 20 Oct 2016 09:26:58 -0400 To: Rose, Steven Cc: Jackson, Donna;Dodson, Jim;Galvin, Dennis; Barillas, Martha

Subject:

Drawing of flaw in weld of previous repair Attachments: Locatio n of Flaw on Repaired Nozzle.pdf

Steve, The attached drawing indicates the area of concern for Nozzle 23. Let me know if you have any questions.

Thanks, Mark J. Riches US NRC I Resident Inspector Shearon Harris Station Phone: 9 19 362-0601 E-mail: Mark.Riches@nrc.gov

                \

112*MAX SEE NOTE 1

                                .70MIN FULL THICKNESS FROM TRIPLE POINT SEE NOTE 1 20*M1N SEE NOTE 1 SURFACE UITABLE PT DETAIL B STEP4 AO.~fl o ~ ~\-~w

!N\\.\~ \'l~\..1' WPS WP3143/F43TBSC3

                                            --~                      ~Jio."'\..

( 1.e.. No~~i..~'22>') I FILLER: ERNICrFe-7A AFTER MACHINING PER STEPS

                                              =t-                                                t  -- t - -
                                                 --t---   WHEN REQUIRED, HOUSING EXTEN WELDED TO THE* ORIGINAL HOUSIN WELD FOR IOTB WELDING AND Ra AFTER MACHINING                                        AND IS REMOVED AFTER REMEDIA1 PERSTEP5                  UT                           SEENOTE5 STEP4                  NOTES:

WELDING

From: Alley, David Sent: 20 Oct 2016 22:07:10 -0400 To: Collins, Jay;Lingam, Siva;Drake, James.;Taylor, Nick;Anchondo, lsaac;Dodson, Douglas;Kopriva, Ron Cc: Pascarelli, Robert;Tsao, John; Cumblidge, Stephen;Singal, Balwant

Subject:

RE: Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456)

All, I read through the licensee's response to us and the code case info again. Now I don't think I was quite right in my email below. Not quite sure how I was looking at this this PM but it isn't holding water tonight.

Based on what Jay sent from the code case: 3140 INSERVICE VISUAL EXAMINATIONS (VE) -3141 General (c) Relevant conditions for the purposes of the VE shall include areas of corrosion, boric acid deposits, discoloration, and other evidence of nozzle leakage. This defines relevant conditions. It does not define location on the head or nozzles, i.e., a relevant condition can be at the annulus or between nozzles on the head. The term relevant condition is divided into two categories 3142.2 Acceptance by Supplemental Examination. A nozzle with relevant conditions indicative of possible nozzle leakage .. . And 3142.3 Acceptance by Corrective Measures or Repair/Replacement Activity (a) A component with relevant conditions not indicative of possible nozzle leakage Neither of the concepts, indicative or not indicative of possible nozzle leakage, appear to br defined in the code case. Not indicative seems pretty easy, i.e., boric acid or other things that could indicate leakage which are not connected to a nozzle annulus. Indicative of leakage could be a bit harder. As Jay points out: The NRG considers any relevant condition in the annulus region between the nozzle and head surface that cannot be removed by light cleaning activities to be a relevant condition of possible nozzle leakage. (Jay where is this written down?) Despite the logic and history of the above position, I think we need to recognize that this definition is not an explicit part of the code case. Having said this, I do not see how an alternate conclusion can be reached. The very small leakage from cracks in J groove welds early in a leakage event cannot be expected to generate significant amounts of boric acid residue. At the same time, boric acid debris that falls or is blown onto the head cannot be expected to be adherent with respect to Jay's "light cleaning activities". Based on all the above, statements of significance in the licensee's submittal appear to be Penetrations with relevant conditions identified

All penetrations (referring to the previous sentence with identified 59, 77, 71, 46, 70, 58 and 63) were assessed by the QC level Ill examiners as having no boron in the annulus area. {This statement is inconsistent with the evidence so far presented to the NRC) and Reactor vessel head insulation The remaining nozzles were also carefully reviewed both in person and by video footage. The nozzles with residue buildup were carefully examined to the point that WCNOC is confident the residue was not originating from a crack in the alloy 600 material or the partial penetration weld on each nozzle. And Examination of vessel closure head visual examination results The logic used in evaluating the penetrations with relevant conditions was the ability to determine visually that the accumulation could not have come from the partial penetration weld or a nozzle crack. This appears to indicate that other nozzles had boric acid and/or corrosion products touching the annulus (we need to confirm but this is consistent with evidence presented to us thus far). Based on other statements, they were not successful in vacuuming up much/any debris that may have been present. This is where we get to Jay's precedent statement. If they have boric acid in the annulus and they don't get it up by vacuuming, it doesn't seem possible for them to reach a conclusion that it didn't come from the J groove weld. This appears to be the point that may need to be discussed At the moment it appears that there are three questions to answer. Is there boric acid or corrosion products in contact with the annulus on nozzles other than the original 12? If so, is there a basis by which wolf creek can reach a conclusion that those debris are not indicative of leakage? And the third question is for us, is there a way for us to accept wolf creek's position without reviewing the photos for each nozzle upon which wolf creek based their decisions? Dave From: Alley, David Sent: Thursday, October 20, 2016 3:36 PM To: Collins, Jay <Jay.Collins@nrc.gov>; Lingam, Siva <Siva .Lingam@ nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Cumblidge, Stephen <St ephen.Cumblidge@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

RE: Wolf creek* WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Paragraph 3 of section "Evaluation of Vessel Closure Head Visual Examination Results" says "There were relevant conditions in close proximity to many nozzles". Based on definitions in the code case this puts all those nozzles in need of inspection. Given that they say "in close proximity" indicates to me that they may not understand "relevant condition" as if there is a gap

between the annulus and the "problem" in my mind the problem may not be evidence of leakage and, therefore, not relevant. Based on what they have said, I agree with Jay that at least an internal call is needed. Dave From : Collins, Jay Sent: Thursday, October 20, 2016 3:18 PM To: Lingam, Siva <Siva .Lingam@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Cumblidge, Stephen <Stephen .Cumblidge@nrc.gov>; Singal, Balwant
<Balwant.Singal@nrc.gov>

Subject:

RE: Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. M F8456) Greetings, In my opinion, this is completely inadequate to address the qu estion. If the annuluses of these nozzles are not clear of boric acid or corrosion product, regardless of how the inspector t hought it got there, t hen t he nozzle has a relevant condition of possible nozzle leakage. Per N-729-1, supplement al examinations are required uncler -3200(b). I request at least an internal phone call this afternoon. Jay -3 142.2 Acceptance by Supplemental E xamination. A nozzle with relevant conditions indicative of possible nozzle leakage sha ll be acceptable for continued service if the results of supplemental examinations [-3200(b)] meet the requirements of -3 130. -3.141 General (c) Relevant conditions for the purposes of the VE shall include areas of corrosion, boric acid deposits, discoloration, and other evidence of nozzle leakage. From: Lingam, Siva Sent: Thursday, October 20, 2016 2:55 PM To: Collins, Jay <Jay.Collins@n rc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Cumblidge, Stephen <St ephen .Cumblidge@nrc.gov>; Proulx, David
<David. Prou lx@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Dodson, Douglas
<Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Kopriva, Ron
<Ron.Kopriva@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Anchondo, Isaac
<lsaac.Anchondo@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Attached please find the licensee's response for nozzles with boric acid (other than 12 nozzles) for your review/evaluation for RR 14R-03 ..

From : Stone Lucille M [mailto:lurocke@WCNOC.com) Sent: Thursday, October 20, 2016 2:33 PM To: Lingam, Siva <Siva .Lingam@nrc.gov>

Subject:

[External_ Send er] FW: WCNOC response to verbal RAI for relief request 14R-03 From: Stone Lucille M Sent: Thursday, October 20, 2016 1:27 PM To: 'balwant.singal@nrc.gov'; 'nick.taylor@nrc.gov'; 'ron.kopriva@nrc.gov'

Subject:

WCNOC response to verbal RAI for relief request 14R-03

All, Here is electronic copy. Hard copies in the mail.

Lu Stone WCNOC Licensing

From: Ch eruvenki, Ganesh Sent: 20 Oct 2016 15:02:14 -0400 To: Medoff, James;Coll ins, Jay;Min, Seung K;Hiser, Allen Subje ct: BM l--SLR; Wolf Creek upper head leakage

From: Drake, James Sent: 20 Oct 2016 13:51:32 -0500 To: Alley, David Cc: Collins, Jay;Tsao, John;Cumblidge, Stephen;Hoffman, Keith

Subject:

FW: Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion FYI Jim From : Proulx, David Sent: Thursday, October 20, 2016 1:42 PM To: Kennedy, Kriss <Kriss.Kennedy@nrc.gov>; Morris, Scott <Scott.Morris@nrc.gov>; Pruett, Troy

<Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>;

Werner, Greg <Greg.Werner@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Dricks, Victoir

<Victor.Dricks@nrc.gov>; Maier, Bill <Bill.Maier@nrc.gov>; Moreno, Angel <Angel.Moreno@nrc.gov>;

Bowen, Jeremy <Jeremy.Bowen@nrc.gov>; Lyon, Fred <Fred.Lyon@nrc.gov>; Singal, Balwant

<Balwant.Singal@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Taylor, Nick
<Nick.Taylor@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Dodson, Douglas
<Douglas.Dodson@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>

Subject:

Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Good afternoon, Please see the attached revised one-pager related to the Wolf Creek reactor vessel head leakage and corrosion. Feel free to forward to all interested parties. We will continue to update the information as the licensee completes their cleaning and inspection activities. If you have any additional questions, please contact me at 817-200-1 561. Very respectfully, Davia:fi.oulx SPE- RPBB x1561

From: lingam, Siva Sent: 20 Oct 2016 15:04:48 -0400 To: Collins, Jay;Tsao, John Cc: Alley, David;Cumblidge, Stephen;Pascarelli, Robert

Subject:

FW: Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Attachments: Wolf Creek Vessel Head Nozzle Leakage 10-20-16-rev.docx Attached please find the one-pager from Region IV (as requested by Bill Dean). Please review and provide comments, if any. Thank you. From: Pascarelli, Robert Sent: Thursday, October 20, 2016 2:45 PM To: lingam, Siva <Siva.Lingam@nrc.gov>

Subject:

FW: Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion FYI From: Proulx, David Sent: Thursday, October 20, 2016 2:42 PM To: Kennedy, Kriss <Kriss.Kennedy@nrc.gov>; Morris, Scott <Scott.Morris@nrc.gov>; Pruett, Troy

<Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>;

Werner, Greg <Greg.Werner@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Dricks, Victoir

<Victor.Dricks@nrc.gov>; Maier, Bill <Bill.M aier@nrc.gov>; Moreno, Angel <Angel. Moreno@n rc.gov>;

Bowen, Jeremy <Jeremy.Bowen@nrc.gov>; Lyon, Fred <Fred.Lyon@nrc.gov>; Singal, Balwant

<Balwant.Singal@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Taylor, Nick
<Nick.Taylor@nrc.gov>; Thomas, Fabian <Fabian .Thomas@nrc.gov>; Dodson, Douglas
<Douglas.Dodson@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>

Subject:

Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Good afternoon, Please see the attached revised one-pager related to the Wolf Creek reactor vessel head leakage and corrosion. Feel free to forward to all interested parties. We will continue to update the information as the licensee completes their cleaning and inspection activities. If you have any additional questions, please contact me at 817-200-1561. Very respectfully, 1Javicff/5,.oufx SPE-RPBB x1561

Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Key Messages NOTE: Significant updates annotated by date entered.

   *   (10/20/2016) Continuing reactor vessel head inspections have not identified any significant damage to the head itself, although additional cleaning activities must be completed to observe the bare metal condition. The licensee's current schedule shows them completing this cleaning and doing additional visual inspection around 10/24. The licensee is also removing other components from the reactor vessel head assembly (such as CROM coil stacks) to remove accumulated boron deposits that were carried through the head plenum by ventilation flow.
   *   (10/20/2016) The licensee has completed volumetric examinations of twelve penetration nozzles in the spray area of the leak, which appear to have satisfactory results (the examinations did not reveal any leak paths from inside the vessel). The licensee has requested relief from code-required surface examinations of the penetration welds on the bottom of the head for the twelve affected penetrations. Headquarters is still reviewing the relief request, pending receipt of additional information from the licensee and ongoing inspection efforts by Region IV ISi inspectors who are in the field again this week.
   *   (10/20/2016) The licensee has installed an approved canopy seal clamp assembly (CSCA) on penetration 77, which was the source of the leak requiring the early shutdown of the plant.

Additionally, the licensee has installed CSCAs on two other nozzles which were susceptible to future leakage, and is in the progress of installing two more. The licensee has described their plans to evaluate installing CSCAs on all such nozzles in a future outage to mitigate the risk of future leaks above the head.

• Wolf Creek completed a technical specification (TS) required shutdown of the reactor on Friday, September 2, 2016, in order to locate and repair the source of elevated reactor coolant system leakage. The source of the leak was determined to be a leaking canopy seal weld on a core exit thermocouple penetration nozzle above the reactor vessel head (penetration 77).
• Upon initial inspection on September 19, indication of carbon steel corrosion was noted on the reactor vessel head. The corrosion appears to be limited to a small sector of the reactor vessel head and surrounding structures below the leaking penetration.
• Following the shut down the licensee began a planned refueling outage. The licensee moved the reactor vessel head to the inspection stand, where continuing inspection and repairs to the head are being completed.

Facts

• The resident inspectors monitored reactor coolant system leakage throughout the operating cycle .

Data indicated a steady very small leak rate (approximately 0.05 gallons per minute), that suddenly began to increase on August 31, 2016. On September 2, 2016, Wolf Creek observed RCS unidentified leakage in excess of 1.35 gallons per minute (gpm), exceeding the TS limit of 1.0 gpm. As a result, the licensee initiated a TS required shutdown on September 2, 2016.

Contact:

Nick Taylor, Chief, Reactor Projects Branch B October 20, 2016 (817) 200-1141

• Following shutdown and containment entry, the source of the leak was identified as the canopy seal weld on penetration 77 above the reactor vessel head, which serves one of the core exit thermocouples. Leakage through the threaded mechanical joint serving the core exit thermocouple nozzle assembly is not considered pressure boundary leakage.
• Following the shutdown, the licensee decided to commence their refueling outage, which is planned for 55 days.
• The reactor vessel head is the original head and is approximately 30 years old. The licensee has periodically inspected the head for leakage in accordance with their approved in-service inspection program. The last such inspection was in the spring 2015 refueling outage.
• A Region IV Division of Reactor Safety inspector is currently onsite to assist the resident inspectors in the follow up of these issues.

Contact:

Nick Taylor, Chief, Reactor Projects Branch B October 20, 2016 (817) 200-1141

From: Clark. Theresa To: Alley David; Rud!and David; Chernoff Harold Cc: Bowen I eremy

Subject:

NDE OpE topics - potential 11/2 agenda item Date: Friday, October 21, 2016 8:29:24 AM Dave and Dave-I know your groups are followi ng several materials/ISi issues at t he sites. We haven't been discussing the relat ed event reports at our OEDO morning meeti ngs in much detai l as they come in, but there was a request today to discuss t hem in general at the next quarterly OpE briefing if there is a trend or anything interesting to say from a collective perspective. Ones we've seen recently incl ude Wolf Creek nozzles, Harris nozzles, Palo Verde SI piping. Agai n, no cu rrent action, just potential topic for t he next meeting (November 2 at lOam ) if you thi nk it's appropriate. Thanks! Theresa Valentine Clark Executive Technical Assistant (Reactors) U.S. Nuclea r Regulatory Commission Theresa.Cla rk@orc.gov I 301-415-4048 I 0 -4H10

From: Collins, Jay Sent: 21 Oct 2016 12:36:45 +0000 To: Alley, David;Lingam, Siva; Drake, James;Taylor, Nick;Anchondo, Isaac; Dodson, Douglas;Kopriva, Ron Cc: Pascarelli, Robert;Tsao, John; Cumblidge, Stephen;Singal, Balwant

Subject:

RE: Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Greetings, Below is a dated EPRI reference from 2003, to show this is not a new process of determining if a relevant condition is indicative of possible nozzle leakage. Please note this is EPRI licensed material and not for general disclosure. Visual Examination for Leakage of PWR Reactor H ead Penetrations Revision 2 of I 006296, Include 2002 Inspection Results and MRP Inspection Guidance 6.2.6 Bone acid leakage from above may leak through the insulation and collect on the underside of the insulation. or run down the penetrations as shown in Figure 3-19. This condition can greaUy affect the ability to characterize whether the leakage is occlUring from the annulus of the penetration or from another source. A deposit of this type can take several forms. It can tend to cover portions of the head like a tight adhering coating (Figure 3-20) or take the form of loose granular material that may rest on the uphill side of the penetration (Figure 3-21). 6.2.7 Compressed air, in the range of 40-t>O psi (27~ 14 kPa), or a vacuum directed at deposits has been used to distinguish whether a deposit is loose buildup of material simply resting against a penetration that is easily removed or is a tightly adhering deposit. originating from the annulus of a leaking penetration. Figures 3-22 and 3-23 show *1>efore and after" examples ofa penetration evaluated in this fashion. Figure 3-23 shows Penetration #63 after air. at 6Qi psi (414 kPa), was blown at the boron deposit. Because the leakage was from the atumlus. the force of the air blast did not remove the boron deposit. If the deposirs bad faJlen from anorher leak point. they would have ~own away, leaving a clean appearing penerration. After proper documentation. it is important to remove these deposits before the next inspection.

Figur*l-22 Suspect t..<lking PttWtr.ltion S.fore Blowing High-Pres54A'P Air at Deposits (See 6.27., Sedion 2 .) FigurPl-23 S..- PttWtr.Jtion Aft<< Blowing High-"ress\n Air at O.posits (Shows That the O.posit Was ~used by an Actual L..ak ~~ It Woukl Not Blow Away) (Sff 6.27 in Sedion 2.) From : Alley, David Sent: Thursday, October 20, 2016 10:07 PM To: Collins, Jay ; Lingam, Siva; Drake, James ; Taylor, Nick ; Anchondo, Isaac; Dodson, Douglas; Kopriva, Ron Cc: Pascarelli, Robert; Tsao, John ; Cumblidge, Stephen ; Singal, Balwant

Subject:

RE: Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456)

All, I read through the licenseea's response to us and the code case info again. Now I dona't think I was quite right in my email below. Not quite sure how I was looking at this this PM but it isna't holding water tonight.

Based on what Jay sent from the code case: 3140 INSERVICE VISUAL EXAMINATIONS (VE) -3141 General (c) Relevant conditions for the purposes of the VE shall include areas of corrosion, boric acid deposits, discoloration, and other evidence of nozzle leakage. This defines relevant conditions. It does not define location on the head or nozzles, i.e., a relevant condition can be at the annulus or between nozzles on the head. The term relevant condition is divided into two categories 3142.2 Acceptance by Supplemental Examination.

A nozzle with relevant conditions indicative of possible nozzle leakagea: And 3142.3 Acceptance by Corrective Measures or Repair/Replacement Activity (a) A component with relevant conditions not indicative of possible nozzle leakage Neither of the concepts, indicative or not indicative of possible nozzle leakage, appear to br defined in the code case. Not indicative seems pretty easy, i.e., boric acid or other things that could indicate leakage which are not connected to a nozzle annulus. Indicative of leakage could be a bit harder. As Jay points out: The NRC considers any relevant condition in the annulus region between the nozzle and head surface that cannot be removed by light cleaning activities to be a relevant condition of possible nozzle leakage. (Jay where is this written down?) Despite the logic and history of the above position, I think we need to recognize that this definition is not an explicit part of the code case. Having said this, I do not see how an alternate conclusion can be reached. The very small leakage from cracks in J groove welds early in a leakage event cannot be expected to generate significant amounts of boric acid residue. At the same time, boric acid debris that falls or is blown onto the head cannot be expected to be adherent with respect to Jaya's acelight cleaning activitiesa J. Based on all the above, statements of significance in the licenseea's submittal appear to be Penetrations with relevant conditions identified All penetrations (referring to the previous sentence with identified 59, 77, 71, 46, 70, 58 and 63) were assessed by the QC level Ill examiners as having no boron in the annulus area. (This statement is inconsistent with the evidence so far presented to the NRC) and Reactor vessel head insulation The remaining nozzles were also carefully reviewed both in person and by video footage. The nozzles with residue buildup were carefully examined to the point that WCNOC is confident the residue was not originating from a crack in the alloy 600 material or the partial penetration weld on each nozzle. And Examination of vessel closure head visual examination results The logic used in evaluating the penetrations with relevant conditions was the ability to determine visually that the accumulation could not have come from the partial penetration weld or a nozzle crack. This appears to indicate that other nozzles had boric acid and/or corrosion products touching the annulus (we need to confirm but this is consistent with evidence presented to us thus far). Based on other statements, they were not successful in vacuuming up much/any debris that may have been present. This is where we get to Jaya' s precedent statement. If they have boric acid in the annulus and they dona't get it up by vacuuming, it doesna't seem possible for them to reach a conclusion that it didna't come from the J groove weld. This appears to be the point that may need to be discussed At the moment it appears that there are three questions to answer. Is there boric acid or corrosion products in contact with the annulus on nozzles other than the original 12? If so, is there a basis by which wolf creek can reach a conclusion that those debris are not indicative of leakage? And the third question is for us, is there a way for us to accept wolf creeka's position without reviewing the photos for each nozzle upon which wolf creek based their decisions? Dave From: Alley, David Sent: Thursday, October 20, 2016 3:36 PM

To: Collins, Jay <Jay.Collins@nrc.gov>; Lingam, Siva <Siva.Lingam@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Cumblidge, Stephen <Stephen .Cumblidge@nrc.gov>; Singal, Balwant <Ba lwant.Singal@nrc.gov>

Subject:

RE: Wolf creek* WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Paragraph 3 of section areEvaluation of Vessel Closure Head Visual Examination Resultsa1I says areThere were relevant conditions in close proximity to many nozzlesarl. Based on definitions in the code case this puts all those nozzles in need of inspection. Given that they say arein close proximityaCJ indicates to me that they may not understand arerelevant conditiona 1 as ifthere is a gap between the annulus and the areproblemaCJ in my mind the problem may not be evidence of leakage and, therefore, not relevant. Based on what they have said, I agree with Jay that at least an internal call is needed. Dave From : Collins, Jay Sent: Thursday, October 20, 2016 3:18 PM To: Lingam, Siva <Siva.Lingam@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Cumblidge, Stephen <Stephen .Cumblidge@nrc.gov>; Singal, Balwant
<Balwant.Singal@nrc.gov>

Subject:

RE: Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Greetings, In my opinion, this is completely inadequate to address the question . If the annuluses of these nozzles are not clear of boric acid or corrosion product, regardless of how the inspector thought it got there, t hen t he nozzle has a relevant condition of possible nozzle leakage. Per N-729-1, supplemental examinations are required under -3200(b). I request at least an internal phone call this afternoon.MsoNormal"> Jay -3142.2 Acceptance by Supplemental Examination. A nozzle with relevant conditions indicative o f possible nozzle leakage shall be acceptable for continued service if the results of supplemental examinations (-3200(b)] meet the requirements of -3130. -3141 General (c) Relevant conditions for the purposes of the VE shall include areas of corrosion, boric ac id deposits, discoloration, and other evidence of nozzle leakage. From : Lingam, Siva Sent: Thursday, October 20, 2016 2:55 PM To: Collins, Jay <Jay.Collins@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Tsao, John

<John.Tsao@nrc.gov>; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>; Proulx, David
<David.Proulx@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Dodson, Douglas
<Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Kopriva, Ron
<Ron.Kopriva@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Anchondo, Isaac
<lsaac.Anchondo@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

Wolf creek - WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Attached please find the licenseea TM s response for nozzles with boric acid (other than 12 nozzles) for your review/evaluation for RR 14R-03.. From : Stone Lucille M [9] Sent: Thursday, October 20, 2016 2:33 PM

To: Lingam, Siva <Siva .Lingam@nrc.gov>

Subject:

[External_Sender] FW: WCNOC response to verbal RAI for relief request 14R-03 From: Stone Lucille M Sent: Thursday, October 20, 2016 1:27 PM To: 'balwant.singal@nrc.gov'; 'nick.taylor@nrc.gov'; 'ron.kopriva@nrc.gov'

Subject:

WCNOC response to verbal RAI for relief request 14R-03

All, Here is electronic copy. Hard copies in the mail.

Lu Stone WCNOC Licensing

From: Collins, Jay Sent: 21 Oct 2016 11:27:56 +0000 To: Lingam, Siva

Subject:

Accepted: Wolf Creek Relief Request 14R-03 (CAC No. MF8456)

From: lingam, Siva Sent: 21 Oct 2016 07:38:02 -0400 To: Collins, Jay Cc: Pascarelli, Robert;Alley, David;Tsao, John;Cumblidge, Stephen

Subject:

RE: Internal call with NRR concerning Wolf Creek Head inspection relief request FYI From: Drake, James Sent: Thursday, October 20, 2016 6:14 PM To: Clark, Jeff; Vegel, Anton ; Lantz, Ryan ; Pruett, Troy Cc: Werner, Greg; Anchondo, Isaac; Taylor, Nick; Dodson, Douglas; Thomas, Fabian; Proulx, David; Kopriva, Ron; Lingam, Siva

Subject:

Internal call with NRR concerning Wolf Creek Head inspection relief request We had a call with Dave Alleya's group to discuss the licenseea's recent submittal to support the relief requests they have pending for Relief from the Requirements of ASME Code Case N-729-1 . Based on the wording used in the submittal, there is concern that the licensee may not have properly followed the requirements of the Code case. Per the code, 3140 INSERVICE VISUAL EXAMINATIONS (VE)-3141 General (c) Relevant conditions for the purposes of the VE (visual examination) shall include areas of corrosion, boric acid deposits, discoloration, and other evidence of nozzle leakage. The NRC considers any relevant condition in the annulus region between the nozzle and head surface that cannot be removed by light cleaning activities to be a relevant condition of possible nozzle leakage. The code states in part, ace(c) A nozzle whose VE indic.ates relevant conditions indicative of possible nozzle leakage shall be unacceptable for continued service unless it meets the requirements of-3142.2 or -3142.3. -3142 .2 Acceptance by Supplemental Examination. A nozzle with relevant conditions indicative of possible nozzle leakage shall be acceptable for continued service if the results of supplemental examinations [-3200(b)] meet the requirements of-3130. 3130 is the lnservice Volumetric And Surface Examinations. In the submittal the licensee uses the term acerelevant condition,aD then goes on the state in part, aceRough cleaning was performed using a vacuum cleaner. The suction created by the vacuum cleaner was minimal and incapable of removing particulate from surfacesa:aD The licensee made the following the statements; aceThe logic used in evaluating the penetrations with relevant conditions was the ability to determine visually that the accumulation could not have come from the partial penetration weld or a nozzle crack.aD aceThere were relevant conditions in close proximity to many nozzles as well as a large percentage of the vessel head surface not included in the examination areas adjacent to the nozzles. These encompassed various forms of relevant conditions, but none were/are indicative of pressure boundary leakage from the vessel closure head.aD Based on this information, we are setting up a call with the licensee for 0930 central, 1030 eastern time to discuss the process they used in applying the code case and determine if they have just used the terms incorrectly but did properly apply the code case and disposition the indications correctly. Understanding this is necessary before considering approval of the relief request. Siva Lingam is arranging the logistics for the conference call. Jim

$ mie.r 'f'. :Drake James F. Drake Office phone: 817-200-1558 Cell Phone: l(b)(6) I

From: Lingam, Siva Sent: 21 Oct 2016 07 :26:40 -0400 To: Collins, Jay;Tsao, John;Alley, David;Cumblidge, Stephen;Taylor, Nick;Drake, James;Dodson, Douglas;Thomas, Fabian;Proulx, David;Kopriva, Ron;Anchondo, lsaac;Pick, Greg;Kalikian, Roger;wimulie@WCNOC.com;cyhafen@wcnoc.com;jaknust@WCNOC.com Cc: Pascarelli, Robert;Werner, Greg

Subject:

Wolf Creek Relief Request 14R-03 (CAC No. MF8456) Attachments: ET16-0028.pdf Please note the following to discuss the subject RR with the licensee based on the attached response from the licensee: Bridge No.: 877-935-1422 Passcodc: ~ followed by# Date: October 2 1, 20 16 (Friday) Time: I0:30 AM (Eastern Time)

From: Alley, David Sent: 14 Oct 2016 19:42:14 +0000 To: Collins, Jay

Subject:

RE: RE: WCNOC RV pictures Thanks I got into the system but haven't gotten any farther yet. From: Collins, Jay Sent: Friday, October 14, 2016 3:40 PM To: Alley, David <David.Alley@nrc.gov>

Subject:

Fw: RE: WCNOC RV pictures Greetings, It looks like you will get access. The numbering is a bit confusing. Once you get connected, in the first folder is a list of pictures. The file titled, Pen 67 & 54 DSC00068, seems to include a picture of penetration nozzles 67 and 54. They also appear to be labeled in the picture. I believe this is a view t hat Isaac gave us previously, but not the same photo. Note that the vent line in the picture between the two penetration nozzles. Now, if you go to the head map image listed as M-706-00009_REACTOR PEN in the folder, you will find that penetration nozzles numbers 67 and 54 are no where near nozzle 77, the source of the spill. Instead, number 67 is at approximately 320 degrees and nozzle 54 is at 290 degrees, near the periphery, in the south-west quadrant of the head. Note also, they are not near the head vent line, which is at about the 45 degree location in the North-west quadrant of the head. I believe nozzle 67 is the nozzle 76 that Isaac circled with a question mark in the images he sent on Thursday. Either way, 67 or 76, it has remain ing indications in the annu lus region and is not listed for volumetric inspecti on. I will look over the photos this weekend, but i think we will perhaps need an internal discussion on Monday for a bit. Jay From: Good Nicole R <nilyon@WCNOC.com> Sent: Friday, October 14, 2016 10:25 AM To: Singal, Balwant; Ungam, Siva Cc: Lingam, Siva; Collins, Jay; Tsao, John; Alley, David; Pascarelli, Robert

Subject:

[External_Sender) RE: WCNOC RV pictures Access has been provided to: Siva Lingam Jay Collins J ohn Tsao

Access has been r equested for: Balwant Singal David Alley Robert Pascarelli Thank you, Nicole Good From: Singal, Balwant [10] Sent: Thursday, October 13, 20163:12 PM To: Good Nicole R; Lingam, Siva Cc: Lingam, Siva; Collins, Jay; Tsao, John; Alley, David; Pascarelli, Robert

Subject:

RE: WCNOC RV pictures

Nicole, Not clear if we already have the access or will be getting access later. Please have access to following two persons as a minimum:

Siva Lingam Jay Collins Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwant.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222

   .~.

From: Good Nicole R [mailto:nilyon@WCNOC.com) Sent: Thursday, October 13, 2016 4:05 PM To: Lingam, Siva <Siva.Lingam@nrc.gov> Cc: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

[External_Sender] WCNOC RV pictures I was told you would like pictures of the penetrations with labels of the penetration number. I have only been able to locate a few pictures, at this point. I have granted you access to the Certrec IMS Sept 2016 Forced Outage. Item #14 has five pictures that may be hel pful (DCS00006, DCS00039, DCS00029, DCS00019, and DCS00018). I will need to contact Certrec to get access for Mr. Singal. I will work on getting Mr. Signal access and looking for more pictures tomorrow.

Thank you, Nicole Good Licensing nilyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek Nuclear Operating Corporation

From: Singal, Balwant Sent: 24 Oct 2016 10:22:25 -0400 To: Tsao, John;Alley, David;Collins, Jay;Ka likian, Roger Cc: Pascarelli, Robert;Lingam, Siva

Subject:

FW: EB2 Acting BC October 24 - 28 I just spoke with the licensee. They are in the process of putting the information together. They will be calling Jim Drake initially and will request a call with headquarters after initial discussions with Region IV. Thanks. Latest I have heard so far. I w ill be calling the licensee to check on the status as well. Thanks. Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwa nt.Si nga l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 From : Proulx, David Sent: Monday, October 24, 2016 9:57 AM To: Thomas, Fabian ; Singal, Balwant Subje ct: FW: EB2 Acting BC October 24 - 28 From : Werner, Greg Se nt: Monday, October 24, 2016 7:20 AM To: Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; M at eychick, John

<John.Mateychick@nrc.gov>

Cc: Drake, James <James.Drake@nrc.gov> Subje ct: RE: EB2 Acting BC October 24 - 28 Just got through talking with Jim Drake, seems pretty clear what HQs is telling WC. If WC has relevant indications on the head, besides the 12 nozzles, WC has to do the supplemental inspections. WC cana't just do visual from the top and make a claim that they believe it came the spray or some other area. There has to be some quantifiable examination, such as the volumetric done on the 12 nozzles and surface exam on the underside surface of the head. From : Taylor, Nick Se nt: Monday, October 24, 2016 7:05 AM To: Proulx, David <David.Prou lx@nrc.gov>; M ateychick, John <John.Mateychick@nrc.gov> Cc: Werner, Greg <Greg.Werner@nrc.gov>; Drake, James <James.Drake@nrc.gov>

Subject:

FW: EB2 Acting BC October 24 - 28 David I John, I recommend the two of you get synched up quickly this week on the status of the Wollf Creek relief request & request for additional info from headquarters and EB2. The Wolf Creek managers are not happya:

Thanks,

Nick From: Werner, Greg Sent: Monday, October 24, 2016 6:58 AM To: R4DRS-EB2 <R4DRS-EB21@nrc.gov>; Hay, M ichael <Michael.Hay@nrc.gov>; R4ACES

<R4ACES@nrc.gov>; R4DRS-BC <R4DRS-BC1@nrc.gov>; R4DRP-BCandSPE <G-R4-DRP-BCandSPE@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>; Pruett, Troy <Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan .Lantz@nrc.gov>

Subject:

EB2 Acting BC October 24 - 28 John Mateychick will be the acting EB2 branch chief from today, October 24, thru Friday, October 28. I will becheckjnq eroails and my cell voicemails periodically throughout the week. My cell number i~(5}(6 ) I Greg Werner

From: Collins, Jay Sent: 25 Oct 2016 14:32:39 +0000 To: Cumblidge, Stephen;Singal, Balwant Cc: Kalikian, Roger

Subject:

RE: Important details for Wolf Creek Attachments: Wolf Creek verbal auth 14R-03 10-17-2016 Rev 3.docx Attached is the script that Dave likes. I think the question will be, do they need to do additional examinations and therefore expand the number of nozzles that the relief covers. Balwant, I went looking for you on the ath floor, are you working at home today? Jay From: Cumblidge, Stephen Sent: Tuesday, October 25, 2016 10:20 AM To: Collins, Jay

Subject:

Important details for Wolf Creek What are the main issues that you would like covered for the Wolf creek relief? I have no problem with using a leak path assessment, but I will have a hard time saying yes to a relief that says that they are ignoring relevant indications. Also, please send over your drat script. Stephen Cumblidge Materials Engineer US Nuclear Regulatory Commission Mail Stop OWFN/9 H6 Washington, DC 20555-0001 Telephone: (301) 415-2823 (Office)

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELI EF REQUEST 14R-03 ALTERNATIVE TO USE VOLUMETRIC LEAK PATH FOR SUPPLEMENTAL EXAMS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated October 14, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-03 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration weld numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 at the Wolf Creek Generating Station. The licensee proposed (a) to perform a volumetric leak path assessment of each penetration nozzle in lieu of the surface leak path assessment required by Paragraph - 3200(b) of ASME Code Case N-729-1, and (b) if an unacceptable indication by the leak path assessment or volumetric exam is identified , the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that while the demonstrated volumetric leak path is not equiva lent to a fully qualified surface leak path assessment, the licensee identified sufficient operational experience, technical basis and radiological dose hardship to show that regulatory compliance would result in hardship without a compensating increase in the level of quality and safety. For operating experience, the licensee showed that there has been no previous identified cracking or leakage identified from the CRDM nozzle penetrations or welds of the upper head at Wolf Creek. The NRC staff noted that while this fact does not preclude the possibility of cracking to be found as the plant continues to age, plants which have previously identified cracking are more likely to see subsequent and more significant cracking in the future . Given the lack of the initial cracking being identified in the nozzle heats of material, at the operating temperatures of Wolf Creek, the NRC found that the potential for significant cracking this outage was less likely. For technical basis, the licensee identified that their inspection would be in compliance with the Wesdye Technical Justification Document showing an effective demonstration of the volumetric leak path technique. The NRC has accepted the use of a demonstrated volumetric leak path as part of the upper head inspection program under 10 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced NUREG/CR-7142, Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation, which found , in part, the use of a properly focused 0 degree probe could detect a leakage path under low leakage rates during operation that led to minimal wastage of the upper head low alloy steel. While the NRC staff did not find that the volumetric leak path assessment was equivalent to a qualified surface leak path assessment, the information does

demonstrate the effectiveness of the volumetric leak path examination to detect low , Comment ICJ ): Removed due to leakage rates, as performed in accordance with the licensee's proposed alternative.  :' question of specifics by Regional staff. In

                                                                                                        / the NRC SE write-up I would consider For ha rdship, the licensee noted that a qualified surface leak path assessment could be                / adding a statement similar to...

performed in two manners that would require both additional radiolog ical dose and time / versus the performance of a volumetric leak path assessment. {l:A&-lisensee-estima.too i In consid erat ion of t he radiological dose 3.4 Rem and 10 days for an eddy current surface examination and 18 REM to perform a / estimat es, the NRC notes t hat t hese are liquid dye penetrant examination of all of the 12 penetration welds. )I~~J::J_f3_9_~~~ff.f9_y[l_g_j conservatively high estimates. The NRC both of these conditions to be of sufficient hardship given the operational experience and bases this evaluat ion on actual survey technical adequacy of the licensee's proposed alternative versus the regulatory dose rates in the area. Add itionally, t he requirement. NRC staff notes, that w it h prior planning, the effect of this inspection on outage Therefore, the NRC staff finds that the licensee's proposed alternative provides schedule cou ld be significantly reduced. reasonable assurance of structural integrity until the next scheduled examination, and However, the NRC staff did find that even that compliance with the surface examination requirements of Paragraph -3200(b) of a lower bound radiological dose estimate ASME Code Case N-729-1, for the subject welds, would result in hardship without a was st ill sufficient hardship based on t he compensating level of quality of the compensating increase in the level of quality and safety. vo lumetric examination given no previous cracking had been found previously and Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1, during this outage. Office of Nuclear Reactor Regulation I doubt that detail is necessary in t he As Chief of the Plant Licensing Bra nch IV-1 , Office of Nuclear Reactor Regulation, I verbal, but is included In this comment in concur with the Component Perfor mance, Non-Destructive Examination, and Testing case a question is raised by the region as Branch's determinations. part of ou r review conside rations for this relief. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CROM penetration nozzles numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 such that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2 ) and 10 CFR 50.55a(g)(6)(ii)(D). T he refore, the NRC staff authorizes the use of relief request 14R-03 at the Wolf Creek G enerating Station during the current refueling outage subject to the licensee's proposed alternative that if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee w ill revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspecto r. This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding Relief Request 14R-03, while preparing the subsequent written safety evaluation.

From: Collins, Jay Sent: 25 Oct 2016 11:36:33 +0000 To: Alley, David

Subject:

New wolf creek script Attachments: Wolf Creek verbal auth 14R-03 10-17-2016 Rev 3.docx Attached changes for reg ional concerns.

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELI EF REQUEST 14R-03 ALTERNATIVE TO USE VOLUMETRIC LEAK PATH FOR SUPPLEMENTAL EXAMS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated October 14, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-03 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration weld numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 at the Wolf Creek Generating Station. The licensee proposed (a) to perform a volumetric leak path assessment of each penetration nozzle in lieu of the surface leak path assessment required by Paragraph - 3200(b) of ASME Code Case N-729-1, and (b) if an unacceptable indication by the leak path assessment or volumetric exam is identified , the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that while the demonstrated volumetric leak path is not equiva lent to a fully qualified surface leak path assessment, the licensee identified sufficient operational experience, technical basis and radiological dose hardship to show that regulatory compliance would result in hardship without a compensating increase in the level of quality and safety. For operating experience, the licensee showed that there has been no previous identified cracking or leakage identified from the CRDM nozzle penetrations or welds of the upper head at Wolf Creek. The NRC staff noted that while this fact does not preclude the possibility of cracking to be found as the plant continues to age, plants which have previously identified cracking are more likely to see subsequent and more significant cracking in the future . Given the lack of the initial cracking being identified in the nozzle heats of material, at the operating temperatures of Wolf Creek, the NRC found that the potential for significant cracking this outage was less likely. For technical basis, the licensee identified that their inspection would be in compliance with the Wesdye Technical Justification Document showing an effective demonstration of the volumetric leak path technique. The NRC has accepted the use of a demonstrated volumetric leak path as part of the upper head inspection program under 10 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced NUREG/CR-7142, Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation, which found , in part, the use of a properly focused 0 degree probe could detect a leakage path under low leakage rates during operation that led to minimal wastage of the upper head low alloy steel. While the NRC staff did not find that the volumetric leak path assessment was equivalent to a qualified surface leak path assessment, the information does

demonstrate the effectiveness of the volumetric leak path examination to detect low , Comment ICJ ): Removed due to leakage rates, as performed in accordance with the licensee's proposed alternative.  :' question of specifics by Regional staff. In

                                                                                                        / the NRC SE write-up I would consider For ha rdship, the licensee noted that a qualified surface leak path assessment could be                / adding a statement similar to...

performed in two manners that would require both additional radiolog ical dose and time / versus the performance of a volumetric leak path assessment. {l:A&-lisensee-estima.too i In consid erat ion of t he radiological dose 3.4 Rem and 10 days for an eddy current surface examination and 18 REM to perform a / estimat es, the NRC notes t hat t hese are liquid dye penetrant examination of all of the 12 penetration welds. )I~~J::J_f3_9_~~~ff.f9_y[l_g_j conservatively high estimates. The NRC both of these conditions to be of sufficient hardship given the operational experience and bases this evaluat ion on actual survey technical adequacy of the licensee's proposed alternative versus the regulatory dose rates in the area. Add itionally, t he requirement. NRC staff notes, that w it h prior planning, the effect of this inspection on outage Therefore, the NRC staff finds that the licensee's proposed alternative provides schedule cou ld be significantly reduced. reasonable assurance of structural integrity until the next scheduled examination, and However, the NRC staff did find that even that compliance with the surface examination requirements of Paragraph -3200(b) of a lower bound radiological dose estimate ASME Code Case N-729-1, for the subject welds, would result in hardship without a was st ill sufficient hardship based on t he compensating level of quality of the compensating increase in the level of quality and safety. vo lumetric examination given no previous cracking had been found previously and Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1, during this outage. Office of Nuclear Reactor Regulation I doubt that detail is necessary in t he As Chief of the Plant Licensing Bra nch IV-1 , Office of Nuclear Reactor Regulation, I verbal, but is included In this comment in concur with the Component Perfor mance, Non-Destructive Examination, and Testing case a question is raised by the region as Branch's determinations. part of ou r review conside rations for this relief. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CROM penetration nozzles numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 such that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2 ) and 10 CFR 50.55a(g)(6)(ii)(D). T he refore, the NRC staff authorizes the use of relief request 14R-03 at the Wolf Creek G enerating Station during the current refueling outage subject to the licensee's proposed alternative that if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee w ill revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspecto r. This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding Relief Request 14R-03, while preparing the subsequent written safety evaluation.

From: Collins, Jay Sent: 25 Oct 2016 14:53:19 +0000 To: Ka1likian, Roger

Subject:

FW: Wolf Creek - RR 14R-03 (MF8456) From: Collins, Jay Sent: Thursday, October 20, 2016 4:14 PM To: Lingam, Siva <Siva.Lingam@nrc.gov>; Alley, David <David.Allcy@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>

Subject:

RE: Wolf Creek - RR 14R-03 (MF8456) -3141 General -3142.I Acceptance by VE (b) A component whose VE d!etects a relevant conditio shall be unacceptable for continued service unti l the requirements of -3 142.1 (b)( l ), (b )(2), and (c) below are met. (/) Components with relevant co11ditions require further evaluatio11. This evaluation shall include determination of the source of the leakage and correction of the source of leakage in accordance with -3142.3. (2) All relevant conditions shall be evaluated to determine the extent, if any, of degradation. The boric acid crystals and residue shall be removed to the extent necessary to allow adequate examinations and evaluation of degradation, and a subsequent VE of the previously obscured surfaces shall be performed, prior to return to service, and again in the subsequent: refueling outage. Any degradation detected shall be evaluated to determine if any corrosion has impacted the strucniral integrity of the component. Corrosion that has reduced component wall thickness below design limits shall be resolved through repair /replacement activity in accordance with IWA-4000. (£.! A nozzle whose VE indicates elevant conditions jndicative of possible nozzle leakage shall be unacceptable for continued service unless it meets the requirements of -3142.2 or -3142.3. -3142.2 Acceptance by SuJ!Plemcotal .Examination. A nozzle with relevant conditions indicative of possible J10zzle leakage shall be acceptable for continued service if the results of supplemental examinations [-3200(b)] meet the requirements of -3130. -3142.3 Accep tance by Corrective Measures or Repair/Replacement Activity (a) A component with relevant conditions not indicative of possible nozzle leakage is acceptable for continued

service if the source of the relevant condition is corrected by a repair /replacement activity or by corrective measures necessary to preclude degradation. (b) A component with relevant conditions indicative of possible nozzle leakage sha ll be acceptable for continued service if a repair/replacement activity corrects the defect in accordance with IW A-4000. -3 140 INSERVICE VISUAL EXAM INATIONS (VE) -3141 General ~elevant conditions for the purposes of the VE{ s hall include areas of corrosion, boric acid deposits, discoloration, and other evidence of nozzle leakage. The NRC considers any relevant condition in the annulus region between the nozzle and head surface that cannot be removed by light cleaning activities to be a relevant condition of possible nozzle leakage.

---

Original Appointment-----

From: Lingam, Siva Sent: Thursday, October 20, 20 16 3:5 1 PM To: Lingam, Siva; Alley, David; Collins, Jay; Tsao, John; Drake, James; Taylor, Nick; Dodson, Douglas; Thomas, Fabian; Proulx, David; Kopriva, Ron; Anchondo, Isaac; Cumblidge, Stephen Cc: Pascarelli, Robert

Subject:

Wolf C reek - RR 14R-03 (MF8456) When: Thursday, October 20, 2016 4:15PM-5:15 PM (UTC-05:00) Eastern Time (US & Canada). Where: HQ-OWFN- 1OB06-12p Please note the following to discuss the subject RR based on the attached response from the licensee: Bridge No.: 877-935-1422 Passcode: l(b)(6) Ifollowed by # Date: October 20, 2016 (Thursday) Time: 4: 15 PM (Eastern Time) << File: ETl6-0028.pdf >>

From: Dean, Bill Sent: 25 Oct 2016 09:18:26 -0400 To: Pascarelli, Robert

Subject:

Re: Wolf Creek nozzle leakage Thanks Bob On: 25 October 201 6 02 : 17, "Pascarelli, Robert" <Robert.Pascarelli@nrc.gov> wrote: Bill/Michele/Brian, Anne Boland indicated that were interested in the discussions between the program office, Region IV, and the licensee regarding the vessel head repair at Wolf Creek. Attached for your information is a one-pager that Reg ion IV updated late last week that describes the technical specification required shutdown on September 2,2016 and subsequent repair activities. The source of the leak was determined to be a leaking canopy seal weld on a core exit thermocouple penetration nozzle above the reactor vessel head. The large amount of boric acid obscured the visual inspection of twelve vessel head nozzles as a result of the spray pattern from the leaking nozzle. The code case requires a surface examination of the of the partial penetration welds from the bottom side of the reactor head. Wolf creek is requesting relief from this requirement due to dose implications from conducting the surface examination under the head as well as the impact on the outage schedule. FYI, Wolf Creek had a planned refueling outage a few weeks after the forced shutdown and remained shutdown to begin the outage early. The staff has been reviewing the relief request and we have had a few conference calls with the licensee as well as Region IV. We have been fortunate to have some DRS inspectors and the DRP Branch Chief on-site and they have shared information regarding the licensee's progress on cleaning the head and the ongoing visual examinations. We are awaiting some additional information from Wolf Creek that they agreed to provide to the staff during a conference call on Friday. Please let me know if you have any questions. Bob Pascarelli Bob Pascarelli, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

From: Alley, David Sent: 26 Oct 2016 16:14:03 -0400 To: M ateychick, John;Taylor, Nick;Proulx, David;Thomas, Fabian;Singal, Balwant;Lingam, Siva

Subject:

RE: Wolf Creek Relief Request? In my opinion, up to a point, the schedule belongs to them. I just want to make sure everyone understands that it will take my folks some amount of time to go through whatever is submitted. I don't relish the thought of this continuing to be delayed and then the licensee wanting a snap decision. Whatever can be done to ensure that the licensee understands this issue would be appreciated. Dave From: Mateychick, John Se nt: Wednesday, October 26, 2016 4 :11 PM To: Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Thomas, Fabian

<Fabian.Thomas@ nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; Lingam, Siva
<Siva.Lingam@nrc.gov>; Alley, David <David.Alley@nrc.gov>

Subject:

RE: Wolf Creek Relief Request? Not yet. They had promised to get information to us on Thrusday. John M From: Taylor, Nick Sent: W ednesday, October 26, 2016 3:01 PM To: M ateychick, John <John.M ateychick@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Thomas, Fabian <Fabian.Thomas@ nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; Lingam, Siva

<Siva.Lingam@nrc.gov>; Alley, David <David .Alley@nrc.gov>

Subje ct: Wolf Creek Relief Request?

All, Pretty quiet on the Wolf Creek front with regard to the head. Have we heard anything from the licensee on the photos, etc requested in support of their relief request?

Nick Taylor 972-921-6398

From: Greene, Louis Sent: 26 Oct 2016 14:13:00 -0400 To: WolfCreekEIS Resource;Watford, Margaret;Singal, Balwant;RidsRgn4MailCenter Resource; RidsResDE Resource; RidsNrrPMWolfCreek Resource ;RidsNrrDorllpl4-2 Resource;RidsManager Resource;Regner, Lisa;Pascarel li, Robert;Lyon, Fred;Lingam, Siva;Burkhardt, Janet

Subject:

Wolf Creek Response to Request for Additional Information re Relief Request Number 14R-03 for Reactor Vessel Head Penetration Nozzle Welds. Attachments: 8project_Tem pFil es_eRids_ erid6669556580576731470. rtf ADAMS Distribution Notification A047 - OR Submittal: Inservice/Testing/Relief from ASME Code; related correspondence Open ADAMS PS Document(Wolf Creek Response to Request for Additional Information re Relief Request Number I4R-03 for Reactor Vessel Head Penetration Nozzle Welds.) View ADAMS PS Properties ML16300A214 Accession MLJ 6300A2 l 4 Number Wolf Creek Response to Request for Additional Information re Relief Request Title Number I4R-03 for Reactor Vessel Head Penetration Nozzle Welds. Docket 05000482 Number Document 10120120 16 Date Author McCoy J H Name Author Wolf Creek Nuclear Operating Corp Affiliation Addressee Name Addressee NRC/Document Control Desk Affiliation NRC/NRR Document Letter Type Response to Request for Additional Information (RAI) Availability Publicly Available Date to be 11/04/2016 Released Document Non-Sensitive Sensitivity Comment Date Added l 0/26/20 16 daw5 Keyword DPCautoadd

• rr3 ADAMS Distribution Sheet Priority: Normal From : Greene, Louis Assi2ned Recipients:

Watford, Margaret 0 OK Singal Balwant 0 OK RidsNrrPMWolfCrcek Resource 0 OK RidsNrrDorlLpl4 Resource 0 OK Regner Lisa 0 OK Pascarelli, Robert 0 OK Lyon, Fred 0 OK Lingam, Siva 0 OK Burkhardt, Janet 0 OK Internal Recipients: zzzFTLE CENTER 01 0 Not Found WolfCreek Resource 0 OK R idsRgn4Ma i1Center 0 OK RidsResDE Resource 0 OK RidsManager Resource 0 OK Total Copies: 1 Distribution Codes Used I A047 I OR Submittal: Inservice!festing/Relieffrom ASME Code; related corres pondence Document Profile Accession Number ML l 6300A214 Title Wo lf C reek Response to Request for Additional information re Relief Request Number J4R-03 for Reactor Vessel Head Penetration Nozzle Welds. Docket Number 05000482 Document Date 10120/201 6 Author Name McCoy J H Author Affiliation Wo lf C reek Nuclear Operating Corp Addressee Name Addressee Affiliation NRC/Document Control Desk NRC/NRR Document Type Letter Response to Request for Additional Information (RA!) Availability Publicly Avai lable Date to be Released 11 /04/2016 Document Sensitivity Non-Sensitive

Comment Date Added 10/26/2016 Keyword dawS DPCautoadd jrrJ

From: Alley, David To: Sjogal Balwa ot

Subject:

Wolf creek Date: Thursday, October 27, 2016 8:00:00 PM

Balwant, I failed in my duties.

The licensing guy from Wolf Creek left me a voice mail this PM. I managed to listen to it but not respond. I left his name and number in the office. His question was about whether we needed all nozzles or only the nozzles other than those included in the relief request. My answer is that if they have all the nozzles it would be beneficial for us to see them. The real need however, is for the other nozzles. The plant has already acknowledged that the 12 nozzles in the relief request require further inspection (and it is my understanding that they have inspected as proposed in the relief request). The real need at this point is to determine whether additional nozzles require further inspection and whether there is less reason to accept the concept that all the boric acid came from above (which impacts whether the relief is authorized). The plant needs to make sure that they are fully in compliance with the code case and ultimately with other applicable programs such as the boric acid program. As a result after the final cleaning, the region will probably be interested in the condition of all the nozzles, not just the "other" nozzles Dave David Alley PhD. Chief, Component Performa nce NDE and Testing Bra nch US Nuclear Regu latory Commission 11555 Rockville Pike Rockville MD 20852 301-415-2178

From: Alley, David To: Farnan Michael

Subject:

For tomorrow Date: Thursday, October 27, 2016 8:06:00 PM May be some activity on Wolf Creek head tomorrow. Pictures are supposed to be posted overnight. 3 PM phone call. I will try to be on the call. Conceivably we could get as far as issuing a verbal. If so, you can read my part. Stephen has the script. Stephen will be around all day tomorrow and Roger will be in in the morning. Both have been following the issue. Cant think of anything else that is hot at the moment - including my wood stove. Didn't check it often enough Dave David Alley PhD. Chief , Component Performance NOE and Testing Branch US Nuclear Regu latory Commission 11555 Rockvil le Pike Rockvil le MD 20852 301-415-2178

From: Taylor, Nick Sent: 27 Oct 2016 13:44:20 -0500 To: Alley, David; Drake, James; Melfi, Jim Cc: Proulx, David;Werner, Greg;Anchondo, lsaac;Coll ins, Jay; Kopriva, Ron

Subject:

RE: Wolf Creek Head I think Davea's question is a good one and should probably be answered before we decide to have a call. Nick From : Alley, David Se nt: Thursday, October 27, 2016 11:09 AM To: Drake, James; Melfi, Jim Cc: Taylor, Nick; Proulx, David ; Werner, Greg; Anchondo, Isaac; Collins, Jay ; Kopriva, Ron

Subject:

RE: Wolf Creek Head

Jim, Does this mean that they are going to submit info to us to review in sufficient time that we can review it before the call? Not sure what a call will accomplish until we have looked at their findings.

Dave From : Drake, James Se nt: Thursday, October 27, 2016 11:55 AM To: M elfi, Jim <Jim.Melfi@nrc.gov> Cc: Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Werner, Greg

<Greg.Werner@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Alley, David
<David.Alley@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>

Subject:

Wolf Creek Head

Jim, Wolf Creek just contacted me. They want to have a call tomorrow at 1000 to discuss the head relief request.

They are going to call Siva to set up a bridge. Jim $mne.r <.f. :Drake James F. Drake Office phone: 8 17-200-1558 Cell Phonej<b)(6) I

From: Alley, David To: Wl.aWix

Subject:

RE: FW: Post-cleaning pictures of reactor head uploaded to CERTREC Date: Friday, October 28, 2016 9:22:00 At.! Current plan is a call at 3PM eastern. Dave From: Collins, Jay Sent: Friday, October 28, 2016 8:00 AM To: Alley, David <David.Alley@nrc.gov>

Subject:

RE: FW: Post-cleaning pictures of reactor head uploaded to CERTREC Greetings, I looked this morning and do not see them, therefore I guess the 10am call is not happening. If you need me to look at something, give me a call, but I will be in transit from - 9am to -11am. Jay From: Drake, James Sent: Thursday, October 27, 2016 3:25 PM To: Taylor, Nick <Nick Taylor@nrc goy>; Proulx, David <Dayjd proulx@nrc goy> Cc: Werner, Greg <Greg Werner@nrc gov>: Anchondo, Isaac <Isaac Anchondo@nrc gov>; Kopriva, Ron <Ron Kopriva@nrc gov>; Alley, David <Qayjd Alley@nrc gay>; Collins, Jay <Jay Colljns@nrc goy>; Cumblidge, Stephen <Stephen Cumbljdge@nrc eoy>; Melfi, Jim <Jim Melti@orc gov>

Subject:

RE: FW: Post-cleaning pictures of reactor head uploaded to CERTREC

Nick, The pictures that Wolf Creek has loaded to Certrec now are of the flange area. They won't help with the relief request.

Reece said he doesn't think those pictures will be available until late tonight or tomorrow. The 1000 call for the relief request is contingent on them providing the pictures with sufficient time for NRC personnel to review them. Jim From: Taylor, Nick Sent: Thursday, October 27, 2016 1:45 PM To: Proulx, David <Oayjd Proulx@nrc 1,:oy> Cc: Drake, James <James Orake@nrc goy>

Subject:

RE: FW: Post-cleaning pictures of reactor head uploaded to CERTREC I will try to dial in tomorrow morning but right now I have a meeting scheduled onsite during that time. David, can you represent the branch on this call in case I am unable to join?

Thanks, Nick From: Drake, James Sent: Thursday, October 27, 2016 11:16 AM To: Alley, David <David Alley@orc gov>: Collins, Jay <Jay Col!ins@nrc gov>: Cumblidge, Stephen <Stephen Cumbljdge@nrc gov>

Cc: Taylor, Nick <Nick Taylor@nrc gov>; Anchondo, Isaac <Isaac Anchondo@nrc gov>; Werner, Greg <Greg Werner@nrc goy>; Kopriva, Ron <Ron Koprjya@orc goy>; Lingam, Siva <Sjya Lingam@nrc goy>; Proulx, David <Oayjd proulx@ orc gay>

Subject:

FW: FW: Post-cleaning pictures of reactor head uploaded to CERTREC I think I got everyone. The pictures have been uploaded to item 12 in the 402016 Integrated Inspection folder on CERTREC. If you do not have access, let me know and I will try to get you added to the list.

Jim From : Hobby Reece D [m ajlt o*rehobby@ WCNOC com] Sent: Th ursday, October 27, 2016 11:14 AM To: Drake, James <James Prake@ orc goy>

Subject:

[External_Sender] FW: Post-deaning pictures of reactor head uploaded to CERTREC From: Hobby Reece D Sent: Thursday, October 27, 2016 7:33 AM To: Th omas Fabian D; Dodson Douglas E; 'KOPRIVA, Ron A' Cc: Vickery Brad J; Barraclough Richard M; Good Nicole R; Stone Lucille M; Muilenburg William T; Hafenstine Cynt hia R

Subject:

Post-cleaning pictures of reactor head uploaded to CERTREC Fabian, Doug and Ron : The most-recent pictures ta ken after the reactor head was cleaned have been u ploaded to item 12 in the 402016 Int egrated Inspection folder on CERTREC in accord ance w ith your request. Final clean ing of the reactor head is current ly scheduled for 1600 on October 30, 2016 but that schedule could cha nge based on t he progress of work in t he next few days. We w ill not ify t he resident inspectors about the schedule for t he fina l cleaning. Reece

From: Cu mblidge, Stephen Sent: 28 Oct 2016 13:42:47 -0400 To: Alley, David;Collins, Jay;Singal, Balwant;Kalikian, Roger;Tsao, John; Drake, James;Taylor, Nick; Proulx, David; Regner, Lisa; Werner, Greg;Anchondo, lsaac;Kopriva, Ron;Thomas, Fabian

Subject:

RE: Internal NRC Call to Discuss Wolf Creek Relief Request I agree. The language does not appear to be consistent with the code case, the order, or EPRI guidance. Stephen Cumblidge Materials Engineer US Nuclear Regulatory Commission Mail Stop OWFN/9 H6 Washington, DC 20555-000 1 Telephone: (301) 415-2823 (Office) F rom: Alley, David Sent: Friday, October 28, 20 16 1:36 PM To: Collins, Jay <Jay.Collins@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; Kalikian, Roger <Roger.Kalikian@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>; Regner, Lisa <Lisa.Regner@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

Subject:

RE: Internal NRC Call to Discuss Wolf Creek Relief Request

Folks, I have been through the spreadsheet, the evaluation document and many but not all the photos. I haven't seen anything in the photos that would convince me that only the 12 nozzles had relevant conditions of potential nozzle leakage (or whatever the precise words arc). I also didn't sec anything that confirms that all the relevant conditions have been removed IA W guidance on the subject. Jn the spreadsheet and the evaluation document I found the following:

Nothing emanating from the annulus region was confirmed. (from spreadsheet and page 4 of 5 of evaluation document) Portions of the annulus could be observed without removing all residue and it was determined that this was the extent necessaty to allow adequate examination (page 2 of 5 of evaluation document) Because an adequate examination was performed showing that none of the remaining population of 66 nozzles had nozzle leakage, it was not necessa1y to completely remove the accumulation present on many nozzles to satisfy the objective of determining the absence of nozzle leakage from a crack in the nozzle or j groove weld. (page 3 of 5 of the evaluation document) These statements appear to me to indicate that the licensee's actions arc not consistent with the code case - I believe that the standard is possible leakage not confirmed leakage Thoughts?

Dave From: Collins, Jay Sent: Friday, October 28, 2016 11 :34 AM To: Singal, Balwant <I3alwant.Singal@nrc.gov>; Alley, David <David.Allcy@ nrc.gov>; Kalikian, Roger <Rogcr.KalikianCdlnrc.gov>; Tsao, John <John.Tsao@nrc .gov>; Drake, James <Jamcs.Drakc@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <Davicl.Proulx@nrc.gov> ; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>; Regner, Lisa <Lisa.Regner@nrc.gov>; Werner, Greg <Greg.Wemer@ nrc.gov>; Anchondo, Isaac <Isaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron.Kopriva@ nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

Subject:

RE: Internal NRC Call to Discuss Wolf Creek Relief Request ASME Code Case N-729-1 Note I The VE shall consist of the following: (a) A direct examination of the bare-metal surface of the entire outer surface of the head, including essentially I 00% of the intersection of each nozzle with the head. lf welded or bolted obstructions are present (i.e., mirror insulation, insulation support feet, shroud support ring/lug), the examination shall include 2::95% of the area in the region of the nozzles as defined in Fig. I and the head surface uphill and downhill of any such obstructions. The examination may be performed with insulation in place using remote equipment that provides resolution of the component metal surface equivalent to a bare-metal direct examination. (b) The examination may be perforn1cd with the system dcpressurizcd. (c) The examination shall be performed with an illumination level and a sufficient distance to allow reso lution of lower case charncters not greater than 0.105 in. (2. 7 mm) in height.

---

Original Appointment-----

From: Singal, Balwant Sent: Friday, October 28, 2016 10: 17 AM To: Collins, Jay; Alley, David; Kalikian, Roger; Tsao, John; Drake, James; Taylor, Nick; Proulx, David; Cumblidge, Stephen; Regner, Lisa; Werner, Greg; Anchondo, fsaac; Kopriva, Ron; Thomas, Fabian

Subject:

Internal NRC Call to Discuss Wolf Creek Relief Request When: Friday, October 28, 20 16 11 :00 AM-12:00 PM (UTC-05 :00) Eastern Time (US & Canada). Where: Dave Alley's Office Dave Alley and me received a call from Wolf Creek (Cyndia and Jaimme McCoy) at 9.30 this morning. An internal NRC staff meeting is required lo discuss path forward based on information provided during the call. Bridge No. Info. 866-624-3402 Passcodc: l(b)(6)

Lisa: You will need to use Passcode Fb)(6) I(as initiator of the call). I was not bale to search for conference rooms from home. I will be out-of-office j(b)(6) t for about 3 hours and Lisda will be supporting this call. l can be contacted a~....(b-)(-6)_ _ __.I for any questions. Thanks.

From: Cymbljdge. Stephen To: Lubinski ! oho* Evans Michele* McDermott Brian Cc: Alley. Dayjd

Subject:

Wolf Creek Phone Call Date: Friday, October 28, 2016 5:21:28 PM We have completed our phone call with the staff at Wolf Creek regarding the examinations of the upper head at Wolf Creek. A canopy seal leak resulted in boric acid and corrosion product deposits on their upper head, necessitating further examinations to determine if the leakage could possibly have come through any of the control rod drive nozzles. The NRC staff included, but was not limited to, Balwant Singal, David Alley, Jay Collins, and Jim Drake. The phone call largely revolved around a difference of opinion between the NRC staff and the licensee about the acoeptance criteria for bare metal visual examinations (VE). The NRC staff has maintained, based on operating experience, that small amounts of adherent boric acid that remain after a light cleaning are considered relevant. The licensee was of the opinion that small amounts of boric acid obscuring parts of the nozzles was acceptable. The licensee made two statements of significance. "Due to boric acid deposits from the canopy seal weld leak on Nozzle 77, all nozzles were categorized as having relevant conditions." IAW Paragraph -3142.1, this statement requires the licensee to perform a VE exam on all nozzles to determine if there is a relevant condition indicative of possible nozzle leakage. Note 1 of Table 1 of ASME Code Case N-729-1 states that the VE shall consist of (a) A direct examination of the bare-metal surface of the entire outer surface of the head, including essentially 100% of the intersection of each nozzle with the head. The second critical statement from the licensee's report was, "Because an adequate examination was performed showing that none of the remaining population of 66 nozzles had nozzle leakage, it was not necessary to completely remove the accumulation present on many nozzles to satisfy the objective of determining the absence of nozzle leakage from a crack in the nozzle or j groove weld." This statement is incongruent with a VE exam. After a lengthy discussion about the requirements of N-729-1 , the licensee appeared to understand the NRC's position. The licensee also expressed some confusion as to the use and requirements and use of VT-2 examinations and VE examinations. The NRC staff described the applicability of these two examination methods for upper head examinations. After the discussion it was determined that possible paths forward include conducting supplemental ultrasonic examinations of the affected nozzles or to submit a relief request to defer the inspections. The licensee requested a later phone call with the NRC to discuss which path they will take to resolve this situation. Stephen Cumblidge Materials Engineer US Nuclear Regulatory Commission Mail Stop OWFN/9 H6 Washington, DC 20555-0001 Telephone: (301) 415-2823 (Office)

From: Regner, Lisa Sent: 28 Oct 2016 13:49:14 -0400 To: Singal, Balwant Cc: Pascarelli, Robert

Subject:

Wolf Crik update - internal phone call Current status:

• The licensee provided photos on Certrec of the head and annulus region before and after washing.
• The staff is reviewing the photos and will email the group when they've reached a conclusion on their acceptability within the scope of the existing RR
• The staff will have another internal call prior to the call with the licensee
• At about 2 pm ET, a decision needs to be made whether to delay the call with the licensee to allow the staff to align.
• Late breaking emails indicate that the staff has not been convinced that the scope of the existing RR is adequate.

Discussion from call today at 11 am ET:

• the region has seen indications of boric acid and wastage on nozzles other than the 12 nozzles in the RR
• the licensee stated in the existing relief request that there are other indications outside of the 12
• WC states that they have followed the Code Case N-729, but have not, and apparently can not, provided documentation showing this
• WC has retained the equipment necessary to do additional volumetric leak path testing
• acceptance criteria is to do a 360 degree visual examination unless there is physical limitation preventing nothing larger than aspirin-sized adherent boric acid
• staff is likely to want to expand the RR for the licensee to do additional visual or volumetric leak path testing
• additionally, the staff expects to not grant relief until the head is cleaned and the licensee provides the results of visual examinations
• the code requires enough removal of debris to be able to make a determination of leakage; the staff does not believe WC has done this
• region stated WC has scheduled to clean the head on Sunday Lisa Regner Sr. PM NRR/DORL/LPL4-1 301-415-1906 080 08

From: Singal, Balwant Sent: 28 Oct 2016 23:48:17 +0000 To: Pascarelli, Robert; Boland, Anne;Evans, Michele; Benner, Eric;Lubinski, John Cc: Lingam, Siva;Regner, Lisa;Alley, David;Taylor, Nick

Subject:

Wolf Creek Relief Request - Control Rod Drive Mechanism (CROM) Nozzle Penetrations

All, There were multiple calls internally between the NRC staff (headquarters and Region IV) to get internal alignment on path forward and with the licensee to discuss the latest status and path forward.

The call with the licensee at 3.30 PM was participated by several NRC staff members. The NRC staff informed the licensee that based on the available information, the NRC staff does not believe that the licensee is in compliance with the Code Case N-729-1for66 nozzles (in addition to 12 already covered by the relief request). The licensee requested for another call later during the day after considering the staff input and evaluate available options. There was another call at 6.30 PM participated by several members of Wolf Creek Nuclear Operating Corporation, Nick Taylor from Region IV, Dave Alley from DE , and Balwant Singal from DORL. The licensee appeared to be convinced that they are not in compli.ance with the Code Case for 66 penetrations in question. The licensee proposed two separate options in lieu of performing the volumetric leak path assessment of all 66 penetrations. OPTION 1 The first option was to ask a relief request for additional 66 penetrations not to perform any additional examinations based on hardshi p (dose, schedule impact, and cost) and provide justification based on what has already been done. The NRC staff (Dave Alley) informed the licensee this option is not likely to have sufficient justification and does not seem to have a successful path. OPTION 2 The licensee proposed to perform additional examination from underneath the head similar to the one's proposed for previous 12 penetrations and us the results of the examinations to justify all remaining penetrations. The NRC staff (Dave Alley) indicated that it is up to the licensee to choose the desired path, but the potential for success is low. The other option discussed was for the licensee to include rest of the 66 penetrations also in the scope of the relief request by performing the leak path assessments for all the penetrations as proposed in the existing relief request. This path appeared to have greater potential for success. The licensee was informed that the NRC staff is not suggesting any particular option and it is up to the licensee t o pick path forward based on their best judgement. The call ended with this discussion. The licensee will inform the NRC staff about their decision after internal discussion. It appears that the licensee is convinced that they need to perform leak path assessments for all the 66 remaining penetrations.

Thanks. Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwa nt.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222

From: Werner, Greg Sent: 3 Oct 2016 14:32:45 -0500 To: Alley, David

Subject:

RE: Internal communications at Wolf Creek re head corrosion I plans 1 pm CDT From: Taylor, Nick Se nt: Monday, October 03, 2016 2:32 PM To: Werner, Greg <Greg.Werner@nrc.gov>; Alley, David <David.Alley@nrc.gov>

Subject:

RE: Internal communications at Wolf Creek re head corrosion I plans

All, We are a go for 1:00 pm tomorrow. We are getting a conference bridge set up and wi ll send out an appointment notice. I asked Wolf Creek to load up any available images, evaluations, etc before the call. Not sure if we will get anything .. ..

More to follow, Nick From : Werner, Greg Sent: Monday, October 03, 2016 1:34 PM To: Alley, David <David.Alley@nrc.gov> Cc: Taylor, Nick <Nick.Taylor@nrc.gov>

Subject:

RE: Internal communications at Wolf Creek re head corrosion I plans Yes. I will let Nick Taylor know to add you to the appointment and provide any details we might get before then. Greg From : Alley, David Sent: Monday, October 03, 2016 1:18 PM To: Werner, Greg <Greg.Werner@nrc.gov>

Subject:

FW: Internal communications at Wolf Creek re head corrosion I plans Greg We would like to be on the call tomorrow. Dave From : Tsao, Joh n Se nt: Monday, October 03, 2016 1:29 PM To: Alley, David <David.Alley@nrc.gov>

Subject:

RE: Internal communications at Wolf Creek re head corrosion I plans Dave, Yes we should be on the call with Wolf Creek tomorrow

From: Alley, David Sent: Monday, October 03, 2016 1:21 PM To: Tsao, John <John.Tsao@nrc.gov>

Subject:

FW: Internal communications at Wolf Creek re head corrosion I plans John Please take a look at this Greg Just tried to call - no answer . I am tied up for a while this PM. Might be good for us to be on the call tomorrow Dave From : Werner, Greg Sent: Monday, October 03, 2016 1:08 PM To: Alley, David <David.Alley@nrc.gov> Cc: Taylor, Nick <Nick.Taylor@nrc.gov>

Subject:

FW: Internal communications at Wolf Creek re head corrosion I plans FYI. Just giving you a heads up in case WC asks for relief. NO OTHER information other than what is in the attached file, which is part of a CR and an internal WC newsletter. We are planning an informational call with WC sometime tomorrow, would you like to be included on the appointment? We are trying to find out the status of the head cleaning, information on potential relief requests, and how they selected the other 4 penetrations for the clamps. Greg Werner From : Taylor, Nick Sent: Monday, October 03, 2016 11:50 AM To: Werner, Greg <Greg.Werner@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov> Cc: Pruett, Troy <Troy.Pruett@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>; Lantz, Ryan

<Ryan.Lantz@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov>; Proulx, David <David .Proulx@nrc.gov>;

Janicki, Steven <Steven.Janicki@nrc.gov>

Subject:

Internal communications at Wolf Creek re head corrosion I plans

All, I'm still working on setting up a call with the licensee tomorrow. But Doug provided the attached today from the licensee's CAP and internal outage newsletters. I added the red comment boxes.

Thanks, Nick

From : Reimer. Lisa To: Pascarelli Robert

Subject:

PN: Wolf Creek head inspection update 10-31-201 6 Date: Monday, October 31, 2016 4:20:00 PM Attachments: image001.ong Good news ... Lisa Regner Sr. PM NRR/DORL/LPL4-1 301-415-1906 08D08 From: Taylor, Nick Sent: Monday, Oct ober 31, 20 16 4 :18 PM To: Pruett, Troy <Troy.Pruett @nrc.gov>; La ntz, Ryan <Rya n.Lantz @nrc.gov>; Clark, Jeff <Jeff.Clark@ nrc.gov>; Vegel, Anton <Ant on.Vegel@ nrc.gov>; Kennedy, Kriss <Kriss.Kennedy@ nrc.gov>; Morris, Scott <Scott. M orris@nrc.gov>; Alley, David <David.Alley@nrc.gov>; Werner, Greg <G reg.Werner@nrc.gov>; Si nga l, Ba lwant <Balwant.Singa l@nrc.gov>; Lingam, Siva <Siva.Lingam@ nrc.gov>; Regner, Lisa <Lisa. Regner@nre.gov> Cc: Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabia n <Fabian.Thomas@ nrc.gov>; Proulx, David <David .Proulx@nrc.gov>

Subject:

W olf Creek head inspect ion updat e 10-31-2016

All, By way of updates .. . I spoke a few minutes ago with Fabian Thomas (resident@ Wolf Creek), who observed that the licensee is now making plans and mobilizing equipment to perform ultrasonic inspections of the remaining 66 nozzles on the reactor vessel head.

Fabian reports that the licensee's schedule shows this activity commencing on November

4. We don't yet know how long the activity will take.

Fabian also reports that the licensee is scheduled to begin their bare metal visual inspection of the vessel head on November 5. We are still expecting the licensee to need relief from the surface examinations of the j-groove welds for the 66 additional penetrations. The licensee has not yet communicated their plans for requesting this additional relief (or anything else for that matter).

Thanks, Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV 0 : 817 200-1141 C: (b)(6)

E: ojck.taylor@nrc.gov 4 l'tHnt*-t fttt/14...J ,N/_,,,.,,.._.,,

From: Pascarelli, Robert Sent: 1Nov2016 12:17:49 +0000 To: Wilson, George Cc: Boland, Anne

Subject:

FW: Wolf Creek head inspection update 10-31-2016 FYI, good news from Wolf Creek. From: Lingam, Siva Sent: Monday, October 31, 2016 4:22 PM To: Collins, Jay <Jay.Collins@nrc.gov>; Tsao, John <John.Tsao@nrc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>

Subject:

RE: Wolf Creek head inspection update 10-31-2016 From: Taylor, Nick Sent: Monday, October 31, 2016 4:18 PM To: Pruett, Troy <Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>; Clark, Jeff

<Jeff.Clark@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>; Kennedy, Kriss <Kriss.Kennedy@nrc.gov>;

Morris, Scott <Scott.Morris@nrc.gov>; Alley, David <David .Alley@nrc.gov>; Werner, Greg

<Greg.Werner@nrc.gov>; Singal, Balwant <Balwant.Singal@nirc.gov>; Lingam, Siva
<Siva.Lingam@nrc.gov>; Regner, Lisa <Lisa.Regner@nrc.gov>

Cc: Dodson, Douglas <Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>

Subject:

Wolf Creek head inspection update 10-31-2016

All, By way of updates ... ! spoke a few minutes ago with Fabian Thomas (resident@ Wolf Creek),

who observed that the licensee is now making plans and mobilizing equipment to perform ultrasonic inspections of the remaining 66 nozzles on the reactor vessel head. Fabian reports that the licensee's schedule shows this activity commencing on November 4. We don't yet know how long the activity will take. Fabian also reports that the licensee is scheduled to begin their bare metal visual inspection of the vessel head on November 5. We are still expecting the licensee to need relief from the surface examinations of the j-groove welds for the 66 additional penetrations. The licensee has not yet communicated their plans for requesting this additional relief (or anything else for that matter).

Thanks, Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV

R From: Alley, David Sent: 3 Nov 2016 20:44:18 -0400 To: Pascarelli, Robert Cc: Singal, Balwant;Collins, Jay;Cumblidge, Stephen;Caponiti, Kathleen

Subject:

Request for additional Information, Wolf Creek, 14R-03 and 14R-03, Nozzle exam Attachments: Wolf Creek RAI Rev2.docx, Wolf Creek verbal auth 14R-03 11-03-2016.docx, Wolf Creek verbal auth 14R-04 10-12-2016.docx

1. I concur with the attached RAI

2. I concur with both of the verbal scripts

3. Stephen, please draft a 665 for the RAI and send it to Kathleen

4. If we get the RAI sent out and a response back, I have no objection to doing the verbals on Friday. If we don't get a response back for the RAI , I propose we do the verbals on Monday.

I am open to doing them over the weekend if needed. If done tomorrow, Stephen is acting and will do my part. Dave David Alley PhD. Chief, Component Performance NOE and Testing Branch US Nuclear Regulatory Commission 11555 Rockville Pike Rockville MD 20852 301-415-2178

REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST 14R-03 ALTERNATE EXAMINATION OF REACTOR PRESSURE VESSEL UPPER HEAD NOZZLE PENETRATIONS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 By letter dated November 1, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-03 for the alternate examination of all 78 control rod drive mechanism (CROM) nozzle penetration welds at the Wolf Creek Generating Station. The licensee proposed (a) to perform a volumetric leak path assessment of each penetration nozzle in lieu of the surface leak path assessment required by Paragraph -3200(b) of ASME Code Case N-729-1, and (b) if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. To complete its review, the Nuclear Regulatory Commission (NRC) requests the following additional information.

1. On page 4 of 8 of the request, the licensee states that the radiological dose estimated for the eddy current surface examination of 66 of the penetration welds would be 500 mRem. What is the total estimated radiological dose for the performance of the eddy current surface examination on all 78 of the penetration welds?

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 14R-03 ALTERNATIVE TO USE VOLUMETRIC LEAK PATH FOR SUPPLEMENTAL EXAMS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated November 1, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-03 for the alternate examination of all 78 control rod drive mechanism (CROM) nozzle penetration welds at the Wolf Creek Generating Station. The licensee proposed (a) to perform a volumetric leak path assessment of each penetration nozzle in lieu of the surface leak path assessment required by Paragraph - 3200(b) of ASME Code Case N-729-1, and (b) if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that while the demonstrated volumetric leak path is not equivalent to a fully qualified surface leak path assessment, the licensee identified sufficient operational experience, technical basis and radiological dose hardship to show that regulatory compliance would result in hardship without a compensating increase in the level of quality and safety. For operating experience, the licensee showed that there has been no previous identified cracking or leakage identified from the CRDM nozzle penetrations or welds of the upper head at Wolf Creek. The NRC staff noted that while this fact does not preclude the possibility of cracking to be found as the plant continues to age, plants which have previously identified cracking are more likely to see subsequent and more significant cracking in the future. Given the lack of the initial cracking being identified in the nozzle heats of material, at the operating temperatures of Wolf Creek, the NRC found that the potential for significant cracking this outage was less likely. For technical basis, the licensee identified that their inspection would be in compliance with the Wesdye Technical Justification Document showing an effective demonstration of the volumetric leak path technique. The NRC has accepted the use of a demonstrated volumetric leak path as part of the upper head inspection program under 10 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced NUREG/CR-7 142, Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation, which found, in part, the use of a properly focused 0 degree probe could detect a leakage path under low leakage rates during operation that led to minimal wastage of the upper head low alloy steel. While the NRC staff did not find that the volumetric leak path assessment was equivalent to a qualified surface leak path assessment, the information does

~emonstrate the effectiveness of the volumetric leak path examination to detect low , Comment ICJI: Optional additional leakage rates, as performed in accordance with the licensee's proposed alternative. j t echnical basis for relief. Include if For hardship, the licensee noted that a qualified surface leak path assessment could be 'j desired. performed in two manners that would require both additional radiological dose and time  : versus the performance of a volumetric leak path assessment. The NRC staff found I both of these conditions to be of sufficient hardship given the operational experience and / technical adequacy of the licensee's proposed alternative versus the regulatory / requirementl *-------------------------------------------------------------------------------------------j Therefore, the NRC staff finds that the licensee's proposed alternative provides reasonable assurance of structural integrity until the next scheduled examination, and that compliance with the surface examination requirements of Paragraph -3200{b) of ASME Code Case N-729-1, for the subject welds, would result in hardship without a compensating increase in the level of quality and safety. Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation, I concur with the Component Performance, Non-Destructive Examination, and Testing Branch's determinations. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CROM penetration nozzles numbers 20, 27, 35, 40, 46, 47, 58, 59, 63, 70, 71 and 77 such that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(ii)(D). Therefore, the NRC staff authorizes the use of relief request 14R-03 at the Wolf Creek Generating Station during the current refueling outage subject to the licensee's proposed alternative that if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector. This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding Relief Request 14R-03, while preparing the subsequent written safety evaluation.

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELI EF REQUEST 14R-04 ALTERNATE EXAMINATION OF CONTROL ROD DRIVE MECHAN ISM NOZZLE PENETRATIO NS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated October 11, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-04 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration numbers 77 and 78 at the Wolf Creek Generating Station. The licensee proposed (a) an alternate examination distance for CROM nozzle numbers 77 and 78 in lieu of the required examination distance per ASME Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) and (b) not to perform the surface examination of the portion of the CROM nozzle below the J-groove weld as required by 10 CFR 50.55a(g)(6)(ii)(D)(3). The NRC staff finds that the proposed examination distance are acceptable for CROM nozzle numbers 77 and 78. This is based on the validity of the licensee's stress analysis and fracture mechanics calculation, demonstrating that within four refueling cycles, a potential flaw that initiates in the unexamined zone (below the J-groove weld) of the CROM nozzle numbers 77 and 78 will not propagate into the J-groove weld. At the end of every fourth refueling cycle, the licensee will perform an examination to confirm the structural integrity of CROM nozzles 77 and 78. The NRC staff finds the licensee's hardship justification is acceptable because of the considerable radiation dose and the nozzle configuration that are not conducive for the required examination. The NRC staff finds that the licensee's proposed alternative examination distances for CROM penetration nozzle numbers 77 and 78 provides reasonable assurance of structural integrity and leak tightness until the next scheduled examination, and that compliance with the surface examination requirements of 10 CFR 50.55a(g)(6)(ii)(D)(3) would result in hardship without a compensating increase in the level of quality and safety. Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch IV-1, Office of Nuclear Reactor Regulation, I concur with the Component Performance, Non-Destructive Examination, and Testing Branch's determinations.

The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CRDM penetration nozzles numbers 77 and 78 and that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(ii)(D). Therefore, the NRC staff authorizes the use of relief request 14R-04 at the Wolf Creek Generating Station for the remainder of the fourth 10-year ISi interval, which ends on September 2, 2025. All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third p*arty review by the Authorized Nuclear In-service Inspector. This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding Relief Request 14R-04, while preparing the subsequent written safety evaluation.

From: Collins, Jay Sent: 3 Nov 2016 22:10:48 -0400 To: Alley, David;Pascarelli, Robert Cc: Singal, Balwant;Cumblidge, Stephen;Caponiti, Kathleen

Subject:

RE: Request for additional Information, Wolf Creek, 14R-03 and 14R-03, Nozzle exam Attachme nts: Wolf Creek verbal auth 14R-03 11-03-2016 Rev 2.docx My apologies on the 14R-03 script, but I did not change the number of nozzles affected in the DORL section. I have now fixed that issue in red line strikeout. Attached is Rev 2 of that script. Jay From: Alley, David Se nt: Thursday, November 03, 2016 8:44 PM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov> Cc: Singal, Balwant <Balwant.Singal@nrc.gov>; Collins, Jay <Jay.Collins@nrc.gov>; Cumblidge, Stephen

<Stephen.Cumblidge@nrc.gov>; Caponiti, Kathleen <Kathleen.Caponiti@nrc.gov>

Subject:

Request for additional Information, Wolf Creek, 14R-03 and 14R-03, Nozzle exam

1. I concur with the attached RAI

2. I concur with both of the verbal scripts

3. Stephen, please draft a 665 for the RAI and send it to Kathleen

4. If we get the RAI sent out and a response back, I have no objection to doing the verbals on Friday. If we don't get a response back for the RAI, I propose we do the verbals on Monday.

I am open to doing them over the weekend if needed. If done tomorrow, Stephen is acting and will do my part. Dave David Alley PhD. Chief, Component Performance NOE and Testing Branch US Nuclear Regulatory Commission 11555 Rockville Pike Rockville MD 20852 301-415-2178

VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 14R-03 ALTERNATIVE TO USE VOLUMETRIC LEAK PATH FOR SUPPLEMENTAL EXAMS WOLF CREEK GENERATING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 Technical Evaluation read by David Alley, Chief of the Component Performance, Non-Destructive Examination, and Testing Branch, Office of Nuclear Reactor Regulation By letter dated November 1, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-03 for the alternate examination of all 78 control rod drive mechanism (CROM) nozzle penetration welds at the Wolf Creek Generating Station. The licensee proposed (a) to perform a volumetric leak path assessment of each penetration nozzle in lieu of the surface leak path assessment required by Paragraph - 3200(b) of ASME Code Case N-729-1, and (b) if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made this request in accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that while the demonstrated volumetric leak path is not equivalent to a fully qualified surface leak path assessment, the licensee identified sufficient operational experience, technical basis and radiological dose hardship to show that regulatory compliance would result in hardship without a compensating increase in the level of quality and safety. For operating experience, the licensee showed that there has been no previous identified cracking or leakage identified from the CRDM nozzle penetrations or welds of the upper head at Wolf Creek. The NRC staff noted that while this fact does not preclude the possibility of cracking to be found as the plant continues to age, plants which have previously identified cracking are more likely to see subsequent and more significant cracking in the future. Given the lack of the initial cracking being identified in the nozzle heats of material, at the operating temperatures of Wolf Creek, the NRC found that the potential for significant cracking this outage was less likely. For technical basis, the licensee identified that their inspection would be in compliance with the Wesdye Technical Justification Document showing an effective demonstration of the volumetric leak path technique. The NRC has accepted the use of a demonstrated volumetric leak path as part of the upper head inspection program under 10 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced NUREG/CR-7 142, Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation, which found, in part, the use of a properly focused 0 degree probe could detect a leakage path under low leakage rates during operation that led to minimal wastage of the upper head low alloy steel. While the NRC staff did not find that the volumetric leak path assessment was equivalent to a qualified surface leak path assessment, the information does

~emonstrate the effectiveness of the volumetric leak path examination to detect low , Comment ICJI: Optional additional leakage rates, as performed in accordance with the licensee's proposed alternative. j t echnical basis for relief. Include if For hardship, the licensee noted that a qualified surface leak path assessment could be 'j desired. performed in two manners that would require both additional radiological dose and time  : versus the performance of a volumetric leak path assessment. The NRC staff found I both of these conditions to be of sufficient hardship given the operational experience and / technical adequacy of the licensee's proposed alternative versus the regulatory / requirementl*-------------------------------------------------------------------------------------------j Therefore, the NRC staff finds that the licensee's proposed alternative provides reasonable assurance of structural integrity until the next scheduled examination, and that compliance with the surface examination requirements of Paragraph -3200{b) of ASME Code Case N-729-1, for the subject welds, would result in hardship without a compensating increase in the level of quality and safety. Authorization read by Robert Pascarelli, Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation, I concur with the Component Perfor mance, Non-Destructive Examination, and Testing Branch's determinations. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of li=le--all 78 CROM penetration nozzles numbers 20, 27, 35, 40, 46, 47, 50, 59, 63, 70, 71 and 77 such that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(ii)(D). Therefore, the NRC staff authorizes the use of relief request 14R-03 at the Wolf Creek Generating Station during the current refueling outage subject to the licensee's proposed alternative that if an unacceptable indication by the leak path assessment or volumetric exam is identified, the licensee will revert to the requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). All other requirements of ASME Code, Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector. This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding Relief Request 14R-03, while preparing the subsequent written safety evaluation.

From: Singal, Balwant Sent: 4 Nov 2016 16:03:29 -0400 To: Collins, Jay;Alley, David;Cumblidge, Stephen;Tsao, John Cc: Pascarelli, Robert

Subject:

FW: RAI Response from Wolf Creek (Relief Request 14R-03CAC No. MF8456) Attachments: ET 16-0031.pdf See listing of Records Already Ava ilable to the Public RAI response received from the licensee. The licensee indicated that they are proceeding with the volumetric examinations of the rest of the nozzles with the assumption that they are going to get the approval. They would like to know right away if we have additional questions or concerns. If we do not have any additional questions or concerns, they are ok with the verbal on Monday. Thanks. Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwa nt.Si nga l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 From: Muilenburg William T [11] Sent: Friday, November 04, 2016 3:40 PM To: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

[External_Sender] RAI Response from Wolf Creek

Balwant, Here is our RAI response.

I have verified that Monday morning is fine for a response, we don't need it over the weekend. If there are additional questions I am here Saturday and Sunday.

Thanks, Bill

From: noreply@nrc.gov Sent: 9 Nov 2016 10:56:56 -0500 To: Alley, David

Subject:

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By letter dated October 11, 2016, Wolf Creek Nuclear Operating Corporation (the licensee) submitted Relief Request 14R-04 for the alternate examination of control rod drive mechanism (CROM) nozzle penetration numbers 77 and 78 at the Wolf Creek Generating Station. In accordance with Nuclear Regulatory Commission's (NRC's) process as described in LIC-109, "ACCEPTANCE REVIEW PROCEDURES," the NRC staff has performed an acceptance review to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review was also intended to identify whether the request has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant. The NRC staff has concluded that the subject relief request does provide technical information in sufficient detail to enable the NRC staff to proceed with its detailed technical review and make an independent assessment regarding the acceptability of the proposed relief request in terms of regulatory requirements and the protection of public health and safety and the environment. If needed, the NRC staff may request for additional information to complete its technical review.

             . Relevant Pen       #!. .~.<?..r:!~.i.!.i.CJ~....l                                                                QC Level Il l Comments                                                                                                                      Brief Evaluat ion - See Evaluation document for more detail
             ! Yes ! No !                                                                                                                                                                                             !

• -----**--*-*r*****--***r***-*******r *---*------------*--------------*--*--*---------------*--*------------*----------*------**------------*----------**-------------------*---*------*-----------*----*------*TNoffilri9--em-anatin9--fro*m--itie--aii-nliili*s-re9ioii-*was-c.onfirme<r**----------*--*---------------------------*------------------*--------------

1 X 1 1 !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  : j Leak was repaired using a clamp assembly to preclude future degradation .

..............L............L...........J§l!9~-~.~!.n.2r..~.rx.!.'?..<?.~~-.1?.~~!~.u.!~!~..d.~~!!=:.d...~!.().U.~~-*n.2~!~......................................................l§!i:!-l.~!u.:.::!.!.!.n.~.~9.~!Y...!':-!.9.T~.~.P.!.().~.i~~-*-~Y...!~.~..:?.U.~::!.~.-~!-l-~~..f.'?.~..CJ.f.9.~!.::!.d.~t!2~.:--.............................--......

1 l !Nothing emanating from the annulus region was confirmed.

2 l X l l jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) l l l !Leak was repaired using a clamp assembly to preclude future degradation.

---

*------.l.-----------..!..............i.§1!9~-~~!.n.2r..~_rx_!22~~-.1?.~~i~.u.!~!~..d.~~!!=:.d...~!.()_U.~-~-*~2~1~-------------*---*----------*--*-------**----*------*.)_§!i:!-l.~!U..~::!.U.n.!~.9.~!Y....N..QT~-~.P.!.()_~-!~~--!?.Y...!~-~--:?.U.~9-~.-~!-l.S.!.f.?.~..CJ.f.9.~.r9.d.~!!.?.~.:.......................................

j  :  :  : Nothing emanating from the annulus region was confirmed.

             !      X        !              !                                                                                                                                                                        ! Leak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 3       :               1               l                                                                                                                                                                       !Leak was repaired using a clamp assembly to preclude future degradation .

..............l.............

                            .l..............:lgr.Y..!!?.9.~.~-£~.~~!~-~!.~.~.~.!?.~..Y.~.(.~.P.:~!.1.1.1..~.!~.~.!?.f..~!?.~.~.~~.....................................................................i§~!-l.~.~~.~9.!.!n~.~9.~~
                                                                                                                                                                                                                                                      ...N..9!..9.?.~.P.r.().~.~~~-*?.Y...!~.~..~.~r.!9.~..~~.~!.f.?.~..().f.~~r.9.~~!!.<?.~.:.................................--....

jNothing emanating from the annulus region was confirmed. 1 X j j !Leak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 4 l l l l Leak was repaired using a clamp assembly to preclude future degradation. ..............L.! ...........J.!.............J.~!~.?.!..~l)'.

                                                                 ..1.?..'?.~.~-.P..a.~~.U.!.a.!~..d..U.:?.~~.d...?.!.().U.~~-D.?.~!.~.................................................................... l§~.~!U..~9.!.!.n.!.~9.r.~!Y....N..9.T~.~-P.!.().~.!S.~...~Y...!~.~..:?.U.~9.~.!.!-l.S.!.f.<?.~..().f.~~!.9.d.~~.?.~.: .......................................
                                                                                                                                                                                                                     !Nothing emanating from the annulus region was confirmed.

5 1 X l 1 jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) l l l l Leak was repaired using a clamp assembly to preclude future degradation . ..............L.1 ..........J!.............l.§1!9~-~.~.i.n.2r..~.rx.P~~~~.u.!9.!~

                                                                                                      ..a.!.. ~.CJ.~!.~.............................................................................................. L§!i:!-l.~.~u..@.1.!.n.t.e.9~!Y....N..9T.<?.Cl.~.P.!.().~.!s.~..!?.Y...!~.~..:?.U.~9.~.!.!-l.S.!.f.<?.r.~..Cl.f.~~.r9.d.~!i.<?.~.:.......................................
                                                                                                                                                                                                                     !Nothing emanating from the annulus region was confirmed.

6 l X j l jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  :  : Leak was repaired using a clamp assembly to preclude future degradation .

..............L...........)..............1.P..~..P..<3.!.!!C?.u..1.a.t.~..Cl.n..Y~..s.!~~..2f..n.2~.2'.!~................................................................................................).§!i:!-1.~!u.:.9.!_!.n.!.e.9.~!Y....N..9.TC?.().~_P.!.().~.i~~-*?.Y...!!:!~..:?.U.~9.~..r!-l.S.!.f.<?.r.~..CJ.f.~~!.9.d.~!i.?.~.:.......................................

             !               j               j                                                                                                                                                                        !Nothing emanating from the annulus region was confirmed.

7

             ! x !                          !                                                                                                                                                                        ! Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

1 l !Leak was repaired using a clamp assembly to preclude future degradation .

..............L.: ...........Jj.............J~.!~.?.!..9.r.Y.

                                                                 ..P.~~!~_U.!9.t~ ..<?.~...':!.t!..S.!~e...CJ.f.~()~!e....................................................................................L§~.~.~-~U.!.9.1.!_n.~_e.9.~!Y....N..QT~-~-P.!.().~.!~~~--?.Y...!~.~--:?.U.~9-~..r!-l_S.t..f.<?.~..CJ.f.9.~!.9.d.~t!_<?.~.: .......................................
                                                                                                                                                                                                                      !Nothing emanating from the annulus region was confirmed.
             ! X             !              !                                                                                                                                                                        !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 8 1               !              1                                                                                                                                                                        ! Leak was repaired using a clam p assembly to preclude future degradation .

..............L.: ...........Ll ............l.P..~ l

                                                   ..P..~!.!!C?.U..1.~!~..?..t..n..?.~!~.........................................................................................................................!Nothing l§~!-l.~!U.!.9.!.!.n.~-~9.~!Y...!':-!.9.:r..~.~-P.!.().~.!~~~-*?.Y...!!:!~..:?.U.~9.~.-~!-l.S.~..f.<?.~..CJ.f.9.~.r9.d.~!!.'?.~.:.......................................

emanating from the annulus region was confirmed. 9 j X j j !Leak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) l l l !Leak was repaired using a clamp assembly to preclude future degradation.

---

*--*---_L ____________L ___________ .l.P..~..!().<?.:?.e...!?.<?.r.().n._.e~~~~-u.!~!~!..!?.<?.r.()_n,__s.!9.i~.~~-g--**--*--**--*---***-*--*----------------**-------*****-------***-*-----****------i§!i:!-l.~-~u..@.l.!.n.~.~9.r.~!Y....N..QT~-~-P.!.()_~_!s.~~--!?.Y...!~.~--:?.U.~9-~..r!-l_s.!.f.<?.~..CJ.f.9.~.r9.d.~!!.?.~.:.......................................

             !               !              !                                                                                                                                                                        l Nothing emanating from the annulus region was confirmed.

10 1 X l 1 jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

!  : jLeak was repaired using a clamp assembly to preclude future degradation .

..............L............!l..............l.P..~ l

                                                   ..1.CJ.CJ:?.e...P..?..i:!!C?.u..1.?..~e...Cl.n...~~..~!9.~.......................................................................................................!Nothing i§~.~!u.:.9.!.!.n.!.e.9.r.~!Y....N..QT.c::().~_P..r().~.!~~~--!?.Y...!~-~..:?.u.~9-~..r!-l.~!.f.C?.~..CJ.f.~~.r9.d.~!!.<?.~.:.......................................

emanating from the annulus region was confirmed. 11 X j 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

                             !              l                                                                                                                                                                         j Leak was repaired using a clamp assembly to preclude future degradation.

l  ! Minor dry loose particulate dusted around nozzle lStructural integrity NOT compromised by the surface rust form of degradation.

jNothing emanating from the annulus region was confirmed.

                                                                                                                                                                                                                        ~Leak source - crack in the Canopy Seal weld of Nozzle                                            77. (3142.1 (1}, and 3142.3 (a) 12              x                                                                                                                                                                                                   l Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............l..............l.1?..ry..P..?.!.!!0.!.1.?.t.~..<?.n...n.o.~.Z.!~........................................................................................................................l§!~.~!U.!.~.1.!.n.!.~9.~o/....N..9.!...~.".1.P..r~rri.~s.~..!?.Y...t~.~--S..U.r:f~.~--ru..s.IJ0.~".1..<?.f.9.~W~.d.a.!!.O.r:'.:.......................................

             !              !               !                                                                                                                                                                           !Nothing emanating from the annulus region was confirmed.

1 X 1 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 13

             !              !               !                                                                                                                                                                           l Leak was repaired using a clamp assembly to preclude future degradation.

..............L.! ...........L.! ...........!1.1?..ry__!<?.O.S..~..P..?..'!!~u..1.?.t.~..<?.n...Y.r.!..s.!9.~..o.f..11.0.~!~..................................................................................!Nothing i§!!.u..~!u..r.~.!.!.ri~.~9.!.~o/ ..N..9.!..9.?.".1.P.!.~".1.~S.~..!?.Y...!!2~..s..u.r:t~.~--r.u..s.!.fo.~..'?.f.9.~!.~.d.a.!!.o.r:i.:....................................... emanating from the annulus region was confirmed. 1 X 1 1 jleak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 14

             !              :               l                                                                                                                                                                           !Leak was repaired using a clamp assembly to preclude future degradation .

..............l..............!..............i.1?..r:Y...!'?.?.S..~..P..?..'!!.~u..1.?..~~..<?.n...Y.r.!..S.!9.~..0.f..11.o.~!~..................................................................................i.§!!.U..~.~U..r.9.!.!.ri~.~9.~o/..N..9.!...~.".1.P.!.~".1.~S.~9..!?.Y...!~.~--S..U.r:t9.~..r.u..s.!.f.O.~..'?.f.9.~!.9.d.~!!.O.r:'.:.......................................

!  ! jNothing emanating from the annulus region was confirmed.

15 i X i i !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

             !              :               l                                                                                                                                                                           jLeak was repaired using a clamp assembly to preclude future degradation .

..............l.............J.............l.P. r:Y...!?..O.~.~--P.<l.~!~U..1.<J!~..?..n...Y.r.!..S.!.d.~..O.f..11.0.~!~..................................................................................L~!!.U..~.~U..r.9.1.!.n.~.~9.!.~o/..N..9!...~?..".1.P.!.~".1.!S.~..!?.Y...!~.~--S..U.!.f9.~..ru..S.!.f.O.~.gf.~~!.9.d.~!!.O.r:'.:.......................................

             !              :               !                                                                                                                                                                           !Nothing emanating from the annulus region was confirmed.

1 X 1 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1)._and 3142.3 (a) 16

             !              i               i                                                                                                                                                                           :Leak was repaired using a clamp assembly to preclude future degradation .

..............l..............J..............l.1?..r:Y...P..<l!.!!~.U.!.?.t.~..............................................................................................................................................l§!!.U..~.~U.!.9.!.!.ri~.~9.!.~o/..N..9.!..~?..".1.P.!.~".1.~S.~9..!?.Y...!~.~--s..u.r:t9.~..r.u..s.!.f.O.~..<?.f.9.~!.9.d.~!!.O.r:'.:.......................................

             !              j              !                                                                                                                                                                            ! Nothing emanating from the annulus region was confirmed.

1 X 1 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 17 i  ;  ;  ; Leak was repaired using a clamp assembly to preclude future degradation. .............l..............i..............i.l3.<?.ro.11..9.~~-u..rri.u.1.a.ti.o..n...o.r:i..~.l:l..S.i9.~..<?r~?..~.1.~....................................................................................l§!.~.~-~u..r9.!.!.n.t_~9.~o/..N..9.!...~.rrl.P.!.?..rrl.iS.~..~.Y...t~.~--S..U.r:f9.~..ru..s.!Jo.~..<?.f.9.~!.9.d.~ti.O.r:'.:......................................

             !              i               !                                                                                                                                                                           lNothing emanating from the annulus region was confirmed.

18 l X l l jleak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

             !              :               !                                                                                                                                                                           ! Leak was repaired using a clamp assembly to preclude future degradation.

..............L............L............l.1?..r:Y...P..?.!.!!~U.1.?.t.~..<?.n...Y.r.!..S.!9.~..0.f..r1.0.~.Z.!~................................................................................................).§!!.u..~!U..r.~.!.!.r1~.~9.~o/.~.9.!..9.?.".1.P.!.~".1.iS.~..!?.Y...!~.~--S..U.r:f~.~--ru..S.~..f.O.~..'?.f.9.~!.~.d.<l!i.O.r:'.:.......................................

:  : :Nothing emanating from the annulus region was confirmed.

19

             !      X l                    !                                                                                                                                                                            \eak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a)
             !              !               l                                                                                                                                                                           ! Leak was repaired using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • l**************l**************l*~ir:'.?.!..9.~..P.~~!~.U.!~t~..o.r:i...~.l:l..s.!9.~..9.f..ri.o.~1.~...................................................................................t§!!.U..~.~U..r.9.!.!.ri!.~9.~o/..N..9.!...~'?.rrl.P.r..0.".1.~S.~9...~Y...!~.~--s..u.r:t9.~.!.U..S.!.f.O.~..'?.f.9.~!.~.d.~!!.o.r:i.:.......................................

i i j j Unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3 130, and 3 132.1 (a)) 20 ! X  !  ! jSupplemental Volumetric Examination performed identifying no change in structural i i i !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013. ..............l..............i..............l.l?..r:Y...!?..o.~.~--P..?..'!!~u..1.?.!~..<?.n...Y.r.!..s.!9.~..o.f..11.0.~.z.!~:..~.u.s..~..o.~Pl:l..s.!~.~--~f..n..o.~!~.:.....................lN..o..d.~.@.d.a.~o.r:i..~.?.~..!9.~.11~.i.~~--!~a..!..~.u.!.d...~!!.1.P.!.O.~!s..~..s.!r.~~-!u..r.a.!..i.11!~.!i!!.!t>.'.:...................................................

:  ! !Nothing emanating from the annulus region was confirmed.

21 i X i i !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

             !              !               !                                                                                                                                                                           !Leak was repaired using a clamp assembly to preclude future degradation .

..............l..............L............L!?..r:Y...!'?.?.~.~--P.<l.'!!~u..1.?..~~..<?.n...Y.r.!..S.!9.~..0.!..11.o.~!~..................................................................................L§!!.U..~.~U..@.U.n.~.~9.!.~~---N..9.!..9.?.rT1.P.!.~rrl.~S.~9..!?.Y...!~.~--S..U.r:f9.~..ru..s.!.f.O.!.rT1..'?.f.9.~!.9.d.~!!.O.r:'.:.......................................

             !              !              !                                                                                                                                                                            !Nothing emanating from the annulus region was confirmed.

22 ; l X l j jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) i  ; ;Leak was repaired using a clamp assembly to preclude future degradation . ..............!..............!..............l.l?..r:Y...1.o..O.~.~--P.<l.'!!~U..1.a..t.~..'?.11..Y.r.!..S.!9.~.......................................................................................................L§!!.u..~.~U.!.9.!.!.n.~.~9.!.~o/..N..9.!...~?..".1.P..r.~rri.~S.~9...!?.Y...!~.~--S..U.r:f9.~..r.U..S.!.f.O.~..'?.f.d.~!.9.d.~!!.O.r:'.:.......................................

             !              !               l                                                                                                                                                                           !Nothing emanating from the annulus region was confirmed.

23

             !      X       !              !                                                                                                                                                                            jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)
             !              !               !                                                                                                                                                                           !Leak was repaired using a clamp assembly to preclude Mure degradation.
             !              !               !Dry loose particulate on UH side                                                                                                                                           l Structural integrity NOT compromised by the surface rust form of degradation.

jNothing emanating from the annulus region was confirmed.

                                                                                                                                                                                                                     ~Leak source - crack in the Canopy Seal weld of Nozzle                                               77. (3142.1 (1}, and 3142.3 (a) 24               x                                                                                                                                                                                               l Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............l..............l.1?..ry..!C?.O.~.~--P..Cl.~!.c.u..1.a.!~..C?.ri..lJ.lj. s.!9.~..0.!..ri.o.:z.z.1~..................................................................................i§!~.~!U.!.Cl.l.!.ri!.~9.~o/....N..9.!...~.".1.P..r!?.".1.~5.~..?.Y...t~.~--~-u.r:ta..~..ru..s.t..f.0.~".1..C?.f.9.~WCl.d.~!!.O.r:'.:.......................................

              !               !              !                                                                                                                                                                       !Nothing emanating from the annulus region was confirmed.

1 X 1 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 25 :  !  : l Leak was repaired using a clamp assembly to preclude future degradation. ..............ii..............ii..............il.§!!9.~Yn.i.i.ri.o.!..9.~.P.~~~-u.!a..!~..o.r:i..~9..9.~E.~!7..Cl.r19...~~--~!9.~...o.f..ri.o.~!~.....................................:Nothing i§t_ru..~!u..~a..!.!.ri~.~9.!.~o/ ..N..9.!..!?.C?.".1.P.!.!?.rl'.1.~~~--?.Y...!!2~..~.u.r:ta..~..~U..~!.fo.~..C?.f.9.~!.Cl.d.~!!.o.r:i.:....................................... emanating from the annulus region was confirmed. 1 X 1 1 \eak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 26

! l !Leak was repaired using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • !**************!**************!.1?..':Y...P..Cl~!~u..1.a.t.~..C?.rl..lJ.lj..s.!9.!7-....................................................................................................................j-§t_ru..~!U..~Cl.!.!.ri~.~9.~o/..N..9.!...~.".1.P.!.!?.".1.!~~9..?.Y...!~.~--~-u.r:ta..~..~U..~!.f.O.~..C?.f.9.~!.Cl.d.~!!.O.r:'.:.......................................

              !               !              !                                                                                                                                                                       ! Unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a))

27 i X i l l Supplemental Volumetric Examination performed identifying no change in structural

             !               !               !                                                                                                                                                                       !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

..............L......-.....t.............L!?..':Y...~~.o..ri_o.r:i..~.~---~i.ci..~..o.~.r:i.C?.z.:z!.~............................................................................-............................1~.o..d.~!.a..d.~~.o.r:i..Y.!'.a.~.!~~-ri~.i.~~--!~a..~..~.u.!.d...c.~!!.1.P..r..o.~!~.~--s.~.~~~!u..~~~..i.ri!~.9!.!t>.'.:.............................................-....

:  : l Nothing emanating from the annulus region was confirmed.

28 ! i X i j jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

                              !              !                                                                                                                                                                       ! Leak was repaired using a clamp assembly to preclude future degradation.

..............L.: ...........J.!.............l.1?..':Y...l.C?.O.~.~--P..Cl.~!~u..1.a.~.~..C?.rl..lJ.lj__s.!9.~..0.!..ri.o.:z:z.!~..................................................................................i§~.~-~U..~Cl.!.!.ri!.~9.!.!o/ ..N..9.!..!?.C?.".1.P.!.!?.".1.~~~---~Y...!~.~--~-u.r:ta..~.!.U..~!..f.O.~..C?.f.9.~!.Cl.d.~!!.O.r:'.:.......................................

!Nothing emanating from the annulus region was confirmed.

29 i X i i iLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

              !               l              l                                                                                                                                                                       l Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............i..............i!?..':Y...~C?.o.~.~--r..a..~!~u..1.a..t.~..C?.ri..lJ.lj__~ici.~..o.!..ri.o.:z:z.1~..................................................................................i§t_ru..~.~u..@.'.!.ri!.~9.~o/..N..9.!...~.n.i.P.!.!?.rl'.1.!~~--?.Y...!~.~--~-u.r:ta..~..~u..~!.f.o.r.rl'.1..C?.f.9.~!.a..d.~!!.o.r:i.:.......................................

! l !Nothing emanating from the annulus region was confirmed.

1 X 1 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 30

: l :Leak was repaired using a clamp assembly to preclude future degradation.

..............L...........J............J§1!9.~.~.".1!.r1?.r..9.~.!.°.?.~~-.P.~~!~.U.1~!~..?.r:1..~.~--~!9.El..CJ.f..r1?.~.~El........................................................i§t_ru..~.~U..r~_l.!.r1!.El9.~o/....N..9.!..9.().".1.P.!.()fl'.1.i~~--~Y...!~.El..~.U.r:f~.~--~U..~!.f.°.~..().f_9.~!.~.d.~t!?.r:'.:.......................................

              !               l              !                                                                                                                                                                       !Nothing emanating from the annulus region was confirmed.

31 l X l j \eak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a)

:  : !Leak was repaired using a clamp assembly to preclude future degradation .

..............1..............J..............l.1?.r:x..!C?.o.~.~--P..a..~!.c.u..1.a..~~--C?.ri..lJ.lj..~!.d.~..C?.!..ri.o.:z:z.!~..................................................................................L§t_ru..~!u.!.a..!.!.ri~.~9.!.~o/ ..N..9.!..!?.C?.n.i.P.!.!?.rl'.1.!~~-*9*Y...!~.El..~.u.r:ta..~.!.u..~!.f.o.~..C?.f.9.~!.a..d.~!!.o.r:i.:.......................................

: l l Nothing emanating from the annulus region was confirmed.

32 i X i i !Leak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

              !               !              !                                                                                                                                                                       !Leak was repaired using a clamp assembly to preclude future degradation .

..............l..............L............l.1?..':Y...!C?.O.~.~--P..Cl.~!~U..1.a.!~..C?.ri..lJ.lj..~!9.~..C?.!..ri.o.:z:z.!~..................................................................................l§t_ru..~!U.!.Cl.!.!.ri~.~9.~o/.~.9.!..~C?.rl'.1.P.!.!?.rl'.1.!~~d...?.Y...!!2El..~.u.r:ta..~.!.U..~!.f.o.~..C?.f.9.~!.Cl.d.~!!.C?.r:i.:.......................................

!  : !Nothing emanating from the annulus region was confirmed.

33 i X  ! i !Leak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

!  : !Leak was repaired using a clamp assembly to preclude future degradation .

..............1..............i..............i!?..':Y...!C?.o.~.~--P..a..~!~u..1.a..~~--C?.ri..lJ.lj__~!d.~..o.!..ri.o.:z:z.!~!..'.".~~u..n.i.~~-a..ri.riu..!u..~..c.!~a.!............................i§t_ru..~.~u..@.1.!.r:!~.~9.r.!o/....N..9.!..!?.C?.n.i.P.!.!?.ri:i.!~~d...?.Y...!~.El..~.u.r:ta..~..ru..s.!.f.o.r.ri:i..C?.f.9.~!.a..d.~!!.o.r:i.:.......................................

              !               !              !                                                                                                                                                                       ! Nothing emanating from the annulus region was confirmed.

34 i l X l j jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) i i iLeak was repaired using a clamp assembly to preclude future degradation. i i iSlighUminor dry particulate on UH side of nozzle !structural integrity NOT compromised by the surface rust form of degradation.

                                                                                                                                                                                                                ~ Unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a))

35 : X  :  :  : Supplemental Volumetric Examination performed identifying no change in structural

                             !             !                                                                                                                                                                    !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

..............l..............L............lPr.Y...!?..o.~.e...P..a..r:!!~u..1.a..~e...?..~--~-°-~-~ !~..........................................................................................................l~?..9.~.@.~~~?.r.i..~.a.~..i9.e..~!.i.~~..!~a..! ..~.u.!.?...~~P..r?.~!~.e...s.!.~~~!u..~~!.!.~!e..9r.!t>.'.:...................................................

: j :Nothing emanating from the annulus region was confirmed.

1 X 1 1 jLeak source - crack i~ the Canopy Seal weld of Nozzle 77. (3142.1(1),_and3142.3 (a) 36

             !              !              !                                                                                                                                                                    !Leak was repaired using a clamp assembly to preclude future degradation.

..............L ...........Ll ............lPr.Y. ..P..a.~!~u..1.a.t.e...?..n...Y~..s.!9.~.?.f..e~r.i.e.!~~!i?.r.i......................................................................................l§!.~u..~!u..~a..!.!.n.!.e.9.~~-*~.9.!..~.~-P.r.?.~.is.~9..~Y...!~.e...~.u.:.ta..~ ..~u..s.!.f.o.~..9.f.~~r.a..~~!!?.r.i.:.......................................

: !Nothing emanating from the annulus region was confirmed.

1 X 1 j j Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1). and 3142.3 (a) 37

:  : :Leak was repaired using a clamp assembly to preclude future degradation .

..............i..............i..............i.§1!9.~Yr:r.i_i_~?.r..9.~.!?.?.:!~.P.~~!~.U.!~!~.?..n...LJ.f:!..~!9.6...9.f..rl?.~.!e.........................................................~§~U..~!U..ra._l_!.ri!.6.9.~~---N..9.!..~.~.P.r.?.~-~5.~9..~Y...!~.e...~.U.:.fa..~..ru..5.!.f.O.r.~..9.f.~~r.a..~~!!?.!:i.:.......................................

             !              !              !                                                                                                                                                                    !Nothing emanating from the annulus region was confirmed.

38 : l X l l jLeak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  : Leak was repaired using a clamp assembly to preclude future degradation.

..............L.............t.............L!?..r.Y...!?..o.~.e...P..a..r:!!~u.!.a.t.e...?..~..Y.~..:!!9.~..o.!..~.o.~.~!~..................................................................................l§~u.-~!u..~a..!.!.ri!.e.9.r.~~...N..9!..~-~-P.r.?.~.~s.~..~Y...!~.e...~.u.:.ta..~..~u..:!!.f.o.~.-9.f_~~r.a..~~!!?.~:.......................................

:  : j Nothing emanating from the annulus region was confirmed.

             !      X        !             !                                                                                                                                                                    !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 39 :                    :              :                                                                                                                                                                    : Leak was repaired using a clamp assembly to preclude future degradation.

. . . . . ..+. . . . . . j..............j.§!!9.~-~-r:r.i!.~?.r..~.~-P~~~.U.!~!~..o.r.i..Yf:!..~!9.e...?..f.!:i.?..~.1.e......................................................................t§~-~-~u..~a..!.!.n.!.e.9.r.~~...N..9.!..~.~-P.r.?.~.~:!~...~Y...!~.e...~.u.:.ta..~.!.U..5.!..f.0.~..9.f.~~r.a..~~!!?.!:i.:.......................................

l  : :unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a))

40 ! X l l !supplemental Volumetric Examination performed identifying no change in structural

             !               !             !                                                                                                                                                                    kharacteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

..............i..............i..............!P.r.Y...P..a.~!~u..1.a.t.e...?..n...~~..s.!~e...?.f..n.?.~.~!e.................................................................................................l.~?..9.~.@.~~~?.!:i..~.a.:?..!9.e..~t.i_~~..t~a..t..~_u..1.~..C.9.~.P.!.?.~!~.e...s.!.~~-tu..~~!..i.~!e..9r.it>.'.:..................................................

!  :  : Nothing emanating from the annulus region was confirmed.

41 l X l l  ! Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) i i i l Leak was repaired using a clamp assembly to preclude future degradation . ..............L...........!..............1P.r.Y...!?..o.~.e...P..a..r:!!~u.1_a.!e...?..~..Y.~..s.!9.~..o.t..~.o.~.~!~..................................................................................l§!.~-~-~u..ra._1_!_~t.e.9 ~o/..N..9.!..~.~.P.r.?.~.!s.~9..?.Y...t.h_e.__~_u.:.ta..~..~u..s.!.f.o.~..9.f.~~r.a..~~ti_o.r.i.:.......................................

             !               l             l                                                                                                                                                                    !Nothing emanating from the annulus region was confirmed.
             !      X        !             !                                                                                                                                                                    !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1),_and 3142.3 (a) 42

:  : :Leak was repaired using a clamp assembly to preclude future degradation.

! iDry particulate at nozzle interface (small amount) !Structural integrity NOT compromised by the surface rust form of degradation.

• • • • • • • • • • • • • r***********r***********r********************..***************************************..****************************************************************************************************..******TNoi'tiiri9"eiiiinaiin9..trom..itie**a-;;*r;uiu*;;*;:e-91c;;;**w-as*c:*0r;ii;:me-a*:*****************..****************..***************************************....

43 l X l l !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

: j  : Leak was repaired using a clamp assembly to preclude future degradation.

..............L...........L...........iP.r.Y...!?..O.~.e...P..a..r:!!~u..1.a.t.e...9.~..\!.!j_,5.!.~~..0.f..~.O.~.~!~..................................................................................l§~.~-~u..ra..l.!.~!.e.9.~~---N..9.!...~.~-P.r.?.~.~5.~...~Y...!~.e...~.u.:.ta..~..ru..s.!.f.0.~..9.f.~~r.a..~~!!?.!:i.:....................................... 1  ! 1  ! Nothing emanating from the annulus region was confirmed. 1 X 1 1 jLeak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 44

j  : !Leak was repaired using a clamp assembly to preclude future degradation .

..............i..............i..............i.!?..r.Y...!?..0.~.6...P..a..r:!!~u..1.a.t.e...9.~..Y.~..5.!9.~..0.f..~.O.~!~................................................................................)§!!Y~!U..~a..!.!.~!.e.9.r.~o/..N..9.!...~.~.P.r.?.~.iS.~9..~Y...!~.e...~.U.:.fa..~.!.U..5.!.f.0.~..9.f.~~!.a..~~!!?.!:i.:.......................................

             !              !              !                                                                                                                                                                   !Nothing emanating from the annulus region was confirmed.

45

             ! X !                         !                                                                                                                                                                    !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  : iLeak was repaired using a clamp assembly to preclude future degradation.

l iSlighUminor dry particulate on 90 degree and UH side of nozzle )Structural integrity NOT compromised by the surface rust form of degradation.

                                                                                                                                                                                                                     ~ Unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a))

46 : X  :  :  : Supplemental Volumetric Examination performed identifying no change in structural

                             !             !                                                                                                                                                                         !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

..............1..............!-*************!.1?.~..!?..0.:>.~..P..a..r:!!~u..1.a..~~--?..n...Y.tt..~!9.!::..0.f..n..o.~!!::..................................................................................t~9..9.~.@.~~~9.r.i..~.a.:>.. i9.~Dt_i_fi.~..!~a..! ..~.U.!.?...~~P.E9.~!:>.~..~~-~~~!U..~~!.!.n.!~.9.!.!t>.'.:...................................................

             !               !             !                                                                                                                                                                         ! Unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a))

47 : X  :  : :supplemental Volumetric Examination performed identifying no change in structural

             !              !              !                                                                                                                                                                         !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

..............l..............l.............J§igr.i_~~-a.~!..~.0.!.0.!:1...~.r.i.?...~l:!:>.~.-~!C?.0..~........................................................................................................1~9..9.~.~a..~~~9.r.i..~.a.:>..!9.~.n.~.i.fi.~..!~a..!..~.U..1.?...~.~.P..r9.~!:>.~..~!~~!U..~~U.n.!~.9.!.!t>.'.:...................................................

: l l Nothing emanating from the annulus region was confirmed.

[ X [ [ [Leak source -crack i~ the Canopy Seal weld of Nozzle 77. (3142.1(1).,and3142.3 (a) 48

              !              !             !                                                                                                                                                                         ;Leak was repaired using a clamp assembly to preclude future degradation .

..............L: ............L: ............i.§1!9.~.¥..~!.n.9.:..a..~c:>.u..n.! l

                                                                                    ..?.r9_ry:_J?.a..~i-~l:!!a..!~...o.r.i..Y..~.:>.!9.E'...o.!.r.i.o..~.1.E'............................................... L§~u..~!u..~a..!.!.n.!.E'.9.~o/....t:-!.9.!..~.~.P.!.9.~.i.~~--!?.Y...!~.~--:>.u.!:fa..~..~u..~!.f.o.~..c:>.f.9.~!.il.d.~!!?.r.i.: .......................................

lNothing emanating from the annulus region was confirmed. 49 f X f f l Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  : :Leak was repaired using a clamp assembly to preclude future degradation .

..............L...........L...........L§!!9.~.¥.!!!.iD9.!..~~?..u..n.!..?.r9_ry:_!9.9.~!::..P.~~!~l:!!~!!::.9.r.i..Y..~.~!9.~..?..f.!:1.0..~.1.~.................................l§~u..~!U.!.il.!.!.n.!.E'.9.!.!o/....t:-!.9.!...~.~.P.!.9.~.!~!::9..!?.Y...!~.~..:>.u.!:fa..~..~u..~!.f.o.~..c:>.f.9.~!.il.d.~~9.r.i.:....................................... i i i i Nothing emanating from the annulus region was confirmed. 50 [ X  ! i i Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  :  : Leak was repaired using a clamp assembly to preclude future degradation.

..............L! ...........J! .............l_§l!9.~.¥.!!!.i_n.9.!..~~?..U..n.!

                                                                                    ..0.!.9.ry:_!9.<?.~~-P.a.:~!~l:!!~!~.?r.i..Y..~.~!cj_~_g_f.r.i.0..~.1.~.................................i§~U..~.~U.!.il.l.!.n.!.~9.!.!o/....t:-!.9.!..~.~.P.!.<?.~.!:>~cj__ !?.Y...!~-~..:>.U.!:fil.~..ru..~!.f.<?.~..c:>.f.9.~!.?..d.~!!.<?.r.i.: .......................................
                                                                                                                                                                                                                     !Nothing emanating from the annulus region was confirmed.

51

             ! X i                         [                                                                                                                                                                         iLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)

:  : l Leak was repaired using a clamp assembly to preclude future degradation.

..............i..............i..............).1?..~..!?..0.:>.~..P..a..r:!!~U..1.a..~~..?..n...Y.tl..~!.d.~..<?.f..n..<?.~!~..................................................................................!.§~U..~.~U..ra..!.!.n.!.~9.~o/....t:-!.9.!..~?..~.P.!.<?.~.i~~-*!?.Y...!~.~..:>.U.!:fil.~..~U..~!J<?.!.~..C?.f.9.~!.il.d.~!i?r.i.:.......................................

:  :  : Nothing emanating from the annulus region was confirmed.

52

             !      X        !             !                                                                                                                                                                         !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) i              i             i                                                                                                                                                                         l Leak was repaired using a clamp assembly to preclude future degradation .

..............L...........L............1.1?..~..~.o.o.~!::.P.a..~i~.u.!il.t!::..o.r.i..Y..~.~!9.~......................................................................................................£§!.~.~-~u..ril.1.!_n.t_E'.9.~o/..t:-!.9.!..~.~.P.!.9.~.!~!::9..!?.Y...t_h_~..:>.u.!:fil.~..~u..~Uo.~..c:i.f.9.~.ra..d.~ti_o.r.i.:.......................................

:  : :Nothing emanating from the annulus region was confirmed.

[ X [ [ !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1)., and 3142.3 (a) 53

:  : :Leak was repaired using a clamp assembly to preclude future degradation.

i i i Dry loose particulate on UH side of penetration 1Structural integrity NOT compromised by the surface rust form of degradation. .............T.............1"...........T.............................................................................................................................................................................fNoi'tiiri9"eiiiinaiiri9"trom..itie"aii*nuiu*;;*;:e-910;;..w-as*c:*c;r;ii;:me-cf................................................................................

              !     X        !             !                                                                                                                                                                         jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) 54

1  :  : Leak was repaired using a clamp assembly to preclude future degradation.

..............L............L............i.l?..~..~:>.!..~_lg_r~.d...~9.!.0.~.~!..r.i.O.~!.E'...!C?..~~-a.9..!.n.!!::!:f?..~...................................................................l§~.~-~U..ril.l.!.n.!.E'.9.~o/....t:-!.9.!...~.~.P.!.<?.~.~~~...~Y...!~.~..:>.U.!:fil.~..ru..~!.f.O.~..C?.f.9.~!.il.d.~!!9.r.i.:.......................................

              !              !             !                                                                                                                                                                         ! Nothing emanating from the annulus region was confirmed.

[ X ! [ iLeak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 55

1  : 1Leak was repaired using a clamp assembly to preclude future degradation .

.............L. ..........L............i.g_~..!?..o.:>.~..P..a..r:!!~u..1.a.t.E'...?..n...Y.tt..~!9.!::..o.f..n..o.~!!::..................................................................................!.§~u..~!u..~il.!.!.n.!.E'.9.:.!o/....t:-!.9.!...~.~.rr.9.~.i~!::9..!?.Y...!~.~..:>.u.!:fil.~..ru..~!.f.o.~..c:i.f.9.~.ra..d.~!!?.r.i.:.......................................

             !              !              !                                                                                                                                                                         !Nothing emanating from the annulus region was confirmed.

56

              !     X       !              !                                                                                                                                                                         ! Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1). and 3142.3 (a)

:  : iLeak was repaired using a clamp assembly to preclude future degradation.

i 1 i Dry particulate on UH side of nozzle )Structural integrity NOT compromised by the surface rust form of degradation.

jNothing emanating from the annulus region was confirmed.

                                                                                                                                                                                                                     ~Leak source - crack in the Canopy Seal weld of Nozzle                                           77. (3142.1 (1 }, and 3142.3 (a) 57              x                                                                                                                                                                                                l Leak was repaired using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • l**************l**************l.1?..ry__P..?.!.!!0.!.1.?.t.e...<?.n..Yf:i..~!9.~:..'.".11D.~!.~~--~1.e.9E....................................................................................r?.!~.~!~!.9.l.!.n.!.e.9.~o/....N..9.:r...~.n.i.P..r~n.i.~~~--?.Y...t~.~--~-~r:t9.~..r~-~t..f.<?.~n.i..<?.f.9.~w~.d.~!!.<?.11.:.......................................

             !              !                !                                                                                                                                                                       ! Unable to visually confirm no nozzle leakage (3142.1 (b} and (c), 3142.2, 3130, and 3132.1 (a))

58 ! X  !  ! !Supplemental Volumetric Examination performed identifying no change in structural

: l [characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

• • • • • • • • • • • • • • l**************l**************l*§<?.~.O.!":l..~.~~-~-".1.~!.?.!!.0..ll...O.ll..ll.'?.2'.~.1.e..:..~.o..n.i.e...d.!.~~.0.1.0.~.?.!i.o..n.:.............................................................y:!.o..9.~!.9.d.~~.O.ll..~.?.~.!~.~-n.~.i.fi.~..t~~-~ --~-~-'-d...~ll:'.P.!..O.~!~.e...~!~~~-~~-~~!..i.n.!e..9!.!t.Y.:...................................................

             !              !                !                                                                                                                                                                       ! Unable to visually confirm no nozzle leakage (3142.1 (b} and (c), 3142.2, 3130, and 3132.1 (a))

59 ! X  !  !  ! Supplemental Volumetric Examination performed identifying no change in structural l l lBoron accumulation (loose particulate and hard caked boron) on UH side of l characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013. ..............1..............1..............l.e.e.11.~!!.~.!!.<?.11.:.~.':!~!.?.11.g_~..Ll?..<?.~.ll~.~-!!!.l ..~.i.?.~.:.........................................................................................L~.0..9.~!.9.d.~~.O.ll..~.?.~.!9.e..n.~.i.~~--!~~-~ --~-~!.d...~9.ll:'.P.!..O.~!~.e...~!~~-~~-~~!..i.n.!e..9!.!t.Y.:...................................................

: l lNothing emanating from the annulus region was confirmed.

60 : f X f f lLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) l l [Leak was repaired using a clamp assembly to preclude future degradation . ..............L............L............lY~~-~!ig_~Y~!.n..o.~..9.~.P.9.~g.l:J.!~.~~...O.ll..Y.~..~~d..e...<?!.P..e..n.~!!.?.~.o..n.................................................l:?.!!.~.~!~!.9.!.!.n.!.e.9.!.~o/..N..9.!...~.".1.P.!.~ll:'.~~~d...!?.Y...!~.~--~-~r:t9.~..~~-~!.f.<?.~..9.f.9.~!.~.d.~~.o.11.:....................................... l  !  :  : Nothing emanating from the annulus region was confirmed.

             !      X       !                i                                                                                                                                                                       !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1). and 3142.3 (a) 61       :              l                :                                                                                                                                                                       l Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............l..............l.§1!9.~.~~-i_n..o.r..~~'?.~.n.!..<?!.!0.?.~~--f?.~.r:!.i.~~.!~.~~...O.ll..Y.i::t..~.i.d..e...<?f..ll.<?!.<:!.e..........................................l:?.!!.~.~-~~!.9.1.!.n.!.e.9.!.~o/..N..9.!..0.?.".1.P.!.~ll:'.~~~d...?.Y...!~.~--~-~r:t9.~..r~-~!.f.<?.~..<?.f.9.~!.9.d.~!!.<?.ll.:.......................................

             !              !                !                                                                                                                                                                       !Nothing emanating from the annulus region was confirmed.
             !      X       !                !                                                                                                                                                                       !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 62

!  : l Leak was repaired using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • !***..*********!**************j.l?..r:Y...!'?.0.~.9...P..?..r:!!<?.':!.1.?..~9...<?.n...Y.f:i..~!.d.~..<?.~..ll.<?.~!~..................................................................................j.:?.!!.~.~-~~-r9_!.!.n.!.9.9.~o/..N..9.!..<?.'?.".1.P.!.~ll:'.i~~--?.Y...!~.~--~-':!r:f9.~..'.~.~!..f.<?.!.ll:'..9.f.9.~!.9.d.~!i.O.ll.:.......................................

             !              !                !                                                                                                                                                                       !Unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a))

63 i X i i l Supplemental Volumetric Examination performed identifying no change in structural

             !              !                [Dry rust colored boron cake on 90 side of nozzle - Dry loose particulate on UH !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

..............l..............L...........J~i9.~..0.f..n..0.?..2'.!~...............................................................................................................................................i.~.0..9.~.~.d.~~.O.ll..~.?.~..!~.e._n.t_i_fi.~..~~~.t..~.':!!.d...~ll:'.P.!..O.~i~.9...~~-r~~!~.r~!..i.11!9..9!.it>.'.:...................................................

l l [Nothing emanating from the annulus region was confirmed.

[ X  !  ! [Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 64

!  : l Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............L............l.1?..r:x..P..?.!.!!~.~-1.?.t.e...<?.n...Yf:i..~!9.~...o.!..e~11.e.!!.~!i.o.11......................................................................................i:?.~.'.~.~-~-'.9.1.!.n.!.e.9.~o/..N..9.!..0.?.n.i.P!.<?.n.:i.~~~---~Y...!~.~--~-':!r:t9.~..'.~.~~--f.<?.~..9.t.9.~!.9.d.~!i.o.11.:.......................................

!  : l Nothing emanating from the annulus region was confirmed.

65

             !      X       !                 j                                                                                                                                                                      i Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1),_and 3142.3 (a)
             !              !                !                                                                                                                                                                       ! Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............l..............l.§!!9.~.~.".1.i.n..o.!..i:i.~<?.~.n.!..<?!.d..~.P.~.~.i.~~!9.!~...o.11..Y.~.~-~d..e...'?!.11.o..~.1.e...............................................L§!!.~.~!~.~9.!.!.n.!.e.9.~o/..N..9.!...~.n.:.P.!.~n.i.~~~--!?.Y...!~.~--~-':!r:f9.~.!.~.~~--f.<?.~..9.f.9.~!.i:i.d.~!!.o.11.:.......................................

:  : l Nothing emanating from the annulus region was confirmed.

             !      X       !                !                                                                                                                                                                       iLeak source -crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 66

:  : !Leak was repaired using a clamp assembly to preclude future degradation.

..............L.! ...........!.! .............!l.!?..~..!9.0.?.e...P..~.~!0.!.1.~!e..g.n._.Y.!j..~!9.~..o.f..n..c:>~.?;!~..................................................................................ii§!!.~.9~~-~9.U.n.~.e.9.f.\!Y. ...~.9.I.0.?.~.P.r.2~.i~~9...~Y...!~.~--::?.~!:f9_~.r.~.~!.f.c:>~..9.f.9.~r.9.9~!!.o.~.:....................................... Nothing emanating from the annulus region was confirmed. 67

             !      X       !                !                                                                                                                                                                       !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1). and 3142.3 (a)
             !              :                l                                                                                                                                                                       jLeak was repaired using a clamp assembly to preclude future degradation.
             !              :                :Orv loose particulate on UH side of nozzle                                                                                                                             l Structural inteoritv NOT compromised by the surface rust form of deoradation.

jNothing emanating from the annulus region was confirmed.

                                                                                                                                                                                                                     ~Leak source - crack in the Canopy Seal weld of Nozzle                                    77. (3142.1 (1}, and 3142.3 (a) 68              x                                                                                                                                                                                                l Leak was repaired using a clamp assembly to preclude future degradation.

..............!..............!..............l.1?..ry..!C?.O.~.~--P..Cl.rt!.'?.U..1.a.!~..C?.11..lJ.lj. ~!9.~..0.!..ri.o.:z.z.1~..................................................................................i§!~.~!U.!.Cl.l.!.ri!.~9.~o/....~.9.!...~.".1.P..r~n.i.~~~--:0.Y...t~-~--~-u.r:ta..~..ru..~t..f<?.r.".1..C?.f.9.~wa..d.a.!!.<?.r:'.:.......................................

              !             !                !                                                                                                                                                                       !Nothing emanating from the annulus region was confirmed.

1 X 1 1 1Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 69 !  !  ! l Leak was repaired using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • l**************l**************l*~C?!.9.r1..~rl..~.e.a..d...a..d.ia.~r:i.!..!~.P.~.ri~!!:a..!!C?.r1...........................................................................................t§~u..~!u..~a..!.!.ri!.e.9.r.!o/ ..~.9.!..!?.C?.".1.P.!.~".1.~S.~..:0.Y...!~.~--~-u.r:ta..~..~U..S.!.f<?.~..C?.f.9.~!.Cl.d.a.!!?.r:'.:.......................................

i i i  ! unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3 132.1 (a)) 70 i X i i i Supplemental Volumetric Examination performed identifying no change in structural i i !Significant boron accumulation (loose particulate and hard caked boron) on !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013.

• • • • • • • • • • • • • • l**************l**************l*r1.0.~!~..Clr:':d...'..U.~!..~.l~fl:l..9.r1..~.9..a.d...........................................................................................................y'.°!9..9.~!.Cl.d.a.~9.r:'..~.Cl~..!9.~.r1!.i.~~--!~Cl.!..~.U.!.d...~fl:'.P..~9.~!~.9...~!r~~-~U..~a.!Jr1!~.9.!.!ty:...................................................

i i i  ! u nable to visually confirm no nozzle leakage (3142.1 (b} and (c), 3142.2, 3130, and 3132.1 (a)) 71  ! X  !  ! !Supplemental Volumetric Examination performed identifying no change in structural i  ! !Significant boron accumulation (loose particulate and hard caked boron) and !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013. ..............L............L............L:.1_.1~.!..~!0.?.".1..9.ri..!3.~Y..~~a..d.........................................................................................................................l~9..9.~.~a..d.a.~?.r:i..~.a.~..!d..~.ri!.i.~~-d...!~a..!..~.u.!.d...~rl:'.P.!.?.~!~.e...s.!~~~!u..~a.!..i.ri!~.9.!.!!L.................................................

              !             !                !                                                                                                                                                                       ! Nothing emanating from the annulus region was confirmed.

72 ! l X ! l j Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1). and 3142.3 (a) l  ! l Leak was repaired using a clamp assembly to preclude future degradation. ..............L! ............J.! ............_l.1?..ry

                                                     ..!O..°.~.~--P.Cl.rt!~u..1.a..t.~..O..r1..lJ.t!..~!9.~..°.!..r1.°.:Z.Z.!~..................................................................................i§~U..~.~U.!.C1.l.!.r1!.~9.!.!o/..~.9.!..!?.C?.".1.P!.~rl:'.!~~d...:O.Y...!~.~--~-U.r:fCl.~fU..~!.f<?.~..C?.f.9.~!.C1.d.a.!!.°.r:'.:.......................................
                                                                                                                                                                                                                     !Nothing emanating from the annulus region was confirmed.

73

             ! X l                           l                                                                                                                                                                       jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a)
              !             !                !                                                                                                                                                                       l Leak was repaired using a clamp assembly to preclude future degradation.

..............l..............l..............l.§!!9.~Y~.i.f1?.!..a.~O..U..r1!..d..rx..!C?.O..S.~..f.>Cl.rt.i.9U..!C1.!~.~!.U..S.!..~a.~.~~-C?.r1..lJ..l2.~.i.d..e...a.f.r1.°.~.1.e..-.............l.§~u..~.~u..ra..!.!.ri!.~9.~o/..~.9.!..~0..".1.P!.~fl:l.i~~--!?.Y...!~.~--~-u.r:ta..~..~U..~!..f<?.!.~..C?.f.9.~!.C1.d.a.!i?.r:'.:.......................................

              !             !                !                                                                                                                                                                       ! Nothing emanating from the annulus region was confirmed.

74 i X !  ! !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) i i i l Leak was repaired using a clamp assembly to preclude future degradation . ..............l..............l..............l.1?..ry..!o..a.~.e...P..<1.rt!~u.1.a.!e...o..ri..lJ.lj..~!9.~..a.f..ri.a.:z.z.!~..................................................................................l§!.~.~-~u..ra..1.!.rit.e.g~o/..~.9.!..!?.C?.".1.P.!.~".1.!~~d...!?.Y...t.h.~..~.u.r:ta..~..~u..s.Uo.~..o..f.9.~!.a..d.a.ti_()r:'.:.......................................

              !             !                l                                                                                                                                                                       !Nothing emanating from the annulus region was confirmed.

75

             !      X       !               !                                                                                                                                                                        !Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1)._and 3142.3 (a)
              !             i                i                                                                                                                                                                       :Leak was repaired using a clamp assembly to preclude future degradation.
              !             !                ! Dry loose particulate on UH side of nozzle                                                                                                                            l Structural integrity NOT compromised by the surface rust form of degradation.

• • • • • • • • • • • • • r***********r***********r***************************************************************************************************************************************************************************TNoi"tiin9**e-n;-a-r;a-iin9..trom..itie**a-;;*r;uiu*;;*;:e-91c;;;**w-as*c:*c;r;ii;:iile-cr*******************************************************************************

1 X 1 1 jLeak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) 76

              !             l                l                                                                                                                                                                       ! Leak was repaired using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • l**************l**************l.1?..ry..!C?.°.~.~..!?.°.!.C?.rl.9.r:'..lJ..~..S.!d..~..O..f.!:i.O..Z.:Z.1.~.............................................................................................f§~.~-~U..ra..l.!.ri!.9.9.~o/..~.9.!...~.".1.P!.~".1.~S.~...~Y...!~.~--~-U.r:fCl.~..ru..S.!.f<?.~..C?.f.9.~!.Cl.d.a.!!?.r:'.:.......................................

i i i  ! unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2, 3130, and 3132.1 (a)) 77 ! X  !  ! !Supplemental Volumetric Examination performed identifying no change in structural i [ i !characteristics from previous Volumetric exams of all nozzles performed in 2006 and 2013. ..............L...........!..............l.~.e.9.~..@..~.IJ~P.Y...~~-9!..Y.-'.e.!~:.P!Y...~9.r.d...9.9.~~9..!?.9!.~.rl..9.r19..r.~.~!..g!.~rn.2r:i..~.~-9.~:..............i~2.9.~!.9.d.9.~9.r:i..~.9.~..!9.~.ri~.i.~~--~!:!9.~..~.~!.d...~~-P..r.2rn!~.e...~~~~-~~-~9.!..in~~-9.r.!l>.'.:...................................................

              !             !                !                                                                                                                                                                       i Nothing emanating from the annulus region was confirmed.

78 ! j X j  ! j Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1). and 3142.3 (a)

                            !                l                                                                                                                                                                       !Leak was repaired using a clamp assembly to preclude future degradation.

l  ! !SliqhUminor amount dry loose particulate on UH side of nozzle lStructural inteqritv NOT compromised by the surface rust form of deqradation.

Nothing emanating from the annulus region was confirmed. Leak source - crack in the Canopy Seal weld of Nozzle 77. (3142.1 (1}, and 3142.3 (a) Vent x Leak was repaired using a clamp assembly to preclude future degradation. Vent X Boron on vent line appears to have come from above line Structural inteqrity NOT compromised by the surface rust form of deqradation.

DE-McDermott Periodic Briefing, September 8, 2016 PLANTS I APR1400 Design Certification Review (EEEB)

    >> Phase 2 Chapter 8 SER and Chapter 14.3.6 SER completed and provided to NRO. Staff is awaiting OGG interpretation of SECY 91-0078 to address the applicant's conformance with the guidance in the SECY. Additional RAI was pr ovided to DNRL on Chapter 3 related to EQ methodology concerns.

Phase 2 SER for Chapter 3.1 1 and Chapter 19.3 is due September 19. NuScale Small Modular Reactor Topical Report (EEEB)

    >> NuScale submitted TR-0815- 16497, "Safety Classification of Passive NPP Electrical Systems" for NRG review approval of conditions of applicability. The conditions of applicability comprise a set of passive reactor plant design and operational attributes that, if met in full, justify the applicant's determination that none of the plant electrical systems fulfill functions that would warrant a Class 1E classification. An internal meeting was held to develop a revised RAI and the revised RAI was sent to DNRL on August 30. The NuScale Pre-DCA Readiness Assessment will take place on September 19.

Oconee (EEEB)

    >> Cable TIA. DPR staff held a meeting on August 30 with management to discuss path forward of the staff's responses to the Duke Fact Check. A committee of internal stakeholders has been formed (one person from each office of DPR, OGC, DORL, Region II, and EEEB) to collaborate on a draft response to Duke. DPR has scheduled a meeting on September 8 to begin this collaboration effort.
    >> DVR TIA. Resolving DE management comments prior to sending the draft TIA response to DPR for review. The current schedule will exceed the 2 year metric for OLTs.

Ginna Offsite Power TIA (EEEB)

    >> DPR is revising and processing the Tl for the licensee's fact check.

Perry Degraded Undervoltage TIA (EEEB)

    >> New TIA from Region Ill. Normal review schedule.

Indian Point 2 (EVIB)

    >> Licensee has replaced all 227 potentially defective bolts in IP2 plus an additional 51 bolts to add safety margin. The licensee performed an operability determination for Unit 3, which was found to be sufficient.

Region 1 has issued a violaUon for the licensee's handling of the issue. Browns Ferry (EMCB)

    >> The staff has been notified of a potential hearing on the Browns Ferry EPU. After clarification discussions with the licensee, the staff is in agreement with their approach and is awaiting their final written response to RAls.

Seabrook (EMCB)

    >> The staff is reviewing the license amendment request related to ASR. Following the acceptance review, the staff has decided to non-accept with the opportunity to supplement. The staff has provided input detailing the need for missing information and the PM is assembling the letter to the licensee.

Wolf Creek (EPNB) Wolf Creek experienced a leak in a canopy seal weld above a reactor head penetration. This leak caused the plant to enter its refueling outage early. This is not a pressure boundary leak per tech specs or ASME code as the pressure boundary is a threaded connection. Repair will be via a mechanical clamp. Region IV has the lead. The leak resulted in significant boric acid residue at various locations including the head. There is no known damage to the head at this time. Cleanup and inspection to ensure that none of the boric acid came from a leaking nozzle is an involved process. There will likely be one or more relief requests. Exact nature of the requests has not yet been determined. 1

DE-McDermott Periodic Briefing, September 8, 2016 PROCESS I ,( Comment IRMI: Do we have a date? OPC (EEEB)

    >> The staff received comments from the r egions and is working to finalize the first Tl (on interim                                                   I corrective actions and compensatory measures). The Tl will be issued for regional implementation                                                    !

after the Commission approves the IEP.

I Dl&C Action Plan (EICB)

    >> Issued to the Commission.

Dl&C CCF Policy Review (EICB)

    >> NEI also announced that they are developing a technical methodology for NRC endorsement. A                                                     !

I second tabletop was held on Augu st 22. NEI is now focused on potential changes to our current guida11ce for CCF (BTP 7-19). f I Dl&C 50.59 Guidance Review (EICB)  :

    >> The staff is holding bi-weekly meetings with NEI to resolve the key first-of-a-kind conceptual                                              /

approaches being proposed in Appendix D and the deviations from NEI 96-07. 1 MRP-227-A (EVIB) f

    >> Staff is currently reviewing PWROG-15032 regarding CASS materials (Action Item 7) and expects to                                           !

issue a safety assessment this fa1L.It1-~-~!9.f:f.i.~.<.!!\?9...~!:lJllJJnin.g.it~I..~Yj!:J_'!"_Qff~RQ~::!.~:!.Q153:~JJ:.',.Y'!'.~j~_IL *** J addresses the cold work aspect of Action Item 1 that was recently submitted for information. Rev. 1 to MRP-227-A, which addresses most of the licensee/applicant action items, has been received for staff review. Baffle-Former Bolts (EVIB)

    >> EPRI MRP started a baffle-former bolt focus group to develop an integrated industry response to the generic issue for PWRs . The MRP focus group issued interim guidance on July 25 and the guidance generally lines up with the information in the NSAL. LIC-504 has been completed and is in concurrence. Plan to issue an IN once test results come in this fall.

GALL Report - BMI and Capsule withdraw, liner bulge new issue PEOPLE On Board Budget EOY On Boa rd FY17 Budget 78 82 74 76 Khadijah West on rotation from STSB to EEEB through December. Sergiu Basturescu on rotation from EEEB to STSB through February. Yong Li, Yuan Cheng, and Amit Ghosh will move to NRO COE on October 2 from EMCB. Subinoy Mazumdar, Deirdre Spaulding, Eugene Eagle (EICB), Jack McHale (EVIB BC), and Robert Hardies (SLS) opted for early retirement. Depart October 1. Developing staffing plan input to address 3 vacancies in EICB. Vacancies

    >> SLS Electrical: Preparing ERB package to support backfill for Hardies
    >> GG-14 Electronics Dl&C: EICB beginning process
    >> GG-14 Electronics Dl&C: EICB beginning process
    >> GG-13 Electronics Dl&C: EICB beginning process
    >> GG-14 Materials: Agency-wide SOI closes this week.
    >> GG-14 Materials: Reassignment from JLD identified. Expected on board in January 2017.

Overages

    >> EEEB: one unfunded position in FY17. Reassignment from JLD without FTE.
    >> EPNB: two unfunded positions in FY17. One is a no-backfill and was an EO/BO target.

2

DE-McDermott Periodic Briefing, September 8, 2016 PENNIES I Cost Center Buda et Commitments Obliaations Remainina 1061 448 484 384 0 1065* 2,544 2,544 2,544 0 Total 2,992 3,028 2,928 0

• Topical Reports are budgeted in DPR and funding is moved to DE at the time it is committed . The peening topical report will be moved to RES because in essence it is confirmatory research.

Final contract action for NRR/DE is in negotiation and should be awarded and obligated by the end of the month. 3 to WO 16-0052 Page 3 of 13 Figure 2 in ASME Code Case N-729-1 , as referenced by paragraph -2500, requires that the volumetric or surface examination coverage distance below the toe of the J-groove weld (i.e. dimension "a") be 1.5 inches for incidence angle, e. less than or equal to 30 degrees; 1 inch for incidence angle, e, greater than 30 degrees; or to the end of the tube, whichever is less. These coverage requirements are applicable to Wolf Creek Generating Station (WCGS) reactor vessel head penetrations as shown in Table 1. Table 1: WCGS Reactor Vessel Head Penetration Coverage Requirements Penetration Numbers *red Coverage, " a" (inches) 1 to 29 1.5 30 to 78 1.0

4. Reason for Request styles of ends, referred to gh 73 are Type "Y" that are

                                                            ~lit.1"9r diameter and inner diameter.

meter and an internal taper. tion nozzles 74 through 78, referred to as

                                                           , approximately 1.19 inch in length at the re located at the 48.7 degree location. The at th1            1s such that the distance from the lowest point Id to the top of the threaded region could be less than the "a" shown in Figure 2 of ASME Code Case N-729-1.

ired inspection coverage is sought for reactor vessel

                                       , as the required coverage for these two penetrations

t)~!-P c2-o/\<i{; I< c:__,

From: Muilenburg William T Sent: 11 Oct 2016 19:44:39 +0000 To: Singal, Balwant

Subject:

[External_Sender] Participants from Wolf Creek on today's call

Balwant, On the call today for Wolf Creek were:

Steve Smith, Plant Manager Richard Flannigan, Manager Nuclear Engineering Cindy Hafenstine, Manager Regulatory Affairs Dennis Tougaw, ISi Engineer Mark Barraclough, Boric Acid Program Owner Bill Muilenburg, Licensing Supervisor On the other request Debbie Hendell, our Senior Counsel, said she will call this afternoon.

From: Good Nicole R Sent: 28 Oct 2016 17:24:30 +0000 To: Drake, James;Alley, David;Collins, Jay; Cumblidge, Stephen; Dodson, Douglas;Kopriva, Ron;Taylor, Nick;Thomas, Fabian;Singal, Balwant;lingam, Siva

Subject:

[External_Sender) RE: Penetration pictures for Relief Requests 14R-03 and 14R-04 The upload is complete. Thank you, Nicole Good From: Good Nicole R Sent: Friday, October 28, 2016 10:24 AM To: James.Drake@nrc.gov; Alley, David (David.Alley@nrc.gov); Collins, Jay (Jay.Collins@nrc.gov); Cumblidge, Stephen (Stephen.Cumblidge@nrc.gov); Dodson Douglas E; Ron.Kopriva@nrc.gov; Nick.Taylor@nrc.gov; Thomas Fabian D; Balwant.Singal@nrc.gov

Subject:

Penetration pictures for Relief Requests 14R-03 and 14R-04 Pictures have been uploaded for you in the Certrec IMS Sept 2016 Forced Outage folder Item #27 . The upload of the pictures to the folder in IMS is still in process. I will email when the uploads arc complete. Thank you, Nicole Good Licensing nilyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek Nucleor Operating Corporotion

From: Good Nicole R Sent: 14 Oct 2016 16:09:15 +0000 To: Lingam, Siva Cc: Singal, Balwant;Collins, Jay; Kopriva, Ron;Dodson, Douglas;Thomas, Fabian

Subject:

[External_Sender) RE: WCNOC RV pictures I forgot to include this. -Pictu re # - pen etration numbers DSC00006 - 53, 64, 59, 36, 47, 60 DSC00039 - 63, 52, 34, 26, 20, 27, 58, 70, 75, 57, 33, 16, 12 DSC00029 - 46 DSC00019 - 47, 71, 77 DSC00018 - 40, 46, 70, 71 From: Good Nicole R Sent: Friday, October 14, 2016 11:05 AM To: 'Lingam, Siva' Cc: Singal, Balwant; Collins, Jay; Ron.Kopriva@nrc.gov; Dodson Douglas E; Thomas Fabian D

Subject:

RE: WCNOC RV pictures Mr. Lingam, I have been told by Certrec that you have been contacted with a link to Certrec and your password. You have printing rights. IMS Sept 2016 Forced Outage Item #21 has several pictures of the penetrations with labels. Inclu ded is M-706-00009 Reactor Pen, it will help with orientation of the picture labels. Thank you, Nicole Good From: Lingam, Siva [ mailto:Siva.Linqam@nrc.gov] Sent: Thursday, October 13, 2016 3:15 PM To: Good Nicole R Cc: Singal, Balwant; Collins, Jay

Subject:

RE: WCNOC RV pictures Please provide me the Certrec link and password with printing rights. Thank you . From : Good Nicole R [12] Sent: Thursday, October 13, 2016 4:05 PM To: Lingam, Siva <Siva .Lingam@nrc.gov> Cc: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

[External_Sender] WCNOC RV pictures I was told you would like pictures of the penetrations with labels of the penetration number. I have only been able to locate a few pictures, at this point. I have granted you access to the Certrec IMS Sept 2016 Forced Outage. Item #14 has five pictures that may be

helpful (DCS00006, DCS00039, DCS00029, DCS00019, and DCS00018). I will need to contact Certrec to get access for Mr. Singal. 1 will work on getting Mr. Signal access and looking for more pictures tomorrow. Thank you, Nicole Good Licensing nilyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek ~ Nucteor Operating Corporot1on

From: Good Nicole R Sent: 21 Oct 2016 13:52:47 +0000 To: Pascarelli, Robert;Lingam, Siva;Singal, Balwant;Taylor, Nick;Kopriva, Ron;Kennedy, Kriss;Dodson, Douglas;Thomas, Fabian

Subject:

[External_Sender) Reactor Head Images Uploaded to Certrec IMS To all: Images of the reactor head have been uploaded to the respective items in the CERTREC Inspection Management System folder for the September 2016 Forced Outage Inspection : Flange and Head with Clamps installed - IMS item 23 Flange Area As Found - IMS item 24 Eyebolt and Upper Canopy Seal Weld Design - IMS item 25 Videos of Head Inspection - IMS item 26 Thank you, Nicole Good Licensing nilyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek Nucteor Operoting Corporotion

From: Hafenstine Cynthia R Sent: 12 Oct 2016 22:10:42 +0000 To: Singal, Balwant;'siva.lingman@nrc.gov' Cc: Muilenburg William T;Tougaw Dennis E;Barraclough Richard M

Subject:

[External_Sender) Wolf Creek - Draft revision of Relief Request Document Number WO 16-0052 Attachme nts: W016-0052R5dt.pdf Attached is our current draft revision of the relief request. We have not yet incorporated the questions listed in the draft RAI that you provided. We would like to have a call on Thursday at 1:00 pm Eastern Time I Noon Central Time. Please let me know if that will work for you. We appreciate your support in getting this document revised to support our request.

Thanks, Cindy Hafenstine Office 620-364-4204 Celll(b)(6) I

From: Muilenburg William T Sent: 7 Oct 2016 21:17:37 +0000 To: Singal, Balwant

Subject:

[External_Sender] Wolf Creek Relief Request Anticipated

Balwant, I wanted to give you advance notice that on Monday morning (10/10) Wolf Creek will be sending a Relief Request for review concerning reactor vessel head inspections. We need to perform supplemental exams on certain penetrations and we have two concerns.

First, one penetration is one where we have had relief on before because of access concerns and we will need to request the same relief again (ML12353A241 provided NRC Safety Assessment of the request), and second, we will be asking to perform an alternate exam v. that specified in code case N-729. Can you help us assemble the right people to have a phone call regarding this request on Monday morning? I will be in Saturday and Sunday if there are any questions I cain help answer.

Thanks, Bill Muilenburg 620-364-4186

From: Muilenburg William T Sent: 1 Nov 2016 14:41:44 +0000 To: Singal, Balwant;Taylor, Nick

Subject:

[External_Sender] Wolf Creek Relief Requests Attachments: ET16-0030.PDF See listing of Records A lready Available to Balwant/Nick, Here is the Relief Requests related to the WCNOC reactor vessel closure head. Key Points l. This supersedes all previous correspondence on this topic

2. The request now encompasses all penetrations on the head Bill

From: Hafenstine Cynthia R Sent: 13 Oct 2016 17:00:49 +0000 To: Singal, Balwant

Subject:

[External_Sender] FW: Relief Request for Code Case N-729-1 Attachments: M-706-00009_REACTOR PEN.JPG One-page attachment withheld in full under ex4. New drawing for the draft relief request... From: Barraclough Richard M Sent: Thursday, October 13, 2016 11:53 AM To: Hafenstine Cynthia R Cc: Tougaw Dennis E

Subject:

Relief Request for Code Case N-729-1 This is the image I had Salvador Ferrara put together for the relief request R. Mark Barraclough Wolf Creek Nuclear Boric Acid Engineer I Program Owner Fluid Leak Management I Program Owner AOV Engineer 620-364-8831 x8148 I ribarra@wcnoc.com Fax: 620-364-4154 (b)(4)

From: IMS Sent: 14 Oct 2016 10:14:41 -0500 To: Lingam, Siva Subje ct: [External_Sender] Login Information - Certrec Inspection Management System (IMS) Siva Lingam, Welcome! You have been granted access to the CerlTec Inspection Management System in preparation for an upcoming NRC Inspection. This request was made by Nicole Good from Wolf Creek. Within the next 72 hours, please follow the Link below and enter your Usemame and temporary Verification Code. Once entered, you will be prompted to create a password for your account. https://ims.certrec.com/verify/ Usemame: Db( Verification Code: lf72 hours have already passed, please follow the steps above which will prompt a new temporary Verification Code to be emailed as the one above will have expired. ln the future, you will not need to repeat this process to "Sign In" to the Certrec Inspection Management System. The Ccrtrec Inspection Management System can be accessed at: https://ims.ccrtrcc.com As always, please feel free to contact the Certrec Support Team if you have any questions or concerns. Thank you, Certrec Corporation Certrec Support Team support@certrec.com 817.738.7661

From: Wilson, George Sent: 31 Oct 2016 14:34:22 -0400 To: Orf, Tracy

Subject:

FW: Wolf Creek Vessel Head Corrosion - One Pager and Q&As Attachments: WOLF CREEK-RCS- Q&A.docx, Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion.docx Trace these are the emails that I have for the foia on wolf creek George Wilson Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation USN RC 301-415-1711 Office 08E4 From: Wilson, George Sent: Friday, October 28, 2016 7:01 AM To: Wilson, George <George.Wilson@nrc.gov>

Subject:

FW: Wolf Creek Vessel Head Corrosion - One Pager and Q&As George Wilson Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation USN RC 301-415-1711 Office 08E4 From: Lyon, Fred Sent: Thursday, September 22, 2016 7:33 PM To: Wilson, George <George.Wilson@nrc.gov>; Boland, Anne <Anne.Boland@nrc.gov> Cc: Alley, David <David .Alley@nrc.gov>; Evans, Michele <Michele.Evans@nrc.gov>; Dean, Bil l

<Bill.Dean@nrc.gov>; McDermott, Brian <Brian.McDermott@nrc.gov>; Lubinski, John
<John.Lubinski@nrc.gov>; Ross-Lee, MaryJane <MaryJane.Ross-Lee@nrc.gov>; Pascarelli, Robert
<Robert.Pascarelli@nrc.gov>

Subject:

FYI : Wolf Creek Vessel Head Corrosion - One Pager and Q&As From: Taylor, Nick Sent: Thursday, September 22, 2016 7:19 PM To: Kennedy, Kriss <Kriss.Kennedy@nrc.gov>; Morris, Scott <Scott.Morris@nrc.gov>; Pruett, Troy

<Troy.Pruett@nrc.gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov>; Vegel, Anton <Anton.Vegel@nrc.gov>;

Clark, Jeff <Jeff.Clark@nrc.gov>; R4DRP-BC <G-R4-DRP-BC@nrc.gov>; Werner, Greg

<Greg.Werner@nrc.gov>; Dricks, Victor <Victor.Dricks@nrc.gov>; Maier, Bill <Bill.Maier@nrc.gov>;

Moreno, Angel <Angel.Moreno@nrc.gov>; Bowen, Jeremy <Jeremy.Bowen@nrc.gov>; Lyon, Fred

<Fred.Lyon@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc.gov>

Cc: Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Dodson, Douglas

<Douglas.Dodson@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Galemore, Susan
<Susan.Galemore@nrc.gov>; Kopriva, Ron <Ron .Kopriva@nrc.gov>

Subject:

Wolf Creek Vessel Head Corrosion - One Pager and Q&As Good afternoon, Please see the final one-pager and Q&A document related to the recently discovered corrosion on the Wolf Creek reactor vessel head. This has already been provided to the EDOs office. Please feel free to forward as needed to other interested parties. We will evaluate the need for an update after the licensee completes a more thorough inspection of the vessel head! (anticipated within the next week). Please feel free to contact me with any questions that you have.

Thanks, Nick Taylor Chief, Projects Branch B Division of Reactor Projects USNRC Region IV 0: (8 17) 200-1141 c:l<bl<5 i I E: nick.taylor@nrc.gov R

Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Q's and A's

1. Does the red rust indicate damage to the reactor vessel head? How soon will the extent of possible damage be known?

The red coloration could indicate corrosion of the reactor vessel head from the reactor coolant system leakage that became apparent on August 31, 2016 and continued until the licensee fully depressurized the plant. The extent of this corrosion is currently unknown, but believed to be limited to a small sector of the head area. It will take several weeks to clean the vessel head and determine the extent of the corrosion.

2. What is the cause of the rust on the head?

The rust was likely caused by a small leak spraying onto the vessel head from a seal weld on a mechanical joint on a vessel head nozzle above the area. This leak manifested itself on August 11, 2016 and continued until the plant was depressurized.

3. Please describe the extent of the discoloration? How and when was this discovered?

The discoloration is limited to a small sector underneath the leaking head nozzle. This was discovered following a required shutdown on September 2, 2016. The licensee was able to observe the discoloration on the head after removing insulation from the vessel head on September 19, 2016.

4. If there is damage, will the reactor head have to be replaced or can it be repaired?

Depending on the extent of corrosion, the licensee will evaluate the repair/replace options, prior to the end of the current 55-day refueling outage.

5. Will the licensee need NRC permission to restart?

At this time there is no indication that the licensee will need NRC permission to restart. NRC inspectors will verify that the licensee has adequately addressed the issue in order to safely restart and operate the plant. However, NRC personnel are in close communication with the licensee, and will closely monitor the licensee's actions to repair the leak and the vessel head, if needed.

6. Following the incident at Davis Bessie aren't all reactor licensees required to conduct periodic vessel head inspections? Has Wolf Creek done theirs?

Yes. Wolf Creek performs inspections of the their vessel head each refueling o utage in accordance with their approved in-service inspection program. The last such inspection was in the Spring 2015 outage.

Contact:

Nick Taylor, Chief, Reactor Projects Branch B (817)200-1141

7. Did this undiscovered condition present any damage to public health and safety while the plant was operating?

No. The leak was small and well within the design basis of the plant. The licensee shutdown the plant after indication of a sudden increase in leakage beyond that allowed by the Technical Specifications.

8. What role is the NRC playing in this?

The NRG resident inspectors and Region IV personnel monitored the licensee's actions promptly when indication of a leak first surfaced, and verified that the licensee took action promptly and in accordance with their operating license. NRG inspectors from the Region IV office will be onsite beginning September 26 to assist the resident inspectors with their follow-up of this issue. NRG management is in communication with Wolf Creek management o n their path forward.

9. Is the licensee required to file written reports with the NRC regarding this?

Yes. NRG regulations require a written Licensee Event Report (LER) for completion of the Technical Specification-required shutdown, on September 2, 2016.

Contact:

Nick Taylor, Chief, Reactor Projects Branch B (817)200-1141

Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Key Messages

• Wolf Creek completed a technical specification (TS) required shutdown of the reactor on Friday, September 2, 2016, in order to locate and repair the source of elevated reactor coolant system leakage. The source of the leak was determined to be a leaking canopy seal weld on a core exit thermocouple penetration nozzle above the reactor vessel head.
• Upon initial inspection on September 19, indication of carbon steel corrosion was noted on the reactor vessel head itself. Although the extent of the corrosion is not yet known, it appears to be limited to a small sector of the reactor vessel head directly below the leaking penetration.
• Following the shut down the licensee began a planned refueling outage. The licensee is in the process of removing the reactor vessel head to conduct an evaluation of the impact of the leakage and is identifying plans for further analysis and appropriate actions, including repair of the leaking nozzle. The NRC will continue to monitor the licensee's progress.

Facts

• Wolf Creek noted an upward trend in unidentified RCS leakage on August 31, 2016. On September 2, 2016, Wolf Creek observed RCS unidentified leakage in excess of 1.35 gallons per minute (gpm),

exceeding the TS limit of 1.0 gpm. As a result, the licensee initiated a TS required shutdown on September 2, 2016.

• The resident inspectors monitored reactor coolant system leakage throughout the operating cycle.

Data indicated a steady very small leak rate (approximately 0.05 gallons per minute), that suddenly began to increase on August 31, 2016.

• Following shutdown and containment entry, the source of the leak was identified as the canopy seal weld on penetration 77 on the reactor vessel head, which serves one of the core exit thermocouples.

The threaded mechanical joint serving the core exit thermocouple nozzle assembly is not considered pressure boundary leakage.

• Following the shutdown, the licensee remained shutdown to commence their refueling outage, which is planned for 55 days. During this outage, the licensee is evaluating plans to repair the leak using an applicable ASME code allowable methodology. Previous minor leaks on mechanical joints on the reactor vessel head have been repaired with code-approved mechanical clamps. There are 10 of these clamps currently installed on vessel head nozzle assemblies.
• The reactor vessel head is the original head and is approximately 30 years old. The licensee has periodically inspected the head for leakage in accordance with their approved in-service inspection program. The last such inspection was in the Spring 2015 refueling outage.
• The license plans to remove the vessel head and place it on an inspection stand. The reactor vessel head will be cleaned and examined to determine the extent of the corrosion, and if repairs are necessary.
• Region IV inspectors from the Division of Reactor Safety are scheduled to arrive on September 26 to assist the resident inspectors inspection of this issue.

September 20, 2016

Contact:

Nick Taylor, Chief, Reactor Projects Branch B {817) 200-1141

  ~ Wolf Creek Nuclear Operating Corporation
 ~          -                                 00093697 Condition Report AR #: 00093697     Severity Type:    CAQ      Level :  SSC      Due Date: 1211112015        St atus:COMPLETE      Status Date:     1211012015 A R Subject :  STS PE-040E Penetration 20 Canopy Seal Weld Leakage lndicati                                      Age In Days:   264 Owed To Name: DORATHY, BRIAN D                                            Origination Date: 0311812015 Owed To Department: 4050050 - Dorathy Brian                                              Initiator: HALL, JOHN F Owed To Alert Group :                                                          Orig Department: 0060030 - Heffron Jason Condition Report Summary:

Ty pe AR#-Assign#-Sub-Assign# Owed/Assign To Due Date Status CAQ 00093697 BRDORAT 12111/2015 COMPLETE RTFQ 00093697-01 OPS REVIEW COMPLETE RACT 00093697-01-01 OPS REVIEW COMPLETE RACT 00093697-01-02 OPS REVIEW COMPLETE RACT 00093697-01-03 OPS REVIEW COMPLETE RACT 00093697-01-04 OPS REVIEW COMPLETE RACT 00093697-01-05 OPS REVIEW COMPLETE RACT 00093697-01-06 OPS REVIEW COMPLETE RACT 00093697-01-07 OPS REVIEW COMPLETE RACT 00093697-01-08 OPS REVIEW COMPLETE RACT 00093697-01-09 OPS REVIEW COMPLETE RACT 00093697-01-10 OPS REVIEW COMPLETE RACT 00093697-01-11 OPS REVIEW COMPLETE RACT 00093697-01-12 OPS REVIEW COMPLETE RACT 00093697-01 -13 OPS REVIEW COMPLETE RACT 00093697-01-14 OPS REVIEW COMPLETE RACT 00093697-01 -15 OPS REVIEW COMPLETE RACT 00093697-01-16 OPS REVIEW COMPLETE RACT 00093697-01-17 OPS REVIEW COMPLETE RACT 00093697-01-18 OPS REVIEW COMPLETE RACT 00093697-01-19 OPS REVIEW CANCELED RACT 00093697-01 -20 OPS REVIEW CANCELED BLL 00093697-02 BRDORAT 0412012015 COMPLETE PLAN 00093697-03 BRDORAT 12/11/2015 COMPLETE ACT 00093697-03-01 DAGIEFE1 11/2012015 CANCELED ACT 00093697-03-02 BRDORAT 1211112015 COMPLETE ACT 00093697-03-03 BRDORAT 1013012015 COMPLETE Attachments: CR Detail A sset/Equip: RBB01 Work R,equest: 15-111261

Description:

During the performance of STS PE-040E evidence of leakage was identified at the canopy seal weld on penetration

20. The indication appears on the east side of the seal weld face (as the RPV closure head sits). Boron staining is evident above, adjacent and b,elow the seal weld indication. Rust staining is also prevalent at and below the indication running down the CROM nozzle and onto the top head surface. There is no evidence of degradation to the CROM or closure head surfaces. This seal weld indication appears to have been an active leak in the recent past.

Photos/video available from QC (Jason Heffron). Recommend th is seal weld be evaluated for installation of a seal weld clamp assembly. CR Detail Report Page 1 of 28 1211012015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Immediate Concern: N SM Notified: N/A lnit DNC: N Immediate Actions: Initiate CR Extent of condition: RPV Canopy seal welds Recommended Resolution: Install canopy seal weld clamp assembly Screening Review Operability: 3 OPER/DNC The initiator identified during the perfor mance of STS PE-040E evidence of leakage at the canopy seal weld on penetration 21 . The indication is on the east side of the seal weld face (as the RPV closure head sits). Boron staining is evident above, adjacent and below the seal weld indication. Rust staining is also prevalent at and below the indication running down the CRDM nozzle and onto the top head surface. There is no evidence of degradation to t he CRDM or closure head surfaces. I reviewed the pictures that are located at K:\Data\NDE\Photos\RF-20\CRDM Head Inspection. In these pictures there are visible traces of dried boron and some small amounts of discoloration on the seal weld. As the initiator identified, there is no significant accumulation of boron or wastage of any carbon steel on the head penetration directly below the subject seal weld. Technical Specifications defines Pressure Boundary LEAKAGE as LEAKAGE through a nonisolable fault in an RCS component body, pipe wall or vessel wall. TS 3.4.13 contains the operating limits for RCS Operational LEAKAGE. In MODES 1 through 4, no pressure boundary is allowed, unidentified LEAKAGE is limited to 1 gallon per minute, identified LEAKAGE is limited to 10 gallons per minute, and primary to secondary LEAKAGE is limited to 150 gallons per day in any one Steam Generator. The Control Rod Drive Mechanism is what is used to raise, lower, and trip control rods. The internals of this mechanism is exposed to RCS pressure. The Drive Mechanism Latch Housing is internally threaded and torqued down onto a seating surface at the interface between the housing and the top of the Reactor Head Adapter. This connection is a mechanical joint and leakage via this pathway is not Pressure Boundary LEAKAGE as defined by Technical Specifications. The WCGS reactor vessel head and CRDM assemblies are classified as ASME Boiler and Pressure Vessel Code Section Ill Class 1 items. The Reactor Vessel was designed and fabricated to the 1971 Edition through Winter 1972 Addenda and the CROM housing assemblies were designed and fabricated to the 1974 through Winter 1974 Addenda of Section Ill of the ASME B&PV Code. Section Ill paragraph NB-3671.3 states that threaded joints un which threads provide the only seal shall not be used. The seal weld is not a structural part of the pressure boundary and is not required to meet the structural requirements of ASME B&PV CR Detail Report Page 2 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Code, Section Ill, NB-3000. The threads are the load carrying part of the joint design. The industry indications and past operating experience at WCGS of leaks in the subject seal welds are pinholes or small localized cracks. These flaws have resulted in leak rates that are bound by the limits established in Technical Specification 3.4.13. Completed performances of STS BB-006 were reviewed from the last operating cycle and RCS leakage limits were not challenged. The Reactor Vessel and the subject CROM is OPERABLE but degraded due to the flaw in the lower seal weld.

References:

Technical Specifications 1.1, 3.4.13 and Bases; TR 3.4.17 and Bases; NRC Inspection Manual Part 9900, WCGS Correspondence CT 02-0029, Westinghouse Instruction and Operating Book for Magnetic Control Rod Drive Mechanism for Full-Length Control Rods, and STS PE-040E. TSS 3/21/15 Note- The reference to NRC inspection manual 9900 has been superceded by NRC inspection manual 0326-Additionally, investigation by engineering has determined the actual condition is on the canopy seal weld on penetration 20. This does not change the operablity basis. Reportable: N Env ironmental Issue: N Tech Spec Sec 5: N Personnel Safety Issue: N Reactivity Issue: N Impact Risk Assessment: N OPS Review: BELL, SETH A CR/WR Screening: BELL, SETH A Significance Cat: 99 - NOT APPLICABLE Screen/SRT Notes: General Notes: Updated By Last Updated This is a long standing issue in the industry. Westinghouse BRDORAT 03/20/2015 has previously identified this to be transgranular stress BRDORAT 03/20/2015 corrosion cracking (TGSCC) susceptibility in austenitic BRDORAT 03/20/2015 stainless steel influenced by the environment, stress, and BRDORAT 03/20/2015 microstructure. The residual stress associated with the BRDORAT 03/20/2015 weld is sufficient to promote TGSCC on the annealed type 304 BRDORAT 03/20/2015 stainless steel canopy seal in a corrosive environment. One BRDORAT 03/20/2015 type of environment in which annealed stainless steels are BRDORAT 03/20/2015 known to be susceptible to TGSCC is in high temperature, BRDORAT 03/20/2015 CR Detail Report Page 3 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report stagnant chloride/oxygen environments. This is associated BRDORAT 03/20/2015 with Head Adaptor Plug Canopies, which is location of this BRDORAT 03/20/2015 leak. Westinghouse has recommended to perform inspections BRDORAT 03/20/2015 of these locations to identify any telltale rusty colored BRDORAT 03/20/2015 spots, and white boric acid deposits and tracks. This is BRDORAT 03/20/2015 what was occurring when this leak was identified. With the BRDORAT 03/20/2015 cause already identified and actions in place, which BRDORAT 03/20/2015 identified this leak, an apparent cause would not be BRDORAT 03/20/2015 beneficial to this issue. BRDORAT 03/20/2015 This issue was discussed with the initiator. BRDORAT 10/26/2015 Other Related Information Assignment Status Summary: Total Assigns/Subs: 3 - 23 Open Assigns/Subs: 0 - 3 Overdue Assigns/Subs: 0 - 1 Cross

References:

Type Number Sub Number MPAC WORK REQUEST 15-111261 Status & Due Date History: Responsible Person Date Updated Status Due Date HALL, JOHN F 03/18/2015 INPROG HALL, JOHN F 03/18/2015 H/APPR NEILSON, RHONDA G 03/21/2015 APPROVED 04/20/2015 BELL, SETH A 03/19/2015 PRE-APRV DORATHY, BRIAN D 04/20/2015 12/1112015 DORATHY, BRIAND 12/10/2015 COMPLETE Evaluation/Check! ist BLL Assignment #: 00093697-02 Due Date: 04/20/2015 Status: COMPLETE Status Date: 04/20/2015

Subject:

STS PE-040E Penetration 20 Canopy Seal Weld Leakage lndicati Age In Days: 30 Total Age: 30.00 Assigned To Name: DORATHY, BRIAND Assigned To Organization: 4050050-ENGINEERING DAILY WORK CONTROL SUPV - DORATHY

Description:

Perform a basic evaluation in accordance with AP 28A-100 and Al 28A-100. Use form AIF 28A-100-12, Basic Cause Evaluation. This evaluation does not require a qualified evaluator. Contact the CAP group for further assistance if determined this assignment is not needed. DO NOT status the assignment as Complete. CR Detail Report Page 4 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Condition Statement: PROBLEM STATEMENT: The canopy seal weld on penetration# 20 was found to be leaking during the performance of a remote bare metal visual examination (VE) of the Reactor Vessel (RV) head penetrations (STS PE-040E performed during RF20). The indication appears on the east side of the seal weld face. Boron staining is evident above, adjacent and below the seal weld indication. Rust staining is also prevalent at and below the indication running down the CRDM nozzle and onto the top head surface. There is no evidence of degradation to the CRDM or closur*e head surfaces. Based on the observed staining pattern, this seal weld indication appears to have been an active leak in the recent past. CR INITIATOR CONTACT: HALL, JOHN F. The ISi Engineer involved in this evaluation was also involved in the examinations that identified the canopy seal weld leak. The ISi Engineer thoroughly discussed this issue with the CR Initiator during and after the examination . Extent of Condition: This condition is limited to the lower canopy seal welds at each of the 78 RV head nozzle penetrations because of the design and configuration of this connection (see the evaluation section for description of the design and configuration). This CR addresses all of these lower canopy seal welds in the 78 RV head nozzle penetrations. Therefore, no further extent of condition considerations are applicable. Operating Experience: An Operating Experience search of the INPO website identified a number of similar issues at plants resulting in a leak in the canopy seal weld area. Industry experience using a canopy seal clamp assembly has been successful. The table in the Evaluation section below lists a number of canopy seal leaks that have been successfully repaired at WCGS. As noted in WCAP-12088, the transgranular stress corrosion cracking of the canopy seal weld is a known industry issue. Some applicable OE that can be used for reference, are: OE4046 (Indian Point 3), OE15763 (Indian Point 2), OE23609 (VC Summer), OE27722 (Millstone 3), OE31028 (North Anna 2), Plant Event #40228 (Seabrook), 120388 (Wolf Creek). In addition, WCAP 12088 describes the results of failure analysis of canopy seal weld leaks from S different plants that led to the creation of the WCAP . Wolf Creek has also had history of leaks at the canopy seal weld, as evidenced by the number of canopy seal clamps installed (see the table in the Evaluation section below). Most of the canopy seal clamps are installed at spare penetration locations, which seem to be the most susceptible to the transgranular stress corrosion cracking. As identified in WCAP-12088, the failure mechanism of the canopy seal weld is known. The method of repair is also well-proven and has been successful throughout the industry. CR Detail Report Pages of 28 12/10/201S 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Evaluation and

Conclusion:

EVALUATION: The reactor coolant system (RCS) transfers the heat generated in the reactor core to the steam generators via the reactor coolant pumps. The RCS operates at of 2235 psig and at a cold leg temperature of approximately 557 oF and a hot leg temperature of approximately 618 oF. The RCS must provide the pressure boundary barrier for containing the coolant under all anticipated temperature and pressure conditions. And, at elevated temperatures and wet conditions, the boric acid found in the RCS coolant is highly corrosive to carbon steel (the material of construction for the reactor vessel). At WCGS, the reactor head contains 78 penetrations, of which 53 are for full length CROM Assemblies. The following identifies penetration configurations:

                           - 53 are for full length CROM Assemblies
                           - 13 are for plugged, spare head adapters
                           - 8 are for capped latch housings Assemblies
                           - 4 are for female flanged instrument port columns At the upper end of each penetration is a stainless steel head adaptor flange. The stainless steel flange has male ACME threads and a canopy lip to allow attachment of the reactor control components.

The reactor vessel head penetration assemblies at WCGS consist of a two-piece construction - an lnconel tube section welded to a stainless steel (type 304) flange, referred to as the head adapter flange. The Incon el tube section has an interference fit with the reactor vessel head and is attached by a partial penetration weld. Every head adapter flange attached to the lnconel tube is designed and fabricated the same. Each is fabricated with a canopy lip and a threaded end for installation of attachments. During field installation, the attachment (the CROM, head adapter plug, etc.) is threaded into the head adapter flange. A gas tungsten arc weld process is used to form a seal weld. This seal weld is referred to as the lower canopy seal weld. Each of the attachments has female ACME threads to allow attachment to the stainless steel flange on the end of the CROM penetration and a canopy lip. The head adaptor is designed such that when the attachment is threaded onto the stainless steel flange (during the original construction), the two canopy lips come together. They are seal welded since the ASME Section Il l Code states that threaded joints in which threads provide the only seal shall not be used. Hence the canopy seal weld was provided to seal the ACME threads. It is important to note that the ACME threads of the threaded connection provide the structural design strength and pressure boundary of the joint. The canopy seal weld provides leakage control of the threaded connection, but does not provide any of the ASME Code strength of the connection. The canopy seal welds are thin welds of about 0.070" thickness that serve to seal the threaded pressure boundary connection. The canopy seal weld configuration forms a "dead end" in which impurities in water that works its way past the threads can accumulate. It is suspected that the water used during cold hydrostatic testing and hot functional testing remains in the canopy seal area for the life of the joint (unless a leak develops). Additionally, each time the head is removed from the vessel and re-installed, and the vessel is repressurized, the trapped water in the seal area is oxygenated. This environment establishes conditions that appear to increase the probability of stress corrosion cracking. Annealed stainless steels are known to be susceptible to transgranular stress corrosion cracking (TGSCC) in this type of high temperature, stagnant chloride/oxygen environment. CR Detail Report Page 6 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Westinghouse performed destructive examination and hardware failure analysis on a number of lower canopy seal welds removed from nuclear power plants in the late nineteen eighties (WCAP-12088). In addition, Westinghouse assisted WCNOC personnel in performing a hardware failure analysis report for leaking canopy seal welds (References 4 and 5). These investigations concluded that the failure mechanism was TGSCC. The cracking was identified both in the seal welds and in the base metal of the seal weld joints. All cracking was initiated from the interior of the joint. No sensitization of the material was identified in the components examined. Very low levels of chloride contamination were noted in water samples obtained from removed weld joint areas, as well as residues on the threaded surfaces that were analyzed. Westinghouse determined that the residual stresses associated with the seal welding process, combined with the very low levels of chloride in the oxygenated stagnant region, were sufficient to promote TGSCC on the joint. The Westinghouse hardware failure analysis also included examination of some threaded joints that were removed along with the lower canopy seal welds. There was no evidence of corrosion or cracking on any of the threaded joints that were examined. Austenitic st ainless steel in the presence of chloride/oxygen in the high pressure and temperature environment is susceptible to Transgranular Stress Corrosion Cracking (TGSCC). The residual stress associated with the canopy seal weld is sufficient to promote TGSCC on the annealed 304 stainless steel canopy seal in a corrosive environment. The data from Westinghouse investigations indicate that the cause of the cracking observed in the annealed 304 stainless steel canopy seal is a combination of corrosive media, most likely chloride, and oxygen contamination present in the "dead end cavity" that is formed by the canopy seal (Reference 4). The industry indications of leaks in the seal welds have been characterized as pinholes or small cracks. There have been no industry reports of degradation of canopy seal welds resulting in significant leakage flow rates (Ref. 3). Considering the head adapter flange design, leakage through a crack in the non-pressure boundary seal weld would be expected to be limited by the load carrying component, the flange connection threads. Based upon the WCGS and industry experience, along with the fact that the canopy seal weld is not the load carrying part of the joint design, a gross failure on a lower canopy seal weld is unlikely to occur. If a gross failure of a seal weld does occur, it is expected that leakage would be recognized using indications typical of a small leak inside containment and would be subject to the unidentified leakage Technical Specification limitations. The canopy seal associated with penetration # 20 was determined to be leaking during the RV head inspection in RF20. This penetration is one of the 8 for capped latch housings. Historically, repairs using a Canopy Seal Clamp Assembly (CSCA) have been made to lower canopy seal welds at WCGS at the following locations: Penetration # Type 10 RV Levell Indication System 13 Spare 17 Active CROM 22 Spare 24 Spare 25 Spare 27 Spare 28 Spare CR Detail Report Page 7 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report 29 Spare The CSCA is designed to be installed remotely from above the nozzle housings. The installation can be accomplished with the reactor vessel head located either in the reactor head stand or on the reactor vessel. The above list of canopy seal repairs has been performed at different times during the operating life of the reactor at WCGS. This evaluation and probable cause identify that the leakage may be driven by stress corrosion cracking. Therefore, no pattern has emerged that would enable prediction of the next penetration seal weld crack, - leakage is based on the individual conditions associated with each canopy seal weld. A leaking canopy seal will allow boric acid and radioactive contaminants to leak from the primary system. The boric acid is non-corrosive to the piping and v essel cladding associated with the primary system (which contains the Borated Water). Boric acid leaking onto the hot external surface of the reactor vessel could result in corrosive damage. The extent of the damage depends on the extent, duration of the leak and moisture content of the residue on the carbon steel closure head. At the beginning of each outage, a boric acid walkdown at Mode 3 is conducted in accordance with STN PE-0400. Included in the STN PE-0400 inspection locations are the RV nozzle head adapter canopy seal welds . The STN PE-0400 inspection conducted at the beginning of RF20 did not identify the leakage from the canopy seal weld at head nozzle penetration

                          #20.

It should be noted that this leak was discovered during performance of STS PE-040E, "RPV Head Visual Inspection". The method of visual inspection was by a remote controlled crawler performing a VE inspection on the RPV head. In accordance with STS PE-040E, this VE inspection is conducted every third refueling outage. Although inspection of the canopy seal area is outside the required inspection scope, boric acid staining and a rust staining were noticed on the RPV head surface. This discoloration was traced to the overhead canopy seal area. As noted above, a head visual inspection is scheduled for each outage as part of STN PE-0400 and a head visual examination is scheduled for each refueling outage (either a bare metal visual examination, i.e., VE, every third refueling or a VT-2 with insulation removed at refueling outages between the VE). The boric acid walkdown inspections of STN-PE-0400, along with the visual examinations of STS PE-040E, and the end of outage performance of pressure test STS PE-040A by a VT-2 examination at NOP/NOT, provide several opportunities for discovery of a leaking canopy seal before the RV head would be damaged by the corrosive boric acid liquid. Past identification of leaking canopy seal welds have been identified by one of these inspections. However, the small extent of leakage indicated by the identified boric acid and rus t staining at penetration #20 has identified a weakness in the ability of these inspections and examinations to identify small leaks. Therefore, enhancements to one or more of these inspections are determined appropriate to be able to identify similar small leaks. When a suspected leaking lower canopy seal weld is identified following a visual examination as described above, WCGS has installed a Canopy Seal Clamp Assembly (CSCA). The proposed repair method to encapsulate the canopy seal weld with the CSCA has been evaluated by Westinghouse and determined to be an acceptable repair method. Wolf Creek has previously successfully performed this repair method on active penetration canopy seal leaks a nd multiple spare penetration canopy seal leaks. However, this leak was the first identified on a capped latch assembly and WCNOC did not have a canopy seal clamp assembly approved for installation on CR Detail Report Page 8 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report this configuration. The CSCAs are designed and fabricated as Class 1 components. In the CSCA design, a large ring of Grafoil' (graphite) sealant that is compressed against the canopy seal weld region provides the barrier against fluid leakage. The Grafoil' is held in place by the seal carrier halves, top and bottom housings, and attachment cap screws. The CSCA is designed to be installed remotely from above the CROM housings. The installation is typically performed with the reactor vessel head removed from the reactor vessel and on the head stand. This approach is cost effective and minimizes the radiation dose during the repair process. For the leaking canopy seal weld at penetration #20, WCNOC prepared and issued CCP 012962, approving the use of a CSCA on penetration #20 capped latch assembly. A revision to CCP 012962 will be required to approved CSCAs on any capped latch assembly. CONCLUSIONS: Leakage at these RV canopy seal welds is an inherit design and configuration problem caused by Transgranular Stress Corrosion Cracking . Such leakage has been identified and repaired in the past by installation of canopy seal clamp assemblies. This approach of inspecting and repairing leaks is sufficient to preclude damage to the RV external surface as long as leaks are properly identified and repaired. The small extent of leakage indicated by the identified boric acid and rust staining at penetration #20 has identified a weakness in the ability of these inspections and examinations to identify small leaks. Therefore, enhancements to one or more of these inspections are determined appropriate to be able to identify similar small leaks. However, to preclude delays in repairing leaking canopy seal welds with CSCAs, WCNOC design documents need to have evaluated and approved CSCAs for all the identified configurations of attachments to the nozzle head adapters. WCNOC Engineering needs to complete a revision to CCP 012962 to approve use of CSCAs on additional head adapter configurations. CAUSE CONCLUSIONS (Not required for BGA, BOE) PROBABLE CAUSE STATEMENT: Austenitic st ainless steel in the presence of chloride/oxygen in the high pressure and temperature environment is susceptible to Transgranular Stress Corrosion Cracking (TGSCC). The residual stress associated with the canopy seal weld is sufficient to promote TGSCC on the annealed 304 stainless steel canopy seal in a corrosive environment. The cause of the cracking observed in the annealed 304 stainless steel canopy seal is concluded to be TGSCC from a combination of corrosive media , most likely chloride, and oxygen contamination present in the "dead end cavity" that is formed by the canopy seal (Reference 4). SUPPORTING FACTS: Westinghouse performed destructive examination and hardware fa ilure analysis on a number of lower canopy seal welds removed from nuclear power plants in the late nineteen eighties (WCAP-12088). In addition, Westinghouse assisted WCNOC personnel in performing a hardware failure analysis report for leaking canopy seal welds (References 4 and 5). These investigations concluded that the failure mechanism was TGSCC . The cracking was identified both in the seal welds and in the base metal of the seal weld joints. All cracking was initiated from the interior of the joint. No sensitization of the material was identified in the components examined. Very low levels of chloride contamination were noted in water samples obtained from removed weld joint areas, as well as residues on the threaded surfaces that were analyzed. Westinghouse determined that the CR Detail Report Page 9 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report residual stresses associated with the seal welding process, combined with the very low levels of chloride in the oxygenated stagnant region, were sufficient to promote TGSCC on the joint. The Westinghouse hardware failure analysis also included examination of some threaded joints that were removed along with the lower canopy seal welds. There was no evidence of corrosion or cracking on any of the threaded joints that were examined. The industry indications of leaks in the seal welds have been characterized as pinholes or small cracks. There have been no industry reports of degradation of canopy seal welds resulting in significant leakage flow rates (Ref. 3). Considering the head adapter flange design, leakage through a crack in the non-pressure boundary seal weld would be expected to be limited by the load carrying component, the flange connection threads. Cause: Extent of Cause: Safety Significance: CR Detail Report Page 10 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Actions Taken: ACTIONS TAKEN: Engineering approved (Reference 2) the installation of a canopy clamp seal assembly at the capped latch penetration #20. Additionally, Engineering approved (Reference 2) the installation of a new design "short" canopy seal clamp assembly as a direct replacement for the old design "long" canopy seal clamp assembly. The short CSCA can be installed in any location where the long CSCA is currently approved for installation . The proposed repair method to encapsulate the canopy seal weld with the CSCA has been evaluated by Westinghouse and determined to be an acceptable repair method. Wolf Creek has previously successfully performed this repair method on several active penetration canopy seal leaks and multiple spare penetration canopy seal leaks. As of Jan. 15, 2008, Westinghouse information had identified 150 CSCAs installed on both leaking (56) and non-leaking (94) lower canopy seal welds on head penetrations in 35 plants since 1988. Westinghouse also reports there have been no re-occurrences of leakage from previously leaking welds (Reference 2). The "short" canopy seal clamp assembly was successfully installed on penetration #20. After installation of the CSCA, it was found that the "dummy can" would not fit over the CSCA. The "dummy can" was modified prior to installation. An inspection will be performed during performance of pressure test STS PE-040A to determine if the CSCA installation was successful. ACTIONS PLANNED:

1) The small extent of leakage indicated by the identified boric acid and rust staining at penetration #20 has identified a weakness in the ability of the normal inspections of the RV nozzle penetrations to identify small leaks. Therefore, enhancements to one or more of these inspections are determined appropriate to be able to identify similar small leaks. The following inspection procedures should be evaluated for enhancements to assure small leaks at the canopy seal welds are identified early in the outage to allow time for repairs (installation of CSCAs): STN-PE-040D, STS PE-040E, and STS PE-040A.

2) Westinghouse will be providing follow-up documentation that will approve CSCAs on any capped latch housings. Engineering plans to issue a revision to CCP 012962 to update the appropriate WCNOC documents after receipt of the Westinghouse documentation.

Information Sources: INFORMATION SOURCES ATTACHMENTS:

1. Docket No. 50-482: 30-Day Response for NRC Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity" (CT 02-0029).

2. Change Package# 012962, "Canopy Seal Clamp Assembly in Capped Latch Location", 3/25/2015.

3. WCAP-12088, "Metallurgical Failure Analysis of Leaking Canopy Seals" , C.

M. Pezze (January, 1989).

4. Hardware Failure Analysis Request (HFAR No. MA 92-008, March 1992).

5. Westinghouse letter SAP-92-148 transmitting Westinghouse Report MED-PCE-11788.

Review and Approvals QA Review: Rad Protection Review: Independent Review: CR Detail Report Page 11 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report CARB Review: CAP Liaison: APPROVED BRDORAT - 04/20/2015 Supv. Approval: The evaluation was performed by David Giefer and Richard Gimple. It was intended to attached all the documents to this CR but the folder in Net FYI is locked and will not allow any files to be added. Everything was copied into the applicable categories. The evaluation has been reviewed and approved with the required actions created and assigned. Supt. Approval: Manager Approval: V.P. Approval: CEO Approval: Extentions

  # of Extentions:                0 Extention Notes:

Supv. Ext. Approval: Supt. Ext. Approval: Manager Ext. Approval: V.P. Ext . Approval: CEO Ext. Approval: Other Related Information Assignment Notes: Updated By Last Updated

References:

EVAL St atus & Due Date History: BRIAN D. DORATHY 04/20/2015 COMPLETE BRIAN D. DORATHY 04/20/2015 ACC/ASG RHONDA G. NEILSON 03/21 /2015 INPROG 04/20/2015 CR Detail Report Page 12 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Plan and Actions Plan Assiginment #: 00093697 -03 Status: COMPLETE Status Date: 12/10/2015 Plan

Subject:

Penetration 20 Canopy Seal Weld Leakage Age In Days: 234 Assigned To Name: DORATHY, BRIAN D Assigned To Organization: 4050050-ENGINEERING DAILY WORK CONTROL SUPV - DORATHY

Description:

Action Ass.ignment #: 00093697-03-01 Action Due Date: 11/20/2015 Status: CANCELED Status Date: 05/20/2015 Action

Subject:

Revise STS PE-040E Age In Days: 30 Assigned To Name: GIEFER, DAVID L Assigned To Organization: 4050050-ENGINEERING DAILY WORK CONTROL SUPV - DORATHY

Description:

Revise procedure STS PE-040E to include a camera on a pole that could be manipulated within the CROM tubing to identify canopy seal issues. Action Category: REMEDIAL LTCA: Schedule Requirement: RCMS#: Commitment: Commit To Agency: Work Performed: Review and Approvals Independent Review : CARB Review: CAP Liaison: Supv. Approval: Supt. Approval: Manager Approval: V.P. Approval: CEO Approval: Extensions

   # of Extensions:                  0 Extension Notes:

Supv. Ext. Approval: Supt. Ext. Approval: CR Detail Report Page 13 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Manage r Ext. Approval: V .P. Ext. Approval: CEO Ext. Approval: Action Ass.ignment #: 00093697-03-02 Action Due Date: 12/11/2015 Status: COMPLETE Status Date: 12/10/2015 Action

Subject:

Revise Inspection Procedure Age In Days: 234 Assigned To Name: DORATHY, BRIAN D Assigned To Organization: 4050050-ENGINEERING DAILY WORK CONTROL SUPV - DORATHY

Description:

The small extent of leakage indicated by the identified boric acid and rust staining at penetration #20 has identified a weakness in the ability of the normal inspections of the RV nozzle penetrations to identify small leaks . Therefore, enhancements to one or more of these inspections are determined appropriate to be able to identify similar small leaks. The following inspection procedures should be evaluated for enhancements to assure small leaks at the canopy seal welds are identified early in the outage to allow time for repairs (installation of CSCAs): STN-PE-0400, STS PE-040E, and STS PE-040A. Action Category: REMEDIAL LTCA: Schedule Requirement: RCMS#: Commitme*nt: Commit To Agency: Work Performed: y The procedure that is required to provide a detailed inspection of the RF head is STS PE-040E, "R PV Head Visual Inspection". The other two procedures (STN-PE-0400 and STS-PE-040A) do not provide detailed inspections. Therefore, STS PE-040E revision should close* this action. The revision of the procedure included providing required updating for the Interval 4 ISi Program. In addition. some wording corrections/modifications were made. The main additions to the procedure to support the CR 93697 resolution Include:

1) Addition of a note to include the requirements of a general inspection of the areas above the head to Include the canopy seals (Note included between Section 8.0 and 8. 1) as follows:

                                                           "Although not part of the pressure test inspection, the visual inspection (either VE or VT-2) should include a general inspection of the RV head areas above the CROM nozzle/head interface penetrations that would Include the canopy seals. The results of these inspections should be included in Attachment A - Test Performers Comments"

2) A step to identify acceptance of the visual documentation (Sections 8.1.3 and 8.2.3) as follows:

                                                           "Visual documentation (photographs and/or video) of examination accepted by Boric Acid Corrosion Control Program Owner."

3) A step to assure a good Inspection of the inner areas of the head that are difficu lt to view (Step 8.2.1, Item 1) as follows:

CR Detail Report Page 14 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report "To inspect inner areas of the head that are difficult to view for performance of a VT-2 inspection, a remote viewing device (camera, etc.) mounted to an extension tool can be used. Record approximate azimuth of the access points on Attachment D along with the results of the VT -2 exam using a remote viewing device." With the above revisions to Procedure STS PE-040E, the CR action 93697 02 should be closed. Review and Approvals Independent Review: CARB Review: CAP Liaison: Supv. Approval: APPROVED Supt. Approval: Manager Approval: V.P. Approval: CEO Approval: Extensions

  # of Extensions:                  0 Extension Notes:

Supv. Ext. Approval: Supt. Ext. Approval: Manager Ext. Approval: V.P. Ext. Approval: CEO Ext. Approval: Action Ass.ignment #: 00093697 03 Action Due Date: 10/30/2015 Status: COMPLETE S*tatus Date: 10/26/2015 Action

Subject:

Track Revision to CP 12962 Age In Days: 189 Assigned To Name: DORATHY, BRIAN D Assigned To Organization: 4010020-CIVIL-MECHANICAL DESIGN ENG - CURTEN

Description:

Westinghouse will be providing follow-up documentation that will approve CSCAs on any capped latch housings. Engineering plans to issue a revision to CCP 012962 to update the appropriate WCNOC documents after receipt of the Westinghouse documentation. Action Category: ENHANCEMENT LTCA: Schedule Requirement: RCMS#: CR Detail Report Page 15 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Commitment: Commit To Agency: Work Performed: y FCN 012962 Rev. 02 has been completed to accept the Westinghouse documents into the doucment control system. Additionally, CCN BB-S-018-000-CN004 has been issued for the change to the BB-S-018 calculation. No further action is required. Review and Approvals Independent Review: CARB Review: CAP Liaison: Supv. Approval: APPROVED Supt. Approval: Manager Approval: V .P. Approval: CEO Approval: Extensions

  # of Extensions:                 0 Extension Notes:                y                                    BRDORAT - 09/17/2015 Request extension to 10/30/2015.

The change package revision has been prepared but it is awaiting a review. Additional time is required due to higher priority items preventing the review of the change package. The RF21 design milestone and WANO preparations have taken a priority . This is capturing information provided from Westinghouse in WCNOC documents that will not be needed until the next refueling outage, which the extension due date is well before the next outage. There are not any hazards associated with this extension since it will still be completed prior to next outage. There are no interim actions required for this extension. This action is not addressing the restoration of full qualification of an SSC. Supv. Ext. Approval: APPROVED BRDORAT - 09/17/2015 Peer review provided by John Ashley. 9/17/15 Supt. Ext. Approval: Manager Ext. Approval: V .P. Ext. Approval: CEO Ext. Approval: Other Related Plan and Action Information CR Detail Report Page 16 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Plan Assignment Notes : Updated By Last Updated 00093697-03 93697-03-01 was created in error. Changes could not be made after the BR DORAT 05/20/2015 action was progressed to notify so the action was canceled. Action Ass.ignment Notes: 00093697-01-01 Action auto-closed based on work completion for WO 15-399448-000. INDUS 04/17/2015 00093 697-01-02 Action auto-closed based on work completion for WO 15-399448-001 . INDUS 03/29/2015 00093697-01-03 Action auto-closed based on work completion for WO 15-399448-002. INDUS 03/24/2015 00093697-01 -04 Action auto-closed based on work completion for WO 15-399448-003. INDUS 04/15/2015 00093697-01-05 Action auto-closed based on work completion for WO 15-399448-004. INDUS 03/31/2015 00093697-01-06 Action auto-closed based on work completion for WO 15-399448-005. INDUS 03/31/2015 00093697-01-07 Action auto-closed based on work completion for WO 15-399448-006. INDUS 03/22/2015 00093697-01-08 Action auto-closed based on work completion for WO 15-399448-007. INDUS 03/30/2015 00093697-01 -09 Action auto-closed based on work completion for WO 15-399448-008. INDUS 03/22/2015 00093697-01-10 Action auto-closed based on work completion for WO 15-399448-009. INDUS 04/10/2015 00093 697-01-11 Action auto-closed based on work completion for WO 15-399448-010. INDUS 05/01/2015 00093697-01-12 Action auto-closed based on work completion for WO 15-399448-011 . INDUS 03/30/2015 00093697-01-13 Action auto-closed based on work completion for WO 15-399448-012. INDUS 03/20/2015 00093697-01-14 Action auto-closed based on work completion for WO 15-399448-013. INDUS 04/09/2015 00093 697-01-15 Action auto-closed based on work completion for WO 15-399448-014. INDUS 03/21/2015 00093697-01-16 Action auto-closed based on work completion for WO 15-399448-015. INDUS 03/21/2015 00093697-01 -17 Action auto-closed based on work completion for WO 15-399448-016. INDUS 03/29/2015 CR Detail Report Page 17 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report 00093697-01-18 Action auto-closed based on work completion for WO 15-399448-017. INDUS 03/29/2015 00093697-03-01 93697-03-01 was created in error. It was unknown that changes could be BRDORAT 05/20/2015 made after the action was progressed to notify so the action was canceled. Plan Completion Notes: Action Completion Notes: Plan Cross

Reference:

Type Number Sub Number Action Cross

Reference:

Plan Status and Due Date History: 00093697-03 Respons ible Person Date Updated St atus Due Date DORATHY, BRIAN D 04/20/2015 INPROG 12/1112015 DORATHY, BRIAN D 04/20/2015 ACC/ASG DORATHY, BRIAN D 12/10/2015 COMPLETE Action Status and Due Date Histo ry: 00093697-03-01 Responsible Person Date Updated Status Due Date DORATHY, BRIAN D 04/20/2015 INPROG 11/20/2015 DORATHY, BRIAN D 04/20/2015 NTFY/ASG DORATHY, BRIAN D 04/20/2015 CANCELED DORATHY, BRIAN D 05/20/2015 INPROG DORATHY, BRIAND 05/20/2015 NTFY/ASG DORATHY, BRIAN D 05/20/2015 CANCELED 00093697 02 Responsible Person Date Updated Status Due Date DORATHY, BRIAN D 04/20/2015 INPROG 12/1112015 DORATHY, BRIAND 04/20/2015 NTFY/ASG GIEFER, DAVID L 12/10/2015 ACC/ASG DORATHY, BRIAN D 12/10/2015 NTFY/ASG DORATHY, BRIAN D 12/10/2015 ACC/ASG DORATHY, BRIAN D 12/10/2015 COMPLETE 00093697-03-03 Responsible Person Dat e Updat ed Status Due Dat e DORATHY, BRIAN D 04/20/2015 INPROG 09/18/2015 DORATHY, BRIAN D 04/20/2015 NTFY/ASG DORATHY, BRIAN D 09/17/2015 10/30/2015 PANKASKIE, JASON M 10/08/2015 ACC/ASG DORATHY, BRIAN D 10/26/2015 NTFY/ASG DORATHY, BRIAN D 10/26/2015 ACC/ASG CR Detail Report Page 18 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report DORATHY, BRIAN D 10/26/2015 COMPLETE CR Detail Report Page 19 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Effectiveness Follow-up EFU Assignment#: EFU Due Date: Status: Status Date: EFU

Subject:

Age In Days: Assigned To Name: Assigned To Organization:

Description:

EFU Effective: Review and Approvals Independent Review: CARB Review: CAP Liaison: Supv. Approval: Supt. Approval: Manager Approval: V.P. Approval: CEO Approval: Extensions

 #of Extensions:

Extension Notes: Supv. Ext. Approval: Supt. Ext. Approval: Manager !Ext. Approval: V.P. Ext. Approval: CEO Ext. Approval: Other Related Information Assignment Notes: Updated By Last Updated CR Detail Report Page 20 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Cross

References:

EFU Status and Due Date History: CR Detail Report Page 21of28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Restore to Full Qualification RTFQ 00093697 -01 Status: COMPLETE Status Date: 05/07/2015 WR#: 15-111261 RTFQ

Subject:

STS PE-040E Penetration 21 Canopy Seal Weld Leakage lndicati RTFQ

Description:

During the performance of STS PE-040E evidence of leakage was identified at the canopy seal weld on penetration 21 . The indication appears on the east side of the seal weld face (as the RPV closure head sits). Boron staining is evident above, adjacent and below the seal weld indication. Rust staining is also prevalent at and below the indication running down the CROM nozzle and onto the top head surface. There is no evidence of degradation to the CROM or closure head surfaces. This seal weld indication appears to have been an active leak in the recent past. Photos/video available from QC (Jason Heffron). Recommend this seal weld be evaluated for installation of a seal weld clamp assembly. Equipment: RBB01 On-Line or Refuel: REFUEL CR Detail Report Page 22 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Operability: 3 OPER/DNC The initiator identified during the performance of STS PE-040E evidence of leakage at the canopy seal weld on penetration 21. The indication is on the east side of the seal weld face (as the RPV closure head sits). Boron staining is evident above, adjacent and below the seal weld indication. Rust staining is also prevalent at and below the indication running down the CROM nozzle and onto the top head surface. There is no evidence of degradation to the CROM or closure head surfaces. I reviewed the pictures that are located at K:\Data\NDE\Photos\RF-20\CRDM Head Inspection. In these pictures there are visible traces of dried boron and some small amounts of discoloration on the seal weld. As the initiator identified, there is no significant accumulation of boron or wastage of any carbon steel on the head penetration directly below the subject seal weld. Technical Specifications defines Pressure Boundary LEAKAGE as LEAKAGE through a nonisolable fault in an RCS component body, pipe wall or vessel wall. TS 3.4.13 contains the operating limits for RCS Operational LEAKAGE. In MODES 1 through 4, no pressure boundary is allowed, unidentified LEAKAGE is limited to 1 gallon per minute, identified LEAKAGE is limited to 1 0 gallons per minute, and primary to secondary LEAKAGE is limited to 150 gallons per day in any one Steam Generator. The Control Rod Drive Mechanism is what is used to raise, lower, and trip control rods. The internals of this mechanism is exposed to RCS pressure. The Drive Mechanism Latch Housing is internally threaded and torqued down onto a seating surface at the interface between the housing and the top of the Reactor Head Adapter. This connection is a mechanical joint and leakage via this pathway is not Pressure Boundary LEAKAGE as defined by Technical Specifications. The WCGS reactor vessel head and CROM assemblies are classified as ASME Boiler and Pressure Vessel Code Section Ill Class 1 items. The Reactor Vessel was designed and fabricated to the 1971 Edition through Winter 1972 Addenda and the CROM housing assemblies were designed and fabricated to the 1974 through Winter 1974 Addenda of Section Ill of the ASME B&PV Code. Section Ill paragraph NB-3671 .3 states that threaded joints in which threads provide the only seal shall not be used. The seal weld is not a structural part of the pressure boundary and is not required to meet the structural requirements of ASME B&PV Code, Section Ill, NB-3000. The threads are the load carrying part of the joint design. The industry indications and past operating experience at WCGS of leaks in the subject seal welds are pinholes or small localized cracks. These flaws have resulted in leak rates that are bound by the limits established in Technical Specification 3.4.13. Completed performances of STS BB-006 were reviewed from the last operating cycle and RCS leakage limits were not challenged. The Reactor Vessel and the subject CROM is OPERABLE but degraded due to the flaw in the lower seal weld. CR Detail Report Page 23 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report

References:

Technical Specifications 1.1, 3.4.13 and Bases; TR 3.4.17 and Bases; NRG Inspection Manual Part 9900, WCGS Correspondence CT 02-0029, Westinghouse Instruction and Operating Book for Magnetic Control Rod Drive Mechanism for Full-Length Control Rods, and STS PE-040E. Operations Focus List : Plant System: BB Risk Impact: Risk Review Complete: y WABRAND 04/04/2015 Risk Significance: HIGH Safety Function: RCS Integrity IOA

Conclusion:

A clamp has been installed. QC will perform PMT at NOP. IOA: Sources CAP: Work Orders: Margin Management: Ops Focus List: Single Point Vulnerability: System Health Report: Temporary Modification: Operational Decision Making: Maintenance Rule: MSPI: PDM Watch List: Regulatory Commitment: Other: Vulnerabilities Steam Generator Tube Rupture: Loss of Off-Site Power Rapid Load Reduction: Inadvertent Safety Injection: Fire/Flooding: CR Detail Report Page 24 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Inter-system LOCA: Loss ofRHR: Loss of $.pent Fuel Cooling: Load Reject: Steam Line Break: Loss ofESW: Measures Compensatory Measures: Monitoring Measures: Mitigations Measures: RTFQ Actions: CR Detail Report Page 25 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report 00093697-01-12 Status: COMPLETE WO#: 15-399448-011

Subject:

Remove/Replace cables and unistrut on head as required to su Notes: Action auto-closed based on work completion for WO 15-399448-011. 00093697-01-16 Status: COMPLETE WO#: 15-399448-015

Subject:

PE has request a SWO for new stock item that is SR stock ite Notes: Action auto-closed based on work completion for WO 15-399448-015. 00093697-01-06 Status: COMPLETE WO#: 15-399448-005

Subject:

Remove/Reinstall pipe support BB17H505/251 (HV-6) to support Notes: Action auto-closed based on work completion for WO 15-399448-005. 00093697-01-14 Status: COMPLETE WO#: 15-399448-013

Subject:

Contingency -Access lower shroud to assist clamp installati Notes: Action auto-closed based on work completion for WO 15-399448-013. 00093697-01-15 Status: COMPLETE WO#: 15-399448-014

Subject:

NS92250263 is not tied to asset RBB01 BOM Notes: Action auto-closed based on work completion for WO 15-399448-014. 00093697-01-02 Status: COMPLETE WO#: 15-399448-001

Subject:

STS PE-040E Penetration 21 Canopy Seal Weld Leakage lndicati Notes: Action auto-closed based on work completion for WO 15-399448-001 . 00093697-01-01 Status: COMPLETE WO#: 15-399448-000

Subject:

STS PE-040E Penetration 21 Canopy Seal Weld Leakage lndicati Notes: Action auto-closed based on work completion for WO 15-399448-000. 00093697-01-20 Status: CANCELED WO#: 15-399448-019

Subject:

Residue on CROM nozzles, adapters and housings During RF20 Notes: 00093697-01-13 Status: COMPLETE WO#: 15-399448-012

Subject:

Engineering to evaluate the use of the following NS Stock It Notes: Action auto-closed based on work completion for WO 15-399448-012. 00093697-01-18 Status: COMPLETE WO#: 15-399448-017

Subject:

Engineering to evaluate !riming the dummy can for capped lac Notes: Action auto-closed based on work completion for WO 15-399448-017. 00093697-01-11 Status: COMPLETE WO#: 15-399448-010

Subject:

QC is requested to perform a Pre-Service VT-2 examination on Notes: Action auto-closed based on work completion for WO 15-399448-010. 00093697-01-10 CR Detail Report Page 26 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Status: COMPLETE WO#: 15-399448-009

Subject:

QC is requested to perform a Pre-Service VT-3 examination on Notes: Action auto-closed based on work completion for WO 15-399448-009. 00093697-01-09 Status: COMPLETE WO#: 15-399448-008

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QC is requested to perform a Pre-Service VT-1 examination on Notes: Action auto-closed based on work completion for WO 15-399448-008. 00093697-01-08 Status: COMPLETE WO#: 15-399448-007

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Remove/Replace grollnd cable on Plenum. Notes: Action auto-closed based on work completion for WO 15-399448-007. 00093697-01-04 Status: COMPLETE WO#: 15-399448-003

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Remove/Replace coil stacks at penetration 41 or 9 as require Notes: Action auto-closed based on work completion for WO 15-399448-003. 00093697-01-03 Status: COMPLETE WO#: 15-399448-002

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QC is requested to perform a PT examination on the Canopy Se Notes: Action auto-closed based on work completion for WO 15-399448-002. 00093697-01-05 Status: COMPLETE WO#: 15-399448-004

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Remove/Reinstall Upper Plenum to support installation of a c Notes: Action auto-closed based on work completion for WO 15-399448-004. 00093697-01 -07 Status: COMPLETE WO#: 15-399448-006

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QC is requested to verify/document the Canopy Seal Weld leak Notes: Action auto-closed based on work completion for WO 15-399448-006. 00093697-01-19 Status: CANCELED WO#: 15-399448-018

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STS PE-040E CRDM nozzle to head interface cleaning During th Notes: 00093697 CR Detail Report Page 27 of 28 12/10/2015 11:42:24PM

Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Reportabllity Evaluation Report RER Status: Status Date: Age: Due Date:

Subject:

Date/Time of Discovery:

== Description:==

SCREENING/NOTIFICATIONS (Completed by Shift Manager): Potentially Reportable: RER Number: Per (list applicable reporting criteria met): Person Contacted: Corporate Services Notified: ENS Reportability Determination per 10 CFR 50.72: ENS Worksheet completed and attached: Continuous open channel required: Shift Manager Approval: Last Updated: DISPOSITION (Completed by Licensing): LER#: Ltr. Nlumber: Submittal Date: Event Evaluation: Reportability Evaluation Performed by: REVIEW and APPROVAL (Non-Reportable Events Only) Supervisor Licensing Approval: Last Updated: Manager Regulatory Affairs Approval: Last Updated: ENS Retraction needed: Report Crlterl a CR#: 00093697 CR Visible: y EVAL Visible: y PLAN Visible: y EFU Visible: y Non QA Visible: N RER Visible: y CR Detail Report Page 28 of 28 12/10/2015 11:42:24PM

From: Reimer. Lisa To: Sjngal Balwant: Alley Dayjd; Werner Greg; Pascarelli Robert Subj ect: FYI: Info: Internal NBC Call to Discuss Wolf Creek Relief Request Date: Friday, October 28, 2016 12:43:00 PM Gentlemen, George gave Michele the heads up that the Wolf Creek Relief Request scope may need to be expanded. She is leaving the office today at 3:30 pm, but is fine with any briefing information to be sent to her by email. Please cc: George Wilson as well - he also had to leave early today, but is available by phone. 5 George cell: r_.... l<_l______. George home: 1.... (b-)(6-) _ _ _..... I am also available in the office until 3 pm today, and by cell after: 410-733-9659 Thanks! Lisa Lisa Regner Sr. PM NRR/DORL/LPL4-1 301-415-1906 08D08 From: Collins, Jay Sent: Friday, October 28, 2016 11:34 AM To: Singal, Balwant <Balwant.Singal@nrc.gov>; Al ley, David <David.Alley@nrc.gov>; Kal ikian, Roger <Roger.Kalikian@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Taylor, Nick <Nick.Taylor@nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>; Regner, Lisa <Lisa.Regner@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

Subject:

RE: Internal NRC Call to Discuss Wolf Creek Relief Request ASME Code Case N-729-1

Note 1 The VE sha ll consist of the following: (a) A direct exam ination of the bare-metal surface of the entire outer surface of the head, including essentially 100% of the intersection of each nozzle w ith the head. If welded or bolted obstructions are present (i.e., mirror insulation, insulation support feet, shroud support ring/lug), the exam ination shal l include =95% of the area in the region of the nozzles as defined in Fig. 1 and the head surface uphill and downhill of any such obstructions. The exam ination may be performed w ith insulation in place using remote equipment that provides resolution of the component metal surface equ iva lent to a bare-meta l direct examination . (b) The examination may be performed with the system depressurized. (c) The examination shal l be performed with an illumination level and a sufficient distance to al low resolution of lower case cha racters not greater tha n 0.105 in. (2.7 mm) in height.

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Origina I Appointment-----

From: Singal, Ba lwant Sent: Friday, October 28, 2016 10:17 AM To: Col lins, Jay; Alley, David; Kalikian, Roger; Tsao, John; Drake, James; Taylor, Nick; Proulx, David; Cumbl idge, Stephen; Regner, Lisa; Werne r, Greg; Anchondo, Isaac; Kopriva, Ron; Thomas, Fabian

Subject:

Internal NRC Cal l to Discuss Wolf Creek Relief Request When: Friday, October 28, 2016 11 :00 AM-12:00 PM (UTC-05:00) Eastern Time (US & Canada). Where: Dave Al ley's Office Dave Alley and me received a call from Wolf Creek (Cyndia and Jaimme McCoy) at 9.30 this morn ing. An internal NRC staff meeting is required to discuss path forward based on information provided during the cal l. Bridge No. Info.

866-624-3402 6 Passcode: 1..... (b-)1 _) _ ___. Lisa: You wi ll need to use Passcode1.... 15_)16_) _ _.I (as initiator of the ca ll). I was not bale to search for conference rooms from home. 6 l(b-)(_)_ _ _ _ _ _ _1for about 3 hours and Lisda wi ll be supporting I will be out-of-office .... this ca ll. I can be contacted I a~....'.b_l(_5l _ _ _ _ for any questions. Thanks.

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Ultrasonic inspections performed to see flaws in welds and penetration tubes need to scan interference fit

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above and below the weld, as No Leak the weld is not straight. This scanning, as an unintentional byproduct, produces images from the ultrasound reflecting from the interference fit region. It did not take long for people to figure out that leaking nozzles produced different patterns in Leak the interference fit than non-leaking nozzles.

So, what is going on? Some reflection and some transmission will occur at the interference fit. The amount of sound reflected is affected by the local tightness of the fit, the local smoothness of the metals, and the local presence of boric acid.

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Ultrasound is sensitive to changes in the interference fit as the two metal surfaces are in tight contact. The surfaces were not made mirror-smooth prior to the interference fit, so some odd features will be present. Even so, notches, deep scratches, and a contractor scribing "PNNL in an interference fit can be clearly detected.

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Interference fits without leaks can still have odd features, depending on the smoothness and how the data was collected. False positives are possible if there are gouges and false negatives are possible if thee is little boric acid present. Interference fits with no leakage present

Leaks can produce odd patterns in the ultrasonic examinations of the interference fit. The random-looking patterns imaged by the volumetric leak path assessments can be reproduced . The general pattern remains the same, although different frequencies or methods (Zero degree vs. TOFD) may result in some differences. Westinghouse Data PNNL 2.25 MHz Data Wetted Side Wetted Side

In this case PNNL used a 5 MHz zero-degree probe to inspect the interference fit. Their results closely match industry scans of the same nozzle, with higher resolution and greater sensitivity.

The patterns in the UT images are apparently caused by the presence and absence of boric acid deposits that couple ultrasound through the interference fit. 135 Degrees

The High resolution data closely matches the boric acid pattern in Nozzle 63 from North Anna. Reflections come from areas with little or no boric acid and areas with more boric acid are detectable as areas of greater transmission.

Conclusions Volumetric Leakage Path Assessments can be effectively used to detect boric acid in the interference fit Volumetric leak path Assessments can give ambiguous results, but has been largely reliable ASME has decided not to qualify Volumetric Leakage Path Assessments Further Reading:

• Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation
• NUREG/CR-6996 Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechanism Penetrations
• Materials Reliability Program: Volumetric Leak Path Assessment for Vessel Upper Head Penetrations (MRP-249)}}