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)

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From: Taylor, N ick 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 Pic tu 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 ad d i ti on to boron. Will be really interesting to see what th ey find when the insulation co me s off. Nick From: Janicki, Steven Sent: Thursday, September 15, 2016 9:03 AM To: T aylor, Ni ck <Nick.T aylor@nrc.gov>

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

Dodson, Do u glas <Do u g la s.Dodson@nrc.gov>;

Thomas, Fabian <Fabia n.Thomas@nrc.gov>

Subject:

Wolf Creek Pictures Ni c k , 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 yo u can quickly scan them if you would like at some point (vice having to click the m all individually on Certrec).

S t eve. Respectfully, Steve Janicki Nucl ear Regulatory Commiss i on RIV -Divis io n of reactor Projects Branch B PE (0) 817-200-1457 (C)l (b)(6) I Steven.j an i cki@nrc.gov

. .. .... : : .. * . . -. : .. . -.

From: Sent: To: Cc:

Subject:

Tsao, John 15 Sep 2016 14:14:49 -0400 Alley, David;Hoffman , Keith;Kalikian, Rog e r Hsu, Kaihwa;li, Yong 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 "mec hanical joint", not as a " mechanical clamp". This is because NB-3671.7 permits the i nstalla t ion of mechanical joints (see Keith's email below). NB-3671.7 S l eeve Coupled and Other Patented Joints. M echanica l joint s , for which no s t andar d s exist, and other pate nt ed jo int s m ay be u s ed provided the r e quir e m en ts o f (a), (b), an d (c) below are met. (a) Pr ovision i s m ade to pre ve nt se paration of the joints und e r all Service Loadings. (b) Th e y arc acc e s s ibl e for maintenance , r emova l , and r e plac e ment afte r serv i c e. (c) Either of the following two crite r ia are m et. (I) A prototyp e joint has b e e n s ubject e d t o performance te s t s to det e rm in e the s afoty of th e jo i n t und e r s imul ated s erv i c e conditions.

Wh e n vibration , fa ti g u e , cyclic conditions , l ow temp eratu re , t herma l expansion, or h ydrau lic s ho ck i s a nti c ip ated , the app li cab l e co nditi ons s hall b e in cor p ora t e d in the te s t s. T h e m e c h a ni ca l joints s hall be s ufficiently l eak ti ght to sa ti s f y th e r eq uir e m e nt s of the De s i g n Spec ifi cations. (2) J oin t s are designed in accordance with t he rules o f NB-32 00. A "mechanical clamp" as per ASME Section XI, Appendix IX or Appendix W, is n ot pe r mitted 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

>; T sao, John <Jo hn.T sao@nrc.gov>;

Kalikian, Roger <Roger.K alikian@nrc

.gov> Cc: Hsu, Kaihwa <Kaih wa.H su@nrc.gov>

Li, Yong <Yong.li@nrc.gov>

Subject:

FW: M echanica l Clamp in ASME Se ct ion Ill Based on Ro b ert's view, be lo w, i t appears th at t h e clamp i s acceptable p er the cons t ruction code. Thi s wou l d appear t o m ake t he use o f the cla m p a code repair as it i s i n accordance with t he c on st ru ctio n code. Thi s wo ul d a p pear t o mea n that th e p lant ca n i n s t all a nd l ea v e t h e clamps on forever and that w e have no regulatory h oo k (othe r than, po t ent i a ll y, the condit ion o f t he th r eads for t h is instance based on t he extent of leakage).

Any thoughts?

D ave From: H s u , Kaihwa Sent: Thursday, September 15, 2016 8:25 AM To: Alley, David <Dav i d.Alley@nrc.gov

> Cc: Li, Yong <Yong.Li@nrc

.go v>

Subject:

RE: Mechanical Clamp in ASME Sectio n I ll D ave: I don't see a n y problem fo r the paten t ed clamp to be used over t op of or i ginal joi n t as l ong as the repai r meets AS M E Sec t ion I l l Code cr i ter i a.

Robert From: Li, Yong Sent: Thursday, September 15, 20 1 6 7:29 AM To: H su, K a ihwa <Kaihwa.Hsu@nrc.gov>

Subject:

FW: Mechanical Clamp in ASME Section Ill Please respond to Dave. From: Alley, David Sent: Wednesd a y, September 14, 2016 8: 19 PM To: Hoffman, Keith <Keith.Hoffman@nrc

.gov>; T sao, John <J ohn.Tsao@nrc.gov>; K a l ik i an, Roger <Roger.Kalikian@nrc

.gov>; Li, Yong <Yong.Li@nr c.gov>

Subject:

RE: Mechanical Clam p 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 Il l help. NB-367 1.7 S l eeve C oupl e d and Other P ate nt ed J o in ts 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, Septembe r 14, 20 16 8:08 PM To: Alley, David <Dav i d.Alley@nrc.gov>; Tsao, John <John.T s ao@nrc.gov>; Kalikian , Roger <Roger.Ka l ikian@nrc.gov

>

Subject:

RE: Mechanical Cla m p in ASME Section Ill I believe that is probably a question for Yong Li's branch. Obviously the licensees and ABB -C E/West i nghouse believe i t can be and that is wha t the paten t says it was des i gned to do. From: Alley, David Sent: Wednesday, Septembe r 14, 20 16 4:19 PM To: Hoffman, Keith <Keit h.Hoffman@nrc

.gov>; T sao, J ohn <John.Tsao@nr c.gov>; Ka lik i an, R oge r <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, 20 16 6:39 AM To: Tsao, John <John.Tsao@n r c.gov>; Alley, David <David.Alley@nrc.gov

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

>

Subject:

RE: Mechanical Clamp in ASME Section I ll This was t he section that was referenced as the app l icab l e Section Il l paragraph in one of t he docume nt s I looked at yes t erday. Specifically N B-367 1.7 was refe r enced. N B-36 7 1.3 shows t he t hreaded jo i nt and the requirement fo r the we l d. NB-3670 SP EC IAL PIPING R EQUI R EMENTS N B-3671 Se lec t ion and Limita t ion of Nonwelded P ipin g Join ts The type of piping joint used s hall be s uitabl e for the De s ign Loading s and s hall be s elect e d with con s ideration of joint ti ghtness , mechanica l strength , and t he nature of the fluid handled. Pipin gjoints shall conform to the requirements of this Sub s ection with leak ti g htn ess being a con s ideration in s e l e ction and de s i g n of joint s for piping s yst e m s to s ati s fy the requirements of the Design Specificat i ons. NB-3671.1 F lan ged J oints. Flan g ed joints are p e rmitt e d. NB-3671.2 Expa nd e d Joints. Expanded joints sha ll not be used. NB-3671.3 T hr eaded Joint s. Threaded joint s in which the thread s provide the only s ea l shall not be u s ed. If a seal weld i s employed as the s ea lin g medium , the s tre ss analy s i s of the joint mu s t inc.ludc the s tre ss e s in the weld re s ulting from the relative deflections of the mated parts. NB-367 1.4 F lar e d , F lar e l ess, and Co mpr ess i on Joints. Flared , flarel ess, and compr es sion t ype tubin g fittin gs may b e used for tubing sizes not exceeding 1 in. 0.D. (25 mm) within the limi tations of applicable standa rd s and specifications li s ted in Table N C A-7100-1 and r e quirement s (b) and (c) below. I n the ab s enc e of s uch s tandards or s p e cificat i on s, th e D es i gner sha ll determine that the type of fittin g se l ected is adequa t e and safe for the D es i gn L oadings in accordance w ith the requirements of(a), (b), and (c) be l ow. (a) The pre ss ure de s i g n s hall me e t the r e quirement s of NB-3649. (b) Fittings and their joints shall be su it able for the tubing with which they are to be used in accordance with the minimum wall thickn es s of th e tubing and m e thod of a sse mbly re c ommend e d by th e manufacturer. (c) Fittings sha ll not be used in serv i ces that exceed the manufacturer

's maximum pressur e-temperature recommendations.

NB-3671.S Ca ulked Joint s. Caulk e d or lead e d joint s shall not b e u s ed. NB-3671.6 Brazed and So ld e r e d Joints. (a) Brazed Joint s ( 1) Brazed joint s of a maximum nominal pipe s i ze of I in. (DN 2 5) may be used only at dead end instrument connect i ons and in special applications where space and geometry c onditions prevent the use of joint s permitted under N B-366 1.2 , NB-3661.3 , and NB-3671.4. Th e d e pth of s ock e t s hall be at l e a s t equal to that required for s ocket weldin g fittin gs and sha ll be of sufficient depth to develop a rupture strength equa l to that of the pipe at Design Temperature (N B-4500). (2) Bra z ed joint s that depend upon a fillet rather than a c apillary type filler addition arc not acceptabl e. (3) Brazed joint s sha ll not be used in sy s tems containing flammable flui d s or in areas where fire hazard s are involved. (b) Solder e d Joint s. S o lder e d j o ints s h a ll n o t b e u s ed. NB-3671.7 S l eeve Co upl ed and Other Patented Joints. Mechanical joint s, for which no s t andards exist, and other patented joints may be used provided th e requirem e nt s of(a), (b), and (c) below are m e t. (a) Provi s ion i s made to pr e vent se paration of th e joint s under all Service Loading s. (b) They are accessible for maintenance , removal , and replacement after service. (c) Eith e r of the following two crite r ia ar e m e t. (1) A prototype joint has been subjected to performance tests to determine the safety of th e jo i nt under simu lat e d service condition s. Wh e n vibr a tion , fati g u e, c yclic c ondition s, low t e mp e ratur e, t h e rmal e xpan s ion , or hydraulic s ho ck i s antic ip ated , the applicable conditions s hall be incorporated in the tests. The mechanical joints shall be s uffic i ently l eak tight to s ati s fy th e r e quirem e nt s of t he Desi g n Specifi c ation s. (2) Joints are designed in accordance w ith the rule s of NB-32 00. 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.Al l ey@nrc.gov>; Hoffman, Keith <Ke i th.Hoffman@nrc.gov

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

>

Subject:

Mechanical Clamp in ASME Section Ill I did a word searc h of "Clamp" in NB and NC sect ion s of th e 2007 ed ition an d 2013 e dition of the ASME Code, Section Ill. There were 4 hits in both ed ition s. NB-113 2.1-the clamp in this article is r e l ated to pipe attachment (the clamp used for pipe s upport s) NB-341 1.l(d) pump clamp NB-3651.3 pipe clamp as in pip e supports such as hangers or snubbers tha t u se clamps. NB-4231---clamps used i n welding operations (when we l ding 2 pieces of pipe, welders u se clamps) So ASME SEct i on Ill does not have requirements or specificatio n for the m ec h an i cal clamp app l ication that was used on the CRDM canopy sea l at Wolf Creek.

From: Sent: To:

Subject:

Collins, Jay 6 Sep 2016 20:34:47 +0000 Drake , Jame s RE: Wolf Creek Boric Acid Leaking on H ead 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 Cre e k Boric A cid Leaking on He ad 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 <J ames.Drake@nrc.gov>

Subject:

Wolf Creek Boric Acid Leak ing on Head Greetings , Catchi ng this issue from the s id e lin es , 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*Žt know the in spec tion requirements for Wolf Creek this refueling outage , but I figure they are at l east goi n g to have to clea n the head for a VT-2 in spect ion. Cleaning th e h ead in to o aggressive of a manner can invalidate the v i sual inspect i on and may then trigger a vo l l umetric inspection. Ju s t a heads up for a problem they may not be thinking abo ut , using l essons l earned that Region IV caught earlie r at Fort Calhoun. Thi s was l saaca*Žs issue at Fort Calhoun , so I am sure he has all the fine details. Ju st trying to be h e lpful , Jay From: Hoffman, Keith Sent: 6 Sep 2016 15:36:41 -0400 To: Alley, David;Collins, Jay;Tsao, John;Dav i s, Robert

Subject:

FW: OpE -RCS LEAKAGE RESULTS I N 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 Subj ect: OpE -RCS LEAKAGE RESULTS IN TECHNICAL SPECIFICATIONS SHUTDOWN The following EN is provided for your information.

EN 52218-WOLF CREEK-TECHNICAL SPECIFICATl l ON REQUIRED SHUTDOWN While operat ing in MODE 1 a t 100 percent rated thermal power a nd placing Excess Letdown in service for Reactor Coolant System (RCS) leak detection, RCS opera ti ona l leakage exceeded 1 gpm [gallon per minute] unidentified l eakage as id entified by performing RCS Water Inventory Bal a n ce using the Nucle ar Plant In formation System Computer.

This required the entry into Technical Specification (TS) Limiting Condition of Operation (LCO) 3.4.13 Condition Bat 0808 [CD T] on 9/2/16. The associated action is to place the unit into Mode 3 in 6 hou r s. Trending of containment s ump l eve l indicates the l eakage is in side contai nm ent with the exact l ocation within containment unknown. Containment inspe ct i on i s being performed to try and iden tify the source of Reactor Coolant System leakage. NRC Resident I nspector has been notifi ed. Realignment of the Letdown System back to its norma l arrangement has subsequently reduced RCS l eak rate to 0.521 gpm at 0652 CDT on 9/2/16. Unusual or N ot Understood

-Leak Lo cation is not known at thi s time. Maximum l eak rate r ecorded was 1.358 gpm. Th e l eak was first discovered at 08/31/16 at 1519 CDT. Safety Related Equip ment not operational

-Reactor Vesse l Level I ndicating 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 l ooking down on the vessel head the l i censee i de n ti fi ed s i g n s of a boric acid l eak over a mirrored in su l at i on pane l. After removi n g the panel and using a camera t he licensee saw a p l ume in the area of severa l penetrations. The licensee was able to determine that th e leak was on a core exit th ermocouple nozzle thre aded connection.

Th e licensee a l so determ i ned that this was n o t pressure boundary l eakage. I n addition , the licensee identified th at excess letdown made the leak rate see m worse than th e actual va l ue. The leak rate was eventually quantified at around 0.6 gpm. Without being p r essure boundary leakage and since the leak rate was l ess than 1 gpm , the l i censee was ab le to exi t the LCO. The li censee has decided to go into their planned refueling outage and wil l perform some pre-o uta ge s urveill ances before cool in g down t o MODE 5. The l eak will be repaired dur i ng th e refueling outage while the head is on the stand.

From: Sent: To: C c: Subj e ct: Tha n ks, Jeff F r om: Alley, David Poehler, Jeff r ey 8 Sep 2016 13:09:42 -0400 Alley, David;Butcavage, Alexander Coll i ns, Jay RE: HAVE YOU H EARD ANY DETAILS On TH I S J EFF??? S e nt: Thursday, September 08, 2016 1:05 PM To: Poehler, Jeffrey; Butcavage, Alexander Cc: Collins, Jay Subj e ct: RE: HAVE YOU HEARD ANY DETAILS On THIS JEFF??? I a m pre tt y sure I started on a r esponse to t h is yeste r day and d idna*Žt get it se n t. L eak i s in a canopy sea l weld. Pressure boundary is a th r eaded j o i nt. Weld is j us t for leak tightness.

Weld is stainless to stainless us i ng stainless fil l er. Wolf Cree k was going to go i nto an outage in a coup l e weeks. They j ust a r e sta rti ng the ou t age ear l y. I n itial thoug h ts a r e to u se a mechan i cal c l amp but region IV wi l l be looking into that approach. They appa r ently have something l ike 10 mechanical c l amps a lr eady i nstalled. Dave From: Poehler, Jeffrey S e nt: Wednesday, September 07, 2016 2:09 PM To: Butcavage, Alexander

<A l exander.Butc a vage@nrc.gov> Cc: Collins, Jay <J ay.Collin s@nrc.gov>; Alley, David <Dav i d.Alley@nr c.gov> S ubje c t: RE: HAVE YOU HEARD ANY DETAI L S On THIS JEFF??? No , I had not. Copied folks from the b r anch t hat handles upper head penetrations.

Jeff F r om: Butcavage , A l ex a nder Sent: Wednesday, September 07, 20 1 6 1: 09 PM To: Poehler, Jeffrey <Jeffrey.Poehler@nrc

.go v> Subj e ct: H AVE YOU HEARD ANY DETAILS On TH I S JEFF??? NRC S a y s Wolf Cre e k L e ak C a u s ed By Reactor V es sel H e ad Penetration Noz z le I s su e. WTBW-TV T opeka , KS (9/6 , P a l mer , 78K) reported on its website that NRC Public Affairs Offic e r Victor Dricks a*ceto l d 13 NEWS the leak was not caused by a ba d we l d as some report s have indic a ted. Dricks s aid l eak wa s in a p e netrat i o n no zz le at the top of th e r ea ctor ve ss e l.a*Ža*D But th e is s ue wa s not re l ated t o the weekend ea11 h quake affecting s even s tate s , nor did any radiat i on escape the facility from the water leak. a*ce0n a r elated to p ic, Wo l f Creek officia l s ar e m ee ting with Nuclear Regulatory official s in Arlington, Texa s On S e pt. 2 1 to d i s cu s s w h at th e y ca ll ed , a e ... an appare n t violat i o n in m a i ntaining e merge n cy die s el generator s at the p l ant.a*Ža*D Offic i a l s a*cesa i d that a g enerato r fa il ed 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> into a 24-hour run due to a fau l ty e l ectr i cal component , however , t h ere was no danger , because oth e r mean s were available to s upp l y e mergency power to the plant if n ee ded.a*Ll Th e T e x as me e tin g w i ll dete r mine whet h er add i tiona l inspect i on s and over s i g ht are needed.

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

Subject:

Att a chment s: FW: Wolf Creek CSC.""" A'""a"'---------------.

DCP 012962.docx I

WCNOC Change Pkg. .0 1 2962 withh e l d in fu ll under ex4. Fy i lea k ing at the CROM ca n opy seal we l d at wo l f creek. They used a mechanica l c l amp and 50.59 eva l uat i on. Reg i on I V has questions on t he l icenseea*Žs 50.59 eva lu at i on From: Drake, James S e nt: Monday , September 12 , 2016 1:27 PM To: Lyon, Fred ; Al l ey, David ; Werner, Greg C c: Anchondo, Isa ac; Kopriva, Ron ; Tsao, J ohn ; Hoffman, Ke i th ; Taylor, Nick; Dod so n, Dougla s; Thomas, Fabian

Subject:

RE: Wo l f Cree k CSCA'a Attachment would help. Jim From: Drake, James S e nt: Monday, September 12, 2016 12:13 PM To: Lyon, F red <Fred.Lyon@nrc.gov>; Alley, D a vid <David.Alley@nrc.gov

>; Wern e r , Greg <Greg.Werner@nrc

.gov> C c: Anchondo, Isaac <lsaac.Anchondo@nrc.gov

>; Kopriva, Ron <Ron.Kopriva@nrc

.gov>; Tsao, J ohn <J ohn.T s ao@nrc.gov>; Hoffman , Keith <K e ith.Hoffman@nr c.gov>; Ta ylo r, Ni c k <Nick.T a ylor@nrc.gov>; Dodson, Doug l as <Douglas.Dodson@nrc

.gov>; Thomas, Fa bi an <Fabian.Thomas@nrc.gov

>

Subject:

Wo l f Creek CSCA'a I h i ghlighted so i nte r es t ing statements.

L oo k s l i ke the West i nghouse ana l ys i s was only for stresses o n t h e nozz l e. T he lice n see appea r s to be t rea ti ng t h e en tir e h ead adapte r as t h e nozzle. The firs t comment in the document w ill ta k e you t o the 50.59 screen i ng t h e l icensee performed. Jim .James f:. :Drafi.e J ames F. Drake Off i ce pho n e: 8 1 7-200-1558 Ce ll Phone: l (b)(B) I From: Tsao, J ohn Sent: 12 Sep 2016 14:40:31 -0 400 To: Collins, Jay Subje ct: FW: Wolf Creek CSCA'a Att a chment s: WCNOC 30-day response to BL 2002-01.pdf, WCNOC 60-day response to BL 2002-01.pdf , NRC closeout of BL 2002-0 1 to RVH In s p ec tion Ord er.pdf, NRC re s pon se to WCNOC BL 2002-01.pdf See l i sting of Reco r ds A l ready Ava i lable f yi From: Lyon, Fred Sent: Monday, September 12, 2016 2: 38 PM T o: Drake, James C c: Alley, David; Werner, Greg; T sao, John; Hoffman, Keith; Kopriva, Ron; Anchondo, I saac Subj e ct: RE: Wo l f Cree k CSCA'a Other relevant documents attached.

I havena*Žt run down the Order trail yet. From: Lyon , Fred S e nt: Monday, September 12, 2016 1: 40 PM To: Drake, James <James.Drake@nrc

.gov> C c: Alley, David <David.All e y@nr c.gov>; Werner, Greg <Greg.Werner@nrc.gov

>; Tsao, John <J ohn.T s ao@nrc.gov

>; Hoffman , Keith <Keith.Hoffman@nrc.gov

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

>; A n chondo, I saac <l s aac.Anchondo@nrc

.gov> Subje c t: RE: Wo l f Creek CSCA'a Only found one CROM canopy seal weld repair u sing a mechanical clamp -Salem in 1988. The Turkey Point one is conta i nment boundary l eakage repaired with a mechanica l c l amp (though the title says pressure boundary).

Saw plenty use of MNSAs. I included the Robinson and Seabrook examples simply because the licensee o r ig i na ll y considered mechanical clamps, but then did weld overlays. From: Dr a ke, James S e 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@nr c.gov> Cc: Anchondo , Isaac <lsaac.Anchondo@nrc.gov

>; Kopriva, Ron <Ron.Ko p riva@nrc.gov>; Tsao , J ohn <John.Tsao@nrc.gov

>; H offman, Keith <Keith.Hoffman@nrc.gov

>; Taylor, Nick <N i ck.Taylor@nrc.gov

>; Dodson, Doug l as <Doug l a s.Dod s on@nrc.gov>; Thomas, Fabian <Fabian.Thoma s@nrc.gov> Subj e ct: RE: Wo l f Cree k CSCA'a Attachment would help. Jim From: Drake, James S e nt: Monday , September 12, 2016 12:13 PM T o: Lyon , Fred <Fr e d.Lyon@nrc.gov

>; Alley, David <Dav i d.Alley@nrc.gov

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

> Cc: Anchondo , Isaac <l saac.Ancho n do@n r c.gov>; Kopriva , Ron <Ron.Kopriva@nrc.gov

>; Tsao, J ohn <John.Tsao@nrc.gov

>; H offman , Keith <Keith.Hoffman@nrc

.gov>; Taylor, Nick <Nick.Taylo r@n r c.gov>; Dod s on, Doug l as <Doug l a s.Dod s on@nrc.gov>; Thom as , Fabian <F a bi a n.Thoma s@n r c.gov> Subje c t: Wo l f Creek CSCA'a I hi ghl i ghted so i n t eres ti ng sta t eme n ts. L oo k s l i ke the West i nghouse ana l ys i s was only for stresses on the nozz l e. T he licensee appea r s to be t rea ti ng t he en tir e head adapte r as the nozz l e.

The first comment in the do cument will take you to the 50.59 screening the licensee performed.

Jim $r m es 'f. :15raf e J ames F. Drake Office phone: 817-200-1558 Ce ll Phon ej (b)(6) I From: Sent: To: S ubj e ct: Attachm e nt s: Tsao, Joh n 12 Sep 2016 13:39: 57 -0400 Collins, Jay FW: N-733 ML13263A372.pdf, ML073240650.pdf See listing of Records A lr eady Ava il able to the Public for these attachments.

Background info on Wolf Creek. Pis see ema i l below F ro m: Drake , Jame s Se n t: Monday , September 12, 20 1 6 9:09 AM To: All e y , Da v id Cc: T s ao , John ; Hoffman , Keith S ubject: FW: N-733 Dave, As soon as I get some additional information I will be giving you a call to discuss. Jim F r o m: Ancho n do , I s aac Se n t: Thursday , September 08 , 2016 3: 31 PM T o: Drake, J ames <Ja m e s.Dra k e@nrc.gov> Cc: Wern e r , Gr e g <G r e g.Wern e r@nrc.gov>; Kopr i va , Ro n <Ron.Kopriva

@nrc.gov>; Tay l or , Nick <N i ck.Tay l o r@nrc.gov>; T h oma s, Fab i a n <F abian.Thoma s@nrc.gov>; Dodson, Do u g l as <Douglas.Dodson@nrc.gov> S u b j ect: RE: N-73 3 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 othe r than B-P, "P r essure 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 mechanica l joints (2001 Edition), IWA-4321, "Class 1 Mechanical Joints" (c) Threaded joints in which the threads provide the onjy sea l shall not be used in Class 1 piping systems. I f a sea l we ld is em loyed as the sea ling mediu m , t he s t ress analysis of t he joint s hall incl u de t h e st r esses i n t h e weld res ul t in g from t he relat i ve deflect i o n s of t h e ma t ed 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. h e m echanical connection assembly and the vessel or pi ping l oca t ion where t he mechanical connec t ion assembly is installed s h all be designed tak i n g no s t ructu r al credit f or th e exist i n g Category D or branch connec t ion partial p ene t r at i on w e l d 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 o m: Drake , Jame s Se nt: T h ursday , Sep t embe r 08, 20 1 6 I : 54 PM T o: Anchondo, I s aac <lsaac.Anc h ondo@nrc.gov> Cc: Wern e r , Gre g <G r eg.Werner@nrc.gov>; Kopr i va , Ron <Ron.Kopr i va@nrc.gov>; Tay l or , Nick <Nick.T ay l or@nrc.gov>; Thomas , Fabia n <Fab i an.T h omas@nrc.gov>; D od s o n , Do u g l a s <D o u g l as. D o d so n@nr c.gov> S ubj ect: 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 struc tural 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 s h all n o t be used on the following: (1) C lass 1 p ip in g; (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 sha l l 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 i n accordance with the requirements of IWA-4400.

We l ding shall be documented on an N I S-2 Form. (f) The records required by IWA-6000 shall be maintained by the Owner un t il 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 l eaks of this magnitude.

Jim James F. Drake Office phone: 817-200*-1558 Cell .... (b-)(6_) ___ _.

Exelon Generation RS-13-234 September 19, 2013 U. S. Nuclear Regulatory Commission A TIN: 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 50-457 630 c,-7 2000 011oce 10 CFR 50.55a

Subject:

Relief Request 13R-11 Associated with Alternative Requirements for Repair/Replacement of Control Rod Dr ive 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 197 4 Addenda. During boroscopic inspection of the reactor head assembly dur i ng 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 locat i ons. The locations of the leakage is suspected to be the omega seal welded threaded connection on one or more of these CROM penetrations.

IW A-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) 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 si gnificantly reduces the projected occupational radiation dose when compared to the Code required repair method.

Tennessee Va ll ey Autho ri ty, Post Office Box 2000 , Spr i ng Ci t y, Tennessee 37381-2000 0 1 \9 99 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk W ashington , D.C. 20555 Gentlemen: In the Matter of Tennessee Valley Authority Docket No. 50-390 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 fo r TVA. SIA h as indi cate d that the f r actur e mechanics analysis report SIR-97-089, " Design and Analysis of a W e ld Overlay Repai r for the watts Ba r CROM Lower canopy Seal welds ," Revision 0, previously prepar ed for t he canopy seal repair du ring WBN Unit 1 Cycle 1 refueling outage, is directly applicable to the canopy seal weld for the GS penetrati on. The geometry and materia l s are identical to those previously repaired canopy seal welds. Based o n the comparison, the overlay design for the canopy seal weld on the GS penetration is identical to the welds discussed in the Report SIR-97-\ I 08 9 , Revis ion 0.

in that report can be used (\, 1 9904150261 990407 \t"" PDR ADOCK 05000390 0 PDR U.S. Pp..,.q_e Nuclear Regulato r y 2 0 7 1999 Corrunission 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 co n trol person n el using a remote video camera with approximately BX magnification.

The e x amination was also verified by a T VA L e 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 me at (423) 365-1824. P. L. Pace Manager, Licensing and Industry Affairs Enclosure cc: See page 3 l U.S. Nuclear Regulatory Commission r>age 3 .APR a* 7 1999 cc (Enclosure)

NRC Resident Inspector W atts Bar Nuclear Plant 1260 Nuclear Plant R oad Spring City , Tennessee 37381 Mr. Robert E. Mar tin , Senior Project Manager U.S. Nuclear Regulatory Commission One Wh ite Flint North 11555 Rockville Pike Rockville, Ma rylan d 20852 U.S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth st., sw , suite 23T85 Atlanta , Georgia 30323 ENCLOSURE WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RE LIEF l-RR-2 , Revision 1 Summary: Unit: System: Component: Code Class: Function: Code ENCL O SURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l-RR-2, Revision 1 During the Watts Bar Nuclear Plant Unit 1 Cycle 2 refueling outage activity of disassemb l 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 le aking at the lower canopy seal weld (See Coordinates on Attachment 3). The ASME Code req uires the defects be removed and the configuratio n of the material be reproduced in order to restore the canopy seal to its original design condition.

Due to the physical space lim itations and in consideration o f radiation exposure , Watts Bar proposes as an alternative to pe rform a w eld buildup over the leaking canopy seal weld (See prop osed design in Attachment

4) rather than removing the defect and perf orming a w eld repair. Also, an enhanced visual examination is proposed as an alternative to the l iquid penetrant examination r equired by the origi nal construction code for the final weld buildup. TVA's proposed alternative sea l weld repair and examination methods have been previously implemented at W atts Bar , TVA's Sequoya h 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 10 CFR50.55a(a) (3) (i). 1 R eactor Coolant -System 68 Control Rod Drive Mechanism 1 Vertically position a co ntr o l rod in the nuclear core by raising or lowering an interconnecting drive shaft. Requirement

ASME Section XI, 1989 Editi on , IWA-4110(a), " This Article provides rules and requirements for repair of pressure retaining components and their supports , including appurtenances, subassemblies, parts of a co mponent, and core support structures, by w elding , brazing , or metal rem oval.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, Repa i r Procedure requirements:

E-1 Basis for Relief: ENCLOSURE 1 W ATTS BAR NUCLEA R 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 orig in al Co n str u ction Code of the component or system.

• Later Editions and Addenda of the construct i on Code or*of Sect i on III, either in their entire t y o r portions thereof, and Code Cases may be u sed. I f repair w elding 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 compo nen ts." b. Paragraph I WA-4130(a)

(2), "Repair operations shall be performed in accordance w ith a program deli n eating essential requirements of the complete repair cycle inducting

... ( 2) ... below: ( 2) the flaw removal m ethod , method of measurement of the cavity created by removing the fla w , 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 W eld" requirements

d. Paragraph NB-5271 , "W elds of this type (weld s of specially designed seals , i.e., canopy seal w elds) shall be examined by either the magnetic particle or liquid penetrant m e thod." 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) . Fur ther inspection shows that one CROM has started leaking at the l o w er canopy seal weld. See Attachments 1 , 2 and 3 for configuration and location of the CRD M 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 th e Final Safety Analysis Report (FSAR). The 1971 Edition, Addenda through Winte r 1972 of A SME Section III establis h the design spec ifi cat ion and the constructi o n code for the CRDMs. The 1 971 Edition , Adde nd a through *winter 1971 of ASME Section III establish the design specificat i on and the construction code for the reactor vessel. E-2 Alternative Repair ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEME N T REQUEST FOR R E LIEF l-RR-2, Revision 1 The CRDMs are fabricated in sections w i th th r eaded jo i nts providing the pressure-retaining capabilities.

Since the threaded joint provides pr e ssure retention , the canopy seal weld is not pressure reta ining and is for leakage control. The 1971 Editi on , Addenda through Winter 1972 of ASME Section III does n ot 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 ca n opy seal welds and qualification requirements for welding specially designed welded seals , respective ly. 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 lo w as reasonably

• achievable (750 -800 milli rem per hour on contact and 100 -150 millir em per h o u r general area), removal a nd repair of the defect is n ot the most favorable meth od of repair . . In addition , if the defect was rem oved , it would be : impossible. to reproduce the configuration of the canopy seal to its original design condition as required by IWA-400 0. Requirements

WBN w ill apply the following alternative weld overlay rep air requirements:

a. A w eld over.lay re pair designed under the requi re m ents of ASME Section XI, 1989 Edition , Paragraph IWB-3640, "Ev aluation Procedures and A cceptance Criteria for Austenitic Piping ," and Appendix c, "Evaluation of Flaws in Austen i tic Piping," will be used as an alternative repair method. Guidance will als o be taken from ASME Section XI Code Case N-504, "Alte rnative Rules for Repair of Class 1 , 2, and 3 Austenitic Stainless Steel Piping," and NUREG-0313, "Techn ical Report on Material Selection and Pr ocess ing 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 i. Justi f icat i on For The Granting Of Rel ie f: ENCLOSURE 1 WATTS BAR NUCLEAR PLANT UNIT 1 REPAIR AND REPLACEMENT REQUEST FOR RELIEF l-RR-2, Revision 1 T VA's Code of Record for Repairs and Replacements is ASME Section XI , 1989 Edition. IWB-3640 and Append ix C of the 1989 Edition of ASME Section XI wil l be u sed t o perfor m t he requ i red fracture mechanics a n d 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 witho u t 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 Acceptab il ity ASME Section XI Division," Revision 11. IWB-3640 provides criteria for acceptance of flaws without : repair in ductile , austenitic materials. The basis for ; such 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 "Technica l Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping , Final Report ," Rev isio n 2 , is used for guidance.

The use of NUREG-0313 wil l 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 co ntinued 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 110 kips per square inch (ksi). The Inconel weld ma te rial 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 co rrosion cracking, and is highly ductile. The load carrying *capability of the repaired l oca tion will be greater th an : the original component. : Liquid penetrant examinations that are required by NB-527 1 : wi l l not be performed because of space limitations, which prevent examine rs the needed access to successfull y perf o rm the examination and in consideratio n of maintaining worker dose as low as reasonably achievab l e. As an alternative, TVA will use a remote video camera w ith a magnif i cation of approximately BX and perform a visual examination of the final weld at the enhanced magnifica tion. The basis for this approach is that weld liquid penetrant exam i nations are surf ace examinations , and provides minimal assurance of repair i n t egri t y when compared to an enhanced visu al e x amina t io n. E-4

==

Conclusion:==

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 mec han ics analysis , coupled with the enhanced visual exam ination, 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 w elds at SQN and the results documented in a letter to the NRC dated Ap r il 3 , 1996. The demonstration was performed using a mach i nist scale to determine if a 1/32 of an inch graduation could be distinguished and was found acceptable. The proposed alternative w eld overlay repair and visual examination req u irements 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-4 130(a) (3); "In spection" per Subartic l e IWA-4140; "M aterial" per Subarticl e IWA-4200; "Welding and Welde r Qualific at i o ns" ,per Subarticle IWA-4400; and "R ecords" per Subarticle : 4800. The design of the weld overlay repair and the *safety evaluation per 1 0 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-4 130(a) (1) and (4) is considered within the DCN. Based on the above d i scussion , the alternat i ve we ld overlay repair and visual examination provide an acceptable level of quality and safet y. Authorization to implement the proposed alternatives is requested in accordance with 10CFR50.55a(a)

(3) (i). E-5 ENCLOSURE 1 ATTACHMENT 1 FUL L LENGTH CONTROL ROD DRI VE MECHANISM . "' mm '""-----. . , . I '--*--OPEUTIH COIL STAtl USEMll.Y IYl IOD USEMtl l 0 I sc:o**EtT 10 ElAl-1 CAIL£ to*ECTIOI location of lower canopy seal weld ENCLOSURE 1 ATTACHMENT 2 TYPICAL LOWER CANOPY SEAL WELD DETAIL : ElA2-1 A p N M L I( J H-<f. G F' E 0 c B A ENCLOSURE 1 ATTACHMENT 3 LOCATION OF CRDMs TO BE REPAIRED t5161Ji211I098 76S4 321 CRDMs with Lower Canopy Seal Weld Overlay Repairs CRDM # I Location 13 l G5 ElA3-l 1e o*

ENCLOSURE 1 ATTACHMENT 4 PROPOSED WELD DESIGN E1A4-l From: Sent: To:

Subject:

Tsao, J ohn 13 Sep 2016 12:36:14 +0000 Drake , J ames Accepted:

Wo l f Creek canopy seal leak r epa ir From: Sent: To:

Subject:

Dave, Tsao, John 13 Sep 2016 22:32:31 +0000 A ll ey, David;Hoffman, Keith;Kalikian, Roger another thing about the wolf creek clamps At the end of phone call wit h 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 sea l we ld s. 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 c lam ps and how they are designed at Wolf Creek, a mechan i cal c l amp provides a stronger support than a seal weld. Simply put, a clamp has more metal than a sea l weld. The clamps are simi l ar to the strongbacks that can support lots of l oads. A sea l weld probably has only two weld passes. Second ly , the existing 10 clamps are installed on existing sea l welds, a lb eit 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/nozz l e i s supported at the RPV head. I f the top of the CROM housing is supported also then i have no concern on the impact of the weight of the mechanica l 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 Sea l Clamp Assemblies Attachments:

2008 10_15_08 Material Engineer in g Counter p art Call Summary.doc, 2008 10_15_08 Material Engine ering 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 l icensees that are using or have used them in the past. We have found that the i ss ue came up back in 2008 and just as we did the NRC Staff struggled with whether the u se of t he CSCA was acceptable. The issue was discussed at a M ate ri als Engineering Counterparts Ca ll on 10/15/2008.

Th e attached PDF file shows a document that descri bes 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 Te d Sullivan.

The decision they re ached in 2008 was that using Appendix J of Section XI the use of the CSCA was a ma intenance activity that did not require a Repair/Replacement Plan. Keith M. Hoffman Materials Engineer NRR/D E/EPNB {301)415-1294

SUMMARY

OF THE OCTOBER 15th 2008 MATERIALS ENGINEERING COUNTERPART CONFERENCE CA LL On October 15th , 2008, the staff of the Divi sion of Component I ntegrity of the Office of Nu c l ear 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) reg a rding ongo ing materia l issues and related activ i ties at nuclear p l ants. In support of the ca ll , the conference participant l eads provided summary inputs , see below: CRDM C l amp Repair Method (CSCA) [George Hopper] R2 discussed the need to document the use of the CRDM clamps as an acceptable repair method fo r use on ca nopy seal weld le aks/f l aws. This device has been used by the industry for a while and we had to reinvent the wheel regarding its use at H arris. A legacy document wou l d be helpful fo r future encounters. Ted Su ll ivan will l ook into this. [Ted Sullivan]

Canopy Seal Clamp Assemblies (CSCA) In August a leak was found in a co ntrol rod drive me cha ni sm (CRDM) canopy sea l we l l d at Shearon H arris. Harris personnel decided to address the leak by install i ng a mechanical canopy seal assembly. During the October 2008 Counte r parts phone call , Region II asked NRR staff to document its position with respect to installation of CSCAs on CRDMs. Th e write-up below is i n response to th at req uest. Although this is not a new method for addressing leaks through CRDM canopy sea l we ld s , the NRC staff addressing this iss ue did not have prior experience with th is application and ra ised questions in th e areas of whether this application was a non-code repair that needed relief f rom ASME code requirements and whether the licensee wou l d h ave to do anything about the ac t ual Class 1 l eaking boundary (i.e., the threaded connec tion). T he following represents a summary of the basis provided t o the staff regard ing the questions raised.

• A threaded connection is a pressure retain in g componen t , but because no st ru ctura l credit was taken for the seal welds in the design of the CRDMs, the CRDM seal we l ds were not designed to be a Code co mpon e nt.

• The decision process presented in Non-Mandatory Appendix J of Section XI , "Guide to Plant Maintenance Activities and Section XI Repair/Replacement Activities," leads t o the conclusion that insta lli ng the clamp assemb l y i s not an activ ity with i n the scope of the repair/replacement requ ire ments of Section XI , I WA -4000. The installat io n of a CSCA does not affect th e p r ess ur e retaining portion of a code item and the activity comes under maintenance. This activity does not affec t tests or examinations or reco r ds of tests or examinations that would be required under Section XI.

• The Engineering Change process requires the licensee to ensure that under this activity th ey co ntinu e to meet their desig n basis, which i s th e ASME Code , Section Ill. H a rri s person n e l ensured that they met the requirements of Sec ti on Ill , N B-3671.3 for threaded joints.

• AN ll involvement is not required , al though H ar ri s provided the AN ll with a courtesy review. No N I S-2 form was prepared.

• Harris conducted a VT-2 leakage examination prior to re turnin g the p lant to service, in accordance with the normal r equi rement s of IWA-5000. The staff concluded that this basis for application of the CSCA a t Harris was reasonable.

In addition, the staff searched on this topic in ADAMS. While a number of documents discussing CSCAs were f ound, no reli e f requests on thi s topic were id en t i fied.

Attachment:

West i nghouse Nuclear Services/Engineering Services Canopy Seal C l amp Assembly (CSCA TM) Brochure Catawba Repair I ssue R2 continues to fo ll ow-up on t h e Catawba service water pipe repair issue. T he li censee used a wooden plug t o stop the leak when th ey did the r epair by modification.

Th ey welded a pipe with cap over the hole and le ft the plug in. There fo re, no pressure test was performed, s i nce there was no way to verify the plug dislodged or a ll owed water to pa ss. Th e ANll refused to sign off on the repair. T ime 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 i s actually occurringcr at the s it es. F a rl ey took ove r 10 days to eva luat e th e structu r a l integrity of a leaking service water pipe a nd 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 co m mu nic ate th e expec t ation via another document.

DC Cook Tu rb ine/Fire Supp r ess ion System F a ilur e On September 20, 2008 at DC Cook Unit 1, exper i enced a turbine f ailure and generator hydrogen fire. Many blades i n the low pressure side of the turbine were damaged or failed. Addit io nally , the fire protection header outside the turbine building ruptured short ly after the turbine failure causing l oss of the fir e protection system and severely damaged two of th e 3 fire pumps due to continued pump operation without water in the sys tem. Region Ill is conduc ting a special inspection to investigate this event. Perry Re-Review of UT Data At Perry, dur in g the li censee re-review of UT data on the N 6A and N6C RHR noz z l e-to-SE dissimilar metal welds, the licensee identified Code rejectable flaws. Licensee is completing ASM E Code Section XI I W B-3 600 fl aw eva luation to accept welds for continued service (Plant was at power dur ing this review and remained a t power). Perry High Cycle Fatigue Failure At P erry, li censee expe rien ced a high cycle f atigue failure causing a thru-wall l ea k on a 6" reacto r water cleanup system line just downstream of the re je native HX. Licensee mainta i ned this non-safety related system in operation for more than 5 days without providing a basis for functiona lity. No NRC policy in this area. FF S Fitne ss-For-Servic e API 579-1/ASME FFS-1, JUNE 5, 2007 The subject recommended pract i ce i s a compendium of analys i s techniques that can be applied to inform run/repair/replace decisions for pressur i zed components.

T he document was adapted from an earlier API petrochemical industry document and it borrows libe r ally from a wide variety of pressure vessel codes and standards.

I t is current l y published unde r a committee that includes ASME sponsorship, so it is an ASME recommended practice.

I t may be accessed through the IHS Codes and Sta n dards l ink on the N RC website under the in format i o n resou r ces page. I n orde r to find it in I HS, you would search fo r the term " FFS-1." The document addresses p r ob l ems of a genera l nature and warns users that particular circumstances require review of app l icable codes, standards and regulations. The applicability section states: "T he 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 fair l y detailed approaches for assess i ng susceptibi l ity to, or damage caused by, common degradation mechanisms.

I t i s arranged by degradat i on mechanism, so there is a chapter on add r essing br i tt l e fracture , one on pitting, and chapters on creep damage , fire damage, general wastage , we l d m is alignment and shell distortions, laminations , crack-like flaw s, gouges, etc. Each chapter includes some standard assessment techniques , describes acceptab l e general approaches and includes acceptance criteria.

Overall, the document provides a wide variety of useful, standard engineering tools and approaches for making run/repa i r/replace assessments.

We have not reviewed this document.

I t is not clear that licensees cou l d use it to assess systems, structu r es or components that a r e designed to Codes or Standards t h at a r e not i ncluded in the applicability section. It is possib l e that a licensee could use the guidance in FFS-1 to assess safety related s ystem components, piping s ystems constructed to B31.1 , or for performing functiona l ity (not operability) assessments of degraded ASME components.

The example we discussed during the Counterparts Call was the Comanche Peak hot l eg boric acid corrosion assessment.

The l icensee could have e l ected to use the gu i dance in t he "Assessment of General Meta l Loss" chapter to determine that the reactor coolant system remained func t iona l (or not) before they completed the actions required to make their operability determination. T hey are not constra i ned to use the ASME Code for a f unctionality dete r mination.

We have not yet collectively d i scussed or decided how this document wou l d fit ointo our regu l atory framework.

Nuclear Services/Engineering Services Canopy Seal Clamp Assembly ( CSCA Ž) Background Small leaks in PWR head penetrations can prevent a return to power and cause expensive delays unti l a fix is devised. An increasing number of plants are reporting primary coolant leaks in the field-welded canopy seal area. Westinghouse offers a full range of products and services to control these kinds of leaks, including a unique mechanical clamp assembly, the CSCA. (.)Westinghouse Description The ca n opy seal is a weld between the reactor vesse l head control rod drive mechanism (C ROM) penetration and th e mating part. This weld ha s a tendency to develop cracks as a result of stress corrosion cracking (SCC) and/or original weld defects. T h ese cracks spread through the walls until there's l eakage. The We st in g h ouse CSCA provides a non-welded mechanical method of stopping leaks in the canopy seal weld. The CSCA seals the leaking weld and introduces a comp re ssive l oad into the canopy seal, which tends to close and arrest the c rack propagation.

The CSCA seals the leaking weld by compressing a graphite seal over the e ntire canopy seal weld area. The CSCA consists of a hous i ng, a top plate, sea l carrier halv es, a split graph it e seal, cap screws, and Belleville washers. The general arrangement is the same for spare capped CROM penetrations, full-l e ngth active CRD M penetrations, and for core exit therrnocoup[e (CET) CROM penetrations.

'fo be repaired, the hou s ing is l owered over the penetration, b elow th e eleva tion of the canopy seal weld. The first seal carrier half with graphite is l owered and placed into the housLng; the second sea l carr i er ha l f i s in sta ll ed the same way. The housing is then raised until the graphite seal contacts the ca nop y seal and the top plate, and the cap screws with th e Belleville washers are installed to provide the necessary l oad in g forces. Trained Westinghouse crews will implement the in s tallati o n s with spec ial tooling designed for thi s uniqu e application. W e provide full e ngin ee ring and support services, including design specificatio n, stress report, and ce rtifi ca tion s to Sec tion III of th e ASME Co d e. At severa l plants, the installations were done on an e m e r gency basis. Usually, in this s itu atio n, t h e CSCAs are install e d and th e l eak i s el iminat e d within four to five days. after notification by the utilit y. Benefits The average radiation dose for all installations was Jess than 300 millir e m per CSCA, compari n g favo rably w i th abou t l.5 man-rem per penetration for weld overlay r e p ai r s, and up to 10 to 20 man-r e m per penetration for o th er "c ut-an d-cap" or welding opt i o ns. CSC A installation can be done without dra i ndown of th e r e a cto r coo l ing system (RCS)-a di s tin c t advantage over weld repair opt i ons. A l e ak at o n e plant wasn't discovered until th e ou ta ge was over and th e unit was ret urned to power; th e CSCA w as in sta ll ed w i t h the RCS in M ode 5 at abo ut 300 psi a nd at l 70°F. At another plant, th e CSCA was in sta ll ed during a system h eat-up to about 200°F.

• The CSCA works o n l eaking or n on-leak ing ca nopy seal welds, and can a l so b e modified for installation o n previously ove rl ay-repaired welds. Applying the CSCA on a n o n-l eaki n g sea l weld is a good preventative m easure agains t cont in ued sec. w hi c h i s a primary ca u se of failure.

From: Sent: To: Cc: Taylor, N i ck 15 Sep 2016 14:48:20 -0500 Alley, David;Anchondo, I saac; Drake, J ames;Werner, Greg 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 ca ll ed to ask about the "ASME code non-com pliance" and "improper repairs to the vessel head" at Wolf Creek (as relayed to them by our reg ional management after t he morning meeting here yesterday I think). I let them know the following:

That there i s n o cu rre nt safety iss ue given that the plant is shut down and won't see power operation for at least 2 months , and t h at we will have ample opport unit y to inspect the head and the licensee's actions before restart Regarding the code issue -I let them know that the staff i s sti ll h aving dialogue abo ut what the code require s for the se penetration leaks , and that we are st ill waiting for the licensee to answer some questions before we determine whether or not a compliance issue exis t s Please let me know i f I've communicated anything in error. I look forward to more discussion on this as the facts become m ore clear. DRP m anagement has asked that we provide status periodically as the issue develop. Thank s, Nick Taylor Chief , Pr ojects Branch B Divisi on of Reactor Project s USNRC Region IV 0: (8!7) 20 0-1141 C: l (b)( I E: n i ck.taylor@nrc

.gov I R From: Alley, David Sent: T hursd ay, September 15, 2016 2:12 PM To: Anchondo, Isaac <lsaac.A n chondo@nrc.gov>;

Drake, Jam es <James.Drake@nrc.go v>; Tayl or, Nick <Nick.Taylor@nrc

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

> Cc: Graves , Samuel <Samue l.Grav es@n rc.gov>

Subject:

RE: FW: NR C Questions regarding P enetration 77 No apology necessary.

Your focus is exactly in the right spot which is the safety significant spot. Dave Fr o m: Anchondo , I saac Sent: Thursday, September 15, 2016 3:07 PM To: Alley, David <David.Alley@nrc

.gov>; Drake, James <James.Drake@nrc.gov

>; Tay l or, Nick <Nick.Taylor@nrc

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

> C c: Graves, Samuel <Samuel.Graves@nrc.gov

> Subj e ct: RE: FW: NRC Questions regarding Penetration 77 Dave , I apologize for forgetting to acknow l edge that the question was already out the r e. 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 Se nt: Thursday, September 15, 2016 1:52 PM To: Anchondo , Is aac <lsaac.A n chondo@nrc

.gov>; Drake, J ames <James.Drake@nrc.gov>; T aylor , Nick <Nick.Taylor@nrc

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

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

> Subj e ct: RE: FW: NRC Questio n s regarding Penetration 77 All, Keith Hoffman is working dil i gently 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 thei r basis , given the amount of leakage, fo r 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 Fr o m: Anchondo, I saac Se nt: Thur sday, September 15, 2016 2:39 PM T o: Drake, James <J ames.D r ake@nrc.gov

>; Taylor, N i ck <Nick.Taylor@nrc

.gov>; All e y , David <Dav i d.Alley@nrc

.gov>; Wern er , Greg <Greg.Werner@nrc

.gov> C c: Gra v es, Samue l <Samu e l.Graves@nrc.gov

>

Subject:

RE: FW: NRC Questio n s 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 <N i ck.Tay l or@nr c.gov>; Anchondo, I saa c <l s a a c.Anchondo@n r c.gov>; Alley , David <Dav i d.Alley@nrc.gov>; Werner, Greg <G r eg.Werner@nrc

.gov>

Subject:

RE: FW: N RC Questio n s 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 mee t ing 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 l uate and come to a regu l atory position.

The CSCA's do not appear to be a safety issue, they are designed to Class 1 standards, they have the required stre n gth, and we are not aware of any prob l ems with l eakage from the clamps. Westinghouse completed the stress analysis and there is no prob l em. However, we have not verif i ed the r esults. Where we are currently at is: I s the use of Canopy Sea l C l amp Assemblies allowable by Code and has Wolf Creek comp l ied with a ll regulatory requirements when they insta ll ed them. Unti l we have Wolf Creek's position on the CSCA and all associated docume n ts, we wi ll be making assumptions and won't be able t o make an informed dec i sion. J im From: Taylor, N ic k S e nt: Thursday, September 15, 201611:17 AM To: Anchondo, Isaac <lsaac.Anchondo@nrc

.gov>; Drake, J ames <J ames.Drake@n r c.gov>; Alley, David <Dav i d.Alley@nrc

.gov>; Werner, Greg <Greg.We rn er@nrc.gov> Subj e ct: RE: FW: NRC Que st ions regarding Penetration 77 All , I wo uld li k e t o see us have a meeting t o get m o r e a l igned on code applicabili t y, e t c pr io r to engag i ng wi th t he l icensee or having a n o t her substantia l disc u ss ion a t t h e morning meeting. Thi s i ssue has come up 3 days i n a row now a t t h e morn i ng meet i ng , a nd t h ere are a ll ot of op i nions o ut the r e on what t he code r e q uires , b ut n ot a l o t of facts from th e l icensee. I'd lik e t o see us a ll on the sa m e page p r ior to comm u n i cating with managemen t or the li censee on w h e t her o r not the licensee i mproperly repa ir ed th e head , etc. T hanks, Ni c k Fr o m: Anchondo, I s a ac S e nt: Thursday, September 15, 20 1 6 10:56 AM To: Drake, James <J ames.D r ake@nrc.gov

>; A ll ey, David <Da vid.Alley@nrc.gov

>; Werner, G r eg <Gr e g.W e rn e r@nr c.gov>; Taylor, Nick <N i c k.T ay l or@nrc.gov>

Cc: Lyon, Fred <F r ed.L yon@n r c.gov>; Hoffman, Keith <Ke i th.Hoffman@n r c.gov>; Tsao, J ohn <J ohn.T sao@n r c.gov>; Dodson, Douglas <Douglas.Dodson@nrc.gov

>; Proulx, David <Dav i d.Proulx@nrc

.gov>; Thomas, Fabian <Fab i an.Thomas@nrc

.gov>; Kopriva, Ron <Ron. Kopriva@n r e.gov> Subj e 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 pressu r e boundary shall be designed, fabricated, erected, and tested so as to have an extreme l y l ow probability of abnormal l eakage, of rapidly propagat i ng fai l ure, 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: (b)(5) "There have been no industry reports of degradation o f canopy sea l welds resu l ting in significant leakage flow rates (Ref. 3). Cons i dering the head adapter flange design, leakage through a crack in the non-press u re boundary seal weld would be expected to be l imited by the load carrying component, the flange connection threads." 'T he Westinghouse hardware failure analysis a l so included examination of some threaded j o in ts that were removed a l ong with the lower canopy seal we l ds. There was no evidence of corros i on or cracking on any of the th r eaded joints that were examined."

I" Any comments are we l come. Thanks, Reactor In specto r U.S. Nuclear Regulatory Commission I Region IV Division of Reactor Safety I Engineering Branch 2 (817) 200-115 2 From: Drake, James Sent: Thursday, Septembe r 15, 2016 7:59 AM To: Alley , David <David.Alley@n r c.gov>; Werner, Greg <Greg.Werner@nrc.gov

>; Tay lor , Nick <Nick.Taylo r@nrc.gov> Cc: Lyon, Fred <F r ed.Lyon@n r c.gov>; Hoffman, Keith <Ke i th.Hoffman@nrc

.gov>; Tsao, J ohn <J oh n.T sao@nrc.gov

>; Anchondo, I saac <l saac.Anchondo@n r c.gov>; Dodson, Douglas <Douglas.Dodson@n r c.gov>; Pro u lx, David <Dav i d.Proulx@nrc.gov

>; Thomas, Fabian <Fab i an.Thoma s@nr c.gov>; K opriva, Ron <Ron.Kop ri va@nrc.gov>

Subject:

FW: FW: NRC Que st ion s regarding Penetrat ion 77 Importance:

High Interesting responses From: M uilenburg Will iam T [ma i lto:wimu i l e@WCNOC.com

] Sent: Thursday, September 15, 2016 7:53 AM To: D r ake, James <James.D r ake@nrc.gov>

Subject:

[External_Sender) F W: NR C Que stions r egarding Penet r ation 77 Import ance: 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 quest i on as well. Bill Questions from 9/12 phone ca l l -Everyone, NRC In spec tor s Jim Drake and I saac Anchondo called this morning with the fo llowin g qu est ions re lated to o u r v esse l head and penetration

77. From Jim Drake

1. What code was used to construct the head, B31.7 or Section Ill? I f Section Ill, what year? ASME Section I ll , 1971 Ed i tion through W i nter 1972 Addenda 2. What Code, for I S i , is Wo l f Creek currently commi t ted to? ASME Sect i on XI 2007 Edit i on through 2008 Add e nda 3. What Class of p i ping i s th i s penetration?

The vessel penet r at i on is ASME C l ass 1. T he CRDM housing is a Class 1 component; there i s no pip i ng involved. The pr e s s ure boundary connection i s a threa d ed mechanical connect i on with a non-pressure boundary s ea l weld outs i de of the threaded connect i on. 4. I s Code Case N-733 applicab l e to this condition?

No, Code Case N-733 is not app l icable because this Code Case i s app l icab l e to vessel penetrat i on welds and the leak i s on the sea l weld of the threaded mech a n i ca l connection. From Isaac Anchondo 1. It is noted that there are 10 other penetrations that have the s e rep a irs made, was Code Case N-733 applied to those efforts? No, see above response. T he canopy seal clamp assembly was a mod i ficat i on des i gned to ASME Section Ill requir ements , so no Code Ca s e wa s needed for the clamp assemb l y. The clamp assembly was de s i gned to ASME Section Il l to assure the stresses in the clamp assemb l y and the RV and CRDM threaded conne c t i ons as a result of apply i ng the clamp assemb l y, did not exceed those a ll owed for Class 1 component s, not be c au s e it wa s sealing a l eak of the nonpressure boundary canopy sea l we l d. 2. I s 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 availab l e so that I can provide a respon s e to the NRC as quick l y as po s sible. Thanks, Bill Muilenburg , ext. 4511 From: Sent: To: Cc:

Subject:

Tsao, John 15 Sep 2016 18:14:48 +0000 Alley, David;Hoffman , Keith;Kalikian, Rog e r Hsu, Kaihwa;li, Yong 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 "mec hanical joint", not as a " mechanical clamp". This is because NB-3671.7 permits the i nstalla t ion of mechanical joints (see Keith's email below). NB-3671.7 S l eeve Coupled and Other Patented Joints. M echanica l joint s , for which no s t andar d s exist, and other pate nt ed jo int s m ay be u s ed provided the r e quir e m en ts o f (a), (b), an d (c) below are met. (a) Pr ovision i s m ade to pre ve nt se paration of the joints und e r all Service Loadings. (b) Th e y arc acc e s s ibl e for maintenance , r emova l , and r e plac e ment afte r serv i c e. (c) Either of the following two crite r ia are m et. (I) A prototyp e joint has b e e n s ubject e d t o performance te s t s to det e rm in e the s afoty of th e jo i n t und e r s imul ated s erv i c e conditions.

Wh e n vibration , fa ti g u e , cyclic conditions , l ow temp eratu re , t herma l expansion, or h ydrau lic s ho ck i s a nti c ip ated , the app li cab l e co nditi ons s hall b e in cor p ora t e d in the te s t s. T h e m e c h a ni ca l joints s hall be s ufficiently l eak ti ght to sa ti s f y th e r eq uir e m ent s of the De s i g n Spec ifi cations. (2) J oin t s are designed in accordance with t he rules o f NB-32 00. A "mechanical clamp" as per ASME Section XI, Appendix IX or Appendix W, is n ot pe r mitted 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

>; T sao, John <Jo hn.T sao@nrc.gov>;

Kalikian, Roger <Roger.K alikian@nrc

.gov> Cc: Hsu, Kaihwa <Kaih wa.H su@nrc.gov>

Li, Yong <Yong.li@nrc.gov>

Subject:

FW: M echanica l Clamp in ASME Se ct ion Ill Based on Ro b ert's view, be lo w, i t appears th at t h e clamp i s acceptable p er the cons t ruction code. Thi s wou l d appear t o m ake t he use o f the cla m p a code repair as it i s i n accordance with t he c on st ru ctio n code. Thi s wo ul d a p pear t o mea n that th e p lant ca n i n s t all a nd l ea v e t h e clamps on forever and that w e have no regulatory h oo k (othe r than, po t ent i a ll y, the condit ion o f t he th r eads for t h is instance based on t he extent of leakage).

Any thoughts?

D ave From: H s u , Kaihwa Sent: Thursday, September 15, 2016 8:25 AM To: Alley, David <Dav i d.Alley@nrc.gov

> Cc: Li, Yong <Yong.Li@nrc

.go v>

Subject:

RE: Mechanical Clamp in ASME Sectio n I ll D ave: I don't see a n y problem fo r the paten t ed clamp to be used over t op of or i ginal joi n t as l ong as the repai r meets AS M E Sec t ion I l l Code cr i ter i a.

Robert From: Li, Yong Sent: Thursday, September 15, 20 1 6 7:29 AM To: H su, K a ihwa <Kaihwa.Hsu@nrc.gov>

Subject:

FW: Mechanical Clamp in ASME Section Ill Please respond to Dave. From: Alley, David Sent: Wednesd a y, September 14, 2016 8: 19 PM To: Hoffman, Keith <Keith.Hoffman@nrc

.gov>; T sao, John <J ohn.Tsao@nrc.gov>; K a l ik i an, Roger <Roger.Kalikian@nrc

.gov>; Li, Yong <Yong.Li@nr c.gov>

Subject:

RE: Mechanical Clam p 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 Il l help. NB-367 1.7 S l eeve C oupl e d and Other P ate nt ed J o in ts 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, Septembe r 14, 20 16 8:08 PM To: Alley, David <Dav i d.Alley@nrc.gov>; Tsao, John <John.T s ao@nrc.gov>; Kalikian , Roger <Roger.Ka l ikian@nrc.gov

>

Subject:

RE: Mechanical Cla m p in ASME Section Ill I believe that is probably a question for Yong Li's branch. Obviously the licensees and ABB -C E/West i nghouse believe i t can be and that is wha t the paten t says it was des i gned to do. From: Alley, David Sent: Wednesday, Septembe r 14, 20 16 4:19 PM To: Hoffman, Keith <Keit h.Hoffman@nrc

.gov>; T sao, J ohn <John.Tsao@nr c.gov>; Ka lik i an, R oge r <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, 20 16 6:39 AM To: Tsao, John <John.Tsao@n r c.gov>; Alley, David <David.Alley@nrc.gov

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

>

Subject:

RE: Mechanical Clamp in ASME Section I ll This was t he section that was referenced as the app l icab l e Section Il l paragraph in one of t he docume nt s I looked at yes t erday. Specifically N B-367 1.7 was refe r enced. N B-36 7 1.3 shows t he t hreaded jo i nt and the requirement fo r the we l d. NB-3670 SP EC IAL PIPING R EQUI R EMENTS N B-3671 Se lec t ion and Limita t ion of Nonwelded P ipin g Join ts The type of piping joint used s hall be s uitabl e for the De s ign Loading s and s hall be s elect e d with con s ideration of joint ti ghtness , mechanica l strength , and t he nature of the fluid handled. Pipin gjoints shall conform to the requirements of this Sub s ection with leak ti g htn ess being a con s ideration in s e l e ction and de s i g n of joint s for piping s yst e m s to s ati s fy the requirements of the Design Specificat i ons. NB-3671.1 F lan ged J oints. Flan g ed joints are p e rmitt e d. NB-3671.2 Expa nd e d Joints. Expanded joints sha ll not be used. NB-3671.3 T hr eaded Joint s. Threaded joint s in which the thread s provide the only s ea l shall not be u s ed. If a seal weld i s employed as the s ea lin g medium , the s tre ss analy s i s of the joint mu s t inc.ludc the s tre ss e s in the weld re s ulting from the relative deflections of the mated parts. NB-367 1.4 F lar e d , F lar e l ess, and Co mpr ess i on Joints. Flared , flarel ess, and compr es sion t ype tubin g fittin gs may b e used for tubing sizes not exceeding 1 in. 0.D. (25 mm) within the limi tations of applicable standa rd s and specifications li s ted in Table N C A-7100-1 and r e quirement s (b) and (c) below. I n the ab s enc e of s uch s tandards or s p e cificat i on s, th e D es i gner sha ll determine that the type of fittin g se l ected is adequa t e and safe for the D es i gn L oadings in accordance w ith the requirements of(a), (b), and (c) be l ow. (a) The pre ss ure de s i g n s hall me e t the r e quirement s of NB-3649. (b) Fittings and their joints shall be su it able for the tubing with which they are to be used in accordance with the minimum wall thickn es s of th e tubing and m e thod of a sse mbly re c ommend e d by th e manufacturer. (c) Fittings sha ll not be used in serv i ces that exceed the manufacturer

's maximum pressur e-temperature recommendations.

NB-3671.S Ca ulked Joint s. Caulk e d or lead e d joint s shall not b e u s ed. NB-3671.6 Brazed and So ld e r e d Joints. (a) Brazed Joint s ( 1) Brazed joint s of a maximum nominal pipe s i ze of I in. (DN 2 5) may be used only at dead end instrument connect i ons and in special applications where space and geometry c onditions prevent the use of joint s permitted under N B-366 1.2 , NB-3661.3 , and NB-3671.4. Th e d e pth of s ock e t s hall be at l e a s t equal to that required for s ocket weldin g fittin gs and sha ll be of sufficient depth to develop a rupture strength equa l to that of the pipe at Design Temperature (N B-4500). (2) Bra z ed joint s that depend upon a fillet rather than a c apillary type filler addition arc not acceptabl e. (3) Brazed joint s sha ll not be used in sy s tems containing flammable flui d s or in areas where fire hazard s are involved. (b) Solder e d Joint s. S o lder e d j o ints s h a ll n o t b e u s ed. NB-3671.7 S l eeve Co upl ed and Other Patented Joints. Mechanical joint s, for which no s t andards exist, and other patented joints may be used provided th e requirem e nt s of(a), (b), and (c) below are m e t. (a) Provi s ion i s made to pr e vent se paration of th e joint s under all Service Loading s. (b) They are accessible for maintenance , removal , and replacement after service. (c) Eith e r of the following two crite r ia ar e m e t. (1) A prototype joint has been subjected to performance tests to determine the safety of th e jo i nt under simu lat e d service condition s. Wh e n vibr a tion , fati g u e, c yclic c ondition s, low t e mp e ratur e, t h e rmal e xpan s ion , or hydraulic s ho ck i s antic ip ated , the applicable conditions s hall be incorporated in the tests. The mechanical joints shall be s uffic i ently l eak tight to s ati s fy th e r e quirem e nt s of t he Desi g n Specifi c ation s. (2) Joints are designed in accordance w ith the rule s of NB-32 00. 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.Al l ey@nrc.gov>; Hoffman, Keith <Ke i th.Hoffman@nrc.gov

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

>

Subject:

Mechanical Clamp in ASME Section Ill I did a word searc h of "Clamp" in NB and NC sect ion s of th e 2007 ed ition an d 2013 e dition of the ASME Code, Section Ill. There were 4 hits in both ed ition s. NB-113 2.1-the clamp in this article is r e l ated to pipe attachment (the clamp used for pipe s upport s) NB-341 1.l(d) pump clamp NB-3651.3 pipe clamp as in pip e supports such as hangers or snubbers tha t u se clamps. NB-4231---clamps used i n welding operations (when we l ding 2 pieces of pipe, welders u se clamps) So ASME SEct i on Ill does not have requirements or specificatio n for the m ec h an i cal clamp app l ication that was used on the CRDM canopy sea l at Wolf Creek.

From: Sent: To: Subje ct: Drake, J a m es 19 Sep 2016 11: 27:05 -0500 Alley, David RE: Response to questions on Wolf Creek penetration Thank you Dave. I s aac is r unning with the issue for now. I just got to Grand Gulf to start a heat exchanger inspection. Jim From: Alley, David S e 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

.go v> Cc: Anchondo, I s aac <l saac.Anchondo@nrc.gov

>; Dod s on, Dougla s <Dougla s.Dodson@nrc.gov

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

Proulx, David <D.av i d.Proulx@nrc.gov

>; Lyon, F r ed <F red.L yon@nrc.gov>

Subj e ct: RE: Response to qu es tion s on Wolf Creek pen e tration 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 calculat i on. 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, Jame s S e nt: Saturd a y, S e pt e mb e r 17, 201610:33 PM To: Werner , Greg <Greg.Werner@nrc.gov

>; A ll ey, David <Dav i d.Alley@nrc.gov

>; Tay l or, Nick <Nick.Taylor@nrc.gov> Cc: Anchondo, I s aac <lsaac.Anchondo@nrc.gov

>; Dod s on , Dougla s <Douglas.Dodson@n r c.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov

>; Proulx, David <David.Prou l x@nrc.gov>; Lyon, F r ed

<Fred.Lyon@nrc.gov>

Subject:

Fw: Response to questions on Wolf Creek penetration H e r e is the reply to the questions we asked th e lic ensee. Jim James F. Drake Email: James.D r ake@nrc.gov Office phone:817-276-6558 Cell phone: l (b)(6) I From: Muilenburg William T <wimui l e@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 th i s l eak. No verification of the st r uctural integrity of the joint is required. Westinghouse has calculated that the maximum leakage flow for one canopy seal i s about 3.5 gpm. The obse r ved leakage was l ess than the maximum value. The design of the mechanica l connection is that the canopy sea l we l d is a specially designed sea l between the housing (i.e. Cont r o l Rod Drive Mec h anism (CROM), head adapter plug o r CET) and the reactor vessel head adapter flange. The so l e function of the canopy sea l and sea l we l d is to provide RSC leakage control. The threaded connection between the adapter flange and the hous i ng, i ndependent of the canopy sea l , prov i des the structura l integrity for the pressure bounda r y items of the connection under all service loadings. With the failed seal weld, the leakage does not affect the threaded connection s i nce the mechan i cal connection is pressurized by the RCS and leakage past the threads is not a failure of the pressure boundary. W i th 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 sea l (or across the flaw). In this condit i on, the water does not flash to steam unti l the fai l ed surface or beyond so there is no steam cutting of the threads. Therefore, no impact on the structural integrity of the j oint will occur. 2. What is our plan to repair the penetration?

Install a canopy seal clamp on the l eaking penetrat i on.

3. If we intend to use the canopy seal clamp again, what is our basis and code that we i ntend to apply? The CROM S e a l Clamp As se mb l y i s an a lyz e d to ASME B&PV Code , Section 111, D i v i s i on 1 , 1986 Edit i on (No adde n da). T he Design Spe cifi cat i on for the CROM Seal C l amp Assembly i s ce r t i fied to ASME B&PV Code, Section Ill , D i v i s i on 1, 1 971 Ed. up to and includ i ng the W i nter 19 72 Addenda and th e 1974 E d. Th e De s i gn R e port for th e CROM Se a l C l amp A sse mbly is ce r t i fi e d to ASME B&PV Code , Se c t i on I ll, D i v i s i o n 1 , 1986 Ed i tion (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 l houn had had a simi l ar 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 programmat i c inspections.

From: Sent: To: Subje ct: Att a chment s: DSCF3786.jpg Ross-Lee, MaryJane 20 Sep 2016 15:19:03 -0400 Alley, David;McHale, John FW: Wolf Creek Status -09/20/16 DSCF3798.jpg, DSCF3797.jpg, DSCF3792.jpg, DSCF3789.j pg, DSCF3788.jpg, Mary Jone Ross-Lee (MJ) Deputy Director , Division of Engineering Off ice of Nuclear Reactor Regulation OWFN 9Hl US Nuclear Regulatory Commission Off i ce: 30 1-4 1 5-3298 : 1 e-ma i l: mary j ane.r oss-lee@nrc

.gov From: W i lson, George S e nt: Tuesday, September 20, 2016 1:08 PM To: Dean, Bill <Bill.Dean@nrc.gov>;

McDermott, Brian <Brian.McDermott@nrc.gov>;

Evans, Michele <Miche l e.Evans@nrc.gov>

C c: Boland, Anne <Anne.Bo l and@nrc.gov>; Benner, Eric <Eric.Benner@n r c.gov>; L ub i nski, J ohn <John.Lubinski@nrc

.gov>; Ross-Lee, MaryJane <MaryJane.Ross

-Lee@nrc.gov>

Subj e ct: FW: Wolf Creek Status -09/20/1 6 See t he attac h e d pictures o f the uncleaned RV H at WCGS , it does show some d eg r ada t ion. RI V se nt th ese to the ED O coord in a t or. Outage i nformation fr om R I V The li c ensee's RPV V i sual In s pe c tion pro c edure/a c tivity is s c heduled to c omplete a t -9: 00 PM tonight (09/20/16) a cco rding to the Outage S c h ed ul e i n th e OCC. How e ver , ba s ed o n the delayed start of th i s activity on yesterday , I am almost positive that t his will slip by at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Ba s ed on my conversat i on with the dayshift OCC Manager , the head c o nd i tion will n o t be fully ass e ss ed un til th e RPV in s pe c ti o n ac ti v ity i s co mp l et e d. George Wilson Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reacto r Regulation USN RC 301-415-1 7 11 Office 08E4 From: Lyon, Fred Se nt: Tuesday, September 20, 2016 10: 02 AM To: Pascarelli, Robert <R o b ert.Pascare ll i@n r c.gov>; A ll ey, David <Dav i d.A ll ey@nrc.gov>

Cc: Wilson , George <George.Wilson@n r c.gov>

Subject:

FW: Wolf Creek Status -09/20/16 See the attached pictures of the uncleaned RVH at WCGS. R I V sent these to the EDO coordinator, so if questions come down , no , it's not D-B. From: Thomas, Fabian S e nt: Tuesday, September 20, 2016 9:45 AM To: Taylor, Nick <N i ck.Tay l or@nrc.gov>; Proulx, David <Dav i d.Proulx@nrc

.gov>; Lyon, Fred <Fred.L yon@nrc.gov

>; Janicki, Steven <Steven.Janicki@nrc

.gov> C c: Dodson, Douglas <Doug l as.Dodson@nrc

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

> Subj e ct: 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 Act i vities: 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 detension i ng (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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Based on my conversation with the dayshift OCC Manager , the head condition will not be fully assessed unt i l the RPV inspection activity is completed.

Also, Mode 6 has slipped due to the aforementioned delays in the insulation removal and issues encountered w i th the Large Equipment Lift outside of containment (delay i ng placement of head detensioning equipment). Mode 6 is schedu l ed 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 unt i l 09/23/16 for breaker testing, which will be conducted in the switchyard. The residents have reviewed the outage risk assessment for this condition-i n accordance with the risk process Wolf Creek will be implementing risk mitigation actions when work is occurring i n 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: Sent: To:

Subject:

TS 3.4.13) Dave, Tsao, John 4 Sep 2016 15: 12:18 +0000 A ll e y , D a vid Re: Status 3 PM Saturday 9/3: TS Shutdown at Wolf Creek (RCS leakage thanks. I am glad that the phone ca ll with Wolf Creek went well. I am working on my apprai s als today and hopefully wi ll finish it by tomorrow. John F r om: All ey , D a vid Sent: S unda y , Sep te mb er 4 , 2016 8:50 AM To: T s a o , J oh n; L yon , Fr e d

Subject:

RE: Sta tu s 3 PM Satu rd ay 9/3: TS Shutd o wn a t W o lf C r e e k (R CS L ea ka ge TS 3.4.1 3) John, Phone call went well. You we r e not needed. L ooks l ike t his issue will just move up the start of their refue l ing outage (which was scheduled for later this month. I t 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 , Jo hn Sent: Saturday, September 03, 2 0 16 1 0: 27 PM To: L y on, F r e d <Fred.Ly on@n rc.go v> Cc: All ey , Da v i d <Da v id.All e y@nrc.g ov>

Subject:

Re: Sta tu s 3 PM Sat u rday 9/3: T S S h u t do wn at Wol f Cre e k (R CS L e ak a ge T S 3.4.13) Fred, I am s orry that i mi s sed your call to my home and I mi ss ed the 2 Pm phone call with the licensee.

I was out t oday and did not check my h ome phone for your voice mail until now (Saturday 10:15PM).

If you need me tomorrow Sunday or Monday my cell phone number is l .... <b-)(6_J ___ __, John From: Lyon , Fr e d Sent: Saturday, September 3, 2016 2:53 PM To: Pa s car e lli, Rob er t; All e y , David; T sa o, John; 'p as c a r e lli1991@v e rizon.n e t' C c: W i l s on , G e o r ge; Bo l and, Anne; K l e i n , Alex; Woo d yatt , D i ana; Grov e r, Rav i nd e r

Subject:

R E: S t a t u s 3 PM Satu r d a y 9/3: T S Shutdown a t Wolf C r ee k (R CS L ea k age 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 invo l ve 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. F rom: Lyon, F r ed Sent: Saturday, September 03, 2016 1 0:10 AM To: Pascarelli, Robert <Robert.Pascare ll i@nrc.gov>; Alley, David <David.Alley@nrc.gov

>; T sao, J ohn <John.Tsao@nrc.gov

>; 'pascare ll i199l@ve ri zon.net' <pasca r e l li1991@verizon

.net> Cc: Wi l son, George <George.W il son@nrc.gov

>; Boland, Anne <Anne.Bo l and@nrc.gov

>; K l ein, Alex <Alex.Klein@nrc.gov

>; Woodyatt, Diana <Diana.Woodyatt@nrc.gov

>; Grover, Ravinder <Ravinder.Grover@nrc.gov

>

Subject:

RE: Status 10 AM Sat u rday 9/3: TS Shutdown at Wolf Creek (RCS Leakage TS 3.4 .. 13) Import a nce: 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 l eakage contro l. 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 l icensee's repair plans. I've fo r warded the scheduler to you. Anne , George , D i ana, Alex , Rav i: I d on't t h i nk it's necessary for you to ca ll in, un l ess you fee l you need a piece of this pie. From: Lyon, Fr e d S e nt: Friday, Sept e mber 02, 2016 11:30 PM T o: P asca r e ll i, Ro be rt <Robert.Pascarelli@nrc

.gov>; A lle y, D a vid <David.Alley@nrc.gov

>; T sao , John <John.Tsao@nrc.gov

>; 'p asc ar e lli1 99 l@v e rizo n.n e t' <pascarell i 1991@verizon.net

> C c: Wilson, G e orge <George.Wilson@nrc.gov

>; Boland , Anne <Anne.Bola nd@nrc.gov

>; K l ein , Al ex <Alex.Klein@nrc.gov

>; Woody a tt, D i a na <Diana.Woodyat t@nrc.gov>; G rov e r , R a v i nd e r <Ravinder.Grover@nrc

.gov> Subj e ct: Status 11 PM F r i d a y 9/2: TS S h utdown at Wolf Cr ee k (RCS Leaka g e TS 3.4.13) Import an c e: Hi g h T he l icensee identified t he source o f the l eak as t he RVH, specifically, a nozzle tha t was repaired with a mechanica l clamp l ast outage (spring 20 1 5). I've no other informa t ion yet; t he l icensee is gathering drawings, taking photos, et al., to provide to RIV. Nick Tay l or, the R I V DRP BC, will set up a ca ll about m i dday t omorrow, specific time TBD, wi t h the l i censee to discuss whether t his is pressure boundary l eakage and how they intend to repai r it. I'll provide the i nformation when i t i s avai l able. Dave, Jo h n: J i m Drake , who apparently was involved in the inspect i o n of t h e repa i r l ast o u tage, r ecommended you listen in on the ca l l for consult to RIV. Perhaps he consulted you on the i ss ue? I did an ADA M S s earch back to 11/2014 , but did not find any relief reque s t that might ap p l y. Thanks, Fred From: Lyo n , Fred Se nt: Fr id a y , Sep t em b e r 0 2, 2 01 6 1 2: 1 0 P M To: P asca r e ll i, Ro be rt <Robert.Pascarelli@nrc

.gov> C c: Wilson, Ge o rge <George.Wilson@nrc

.gov>; Boland, Anne <Anne.Bo lan d@nrc.gov>; K l ein , Al ex <Alex.Klein@nrc

.gov>; W o ody a tt, D i ana <Diana.Woodyatt@nr c.gov>; G rov e r, R a v i n de r <Ravinde r.Grove r@nrc.gov> Sub je ct: Statu s: TS S h utdown at Wolf Creek (RCS L eakage T S 3.4.13) Imp or t a n ce: Hi g h On the 1 0 AM CT R I V/l icensee phoncon , the l icensee discussed the i n dications they have so far and p l a n s g o i ng f orwa r d (a ll time s CT). 1. Shutdown to M3 i s in progress and will be done abo ut 2 PM. They w i ll remain at NOP/N OT to enter the b i o s hie l d and search for the l eak source. I f they find th e sou rc e, th e l eak will then be i d e n t i fi e d l eakag e, and the T S l i mit is 1 O gpm; so t h e y will be ab l e to exit the c urrent TS require d act i on. Then , if possible , they would rema i n in M3 to c ondu ct r e p a ir s.

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 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. It is highly unlikely that NRC would grant a NOED for them to remain at NOP/NOT in M3 for troubleshooting activities.

2. I nside 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 l evel decrease rate is slightly increased.

No substantive temperature/humidity changes; there is a small steam leak on a SGBD valve (secondary side). T he 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 1 E-11 to 1 E-9 over the past week. Rad monitor filter levels are slightly elevated over the past week from 1 E-11 to 1 E-10, most l y 1-131 and 1-133; no short-lived isotopes have been detected.

There are no known fue l defects at WCGS. Licensee has not been able to correlate the indications that they have within containment to a particu l ar 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 ab l e to enter containment and do an initial search. From: Lyon, F r ed S e nt: Friday, S e ptember 02, 2016 10:15 AM To: P asca rel l i, Robert <Robert.Pa s c a r e lli@nrc.gov> Cc: Wi l son, George <George.Wilson@nrc.gov

>; Boland, Anne <Anne.Bo l and@nrc.gov>; K l e i n , Alex <A l ex.Klein@nrc

.gov>; Woodyatt, D i ana <D i ana.Woodyatt@nrc.gov

>; Grover , Rav i nder <R a v i nd e r.Grov e r@n rc.gov>

Subject:

UPDA T E: Tech spec shutdown at Wolf Creek th i s morning Importance

High T he 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (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 l ocated 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 1 O 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, F r ed S ent: Friday, S e pt e mber 02, 2016 8:1 2 AM To: Pa scare l l i, Ro be rt <Robert.P as c a r e lli@nrc.gov>

Cc: Wilson, George <Geo r ge.W i lson@nrc.gov

>; Boland, Anne <A n ne.Bo l and@nrc.gov

>

Subject:

FYI: Tech spec shutdown likely at Wolf Creek this morning Importan ce: 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: Pru e tt, Troy <Tro y.Pruett@nrc.gov

>; Lantz, Ryan <Ryan.Lantz@nrc.gov

>; Vegel, Anton <Anton.Vege l@nrc.gov>; Clark, Jeff <Jeff.Clark@nrc.gov

>; Kennedy, Kriss <Kriss.Kenne dy@nrc.gov>; Morris, Scott <Scot t.Morris@nrc

.gov> Cc: Warnick , Greg <Greg.Warnick@nrc.gov

>; Proulx, David <David.Proulx@n r c.gov>; Janicki, Steven <Steven.Janick i@nre.gov>; Dodson, Douglas <Douglas.Dodson@

n rc.gov>; Thomas , Fabian <F abian.T homas@nrc.gov

>; Lyon, Fred <Fred.Ly on@nrc.gov

>

Subject:

Tech spec shutdown likely at Wolf Creek this morning Importan ce: 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 yeste r day and overnight.

The resident inspector spent much of last evening at the site monitoring the si t e'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 l imit for un i dentified leak rate (TS 3.4.13). Action statement A requires them to reduce the leakage to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or enter action statement B, which wou l d require the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 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 t h eir plans are for the control room to commence a normal p l ant 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 l ook 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 n ear-cont i nuous dialogue with the control room t hrough the morning. Plant conditions (aside from leakage) are stable. Containment humidity and temperatures are steady & normal for these plant condit i ons. Pressurizer level is in the normal range and not trending (around 58%). The licensee is in their off-normal procedure for excessive RCS l eakage, which has scram actions should pressurizer level near 6%, but it does not appear that is a like l y 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 fo r 1000 this morning with their p l ant 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 t hrough the day with David and Steve. Thanks , N i ck N ic k Taylor Chief, Projects Br anch B D iv i sio n of Rea cto r Project s USNRC Reg io n I V 0: 817 200-1141 C: (bJ (6 J E: ni ck.ta y l o r@nrc.g ov From: Sent: To:

Subject:

Greetings , Collins, Jay 6 Sep 2016 20:32:14 +0000 Drake, Jame s Wolf Creek Boric Acid Leaking on Head 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 Fo rt Calhoun and the cleaning of their head last year. I dona*Žt know the in spec t ion requirements for Wolf Creek this r efueling ou tag e, but I figure they are at least going to have to clea n the head for a VT-2 inspe ct ion. Cleaning the head in too aggressive of a manner can i nva lidate the visual in spection and may then trigger a vollumetric in spect i on. Ju s t a heads up f o r a p r ob l em they may n ot be thinking abo ut , using l essons l earned that Region IV caught earlie r at Fort Calhoun. This was l saaca*Žs issue at Fort Calhoun , so I a m su r e h e h as all the fine d eta il s. Ju st trying to be h e lpful , Jay From: Sent: To:

Subject:

Tsao, John 3 Oct 2016 1 7:28:52 +0000 Alley, David RE: Internal commun i cations at Wolf Creek re head corrosion I p l ans Dave, Yes we shou l d be on the ca l l w i th Wolf Creek tomorrow From: Alley, David Sent: Monday, October 03, 2016 1:21 PM To: Tsao, John <John.Tsao@nrc.gov>

Subject:

FW: I n te rnal communicatio ns 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: Mond ay, October 03, 2016 1:08 PM To: Alley, David <David.Alley@nrc.gov

> Cc: Taylor, Nick <Nick.Taylor@nrc

.gov>

Subject:

FW: I nternal communicatio n s at Wolf Creek re head co rr os ion/ 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 , Ry an <Ryan.Lantz@nr c.gov>; Clark, Jeff <Jeff.C l ark@nr c.g o v>; Prou lx, David <D a vid.Prou l x@n r c.gov>; Janicki, Steven <Steven.Janicki@nrc

.gov>

Subject:

Internal communications at Wolf Creek re head corrosion I plans All , I'm s till working on se tting up a ca ll with th e li censee tomorrow.

But Doug provided the attached today f rom the licensee's CAP and intern al outage newsletters.

I added the red co mment boxes.

Th a nk s, Nick From: Sydnor, Christopher Sent: 4 Oct 2016 20:10:02 -0400 To: Alley, David;Anchondo, l saac;Baquera, Mica;Bloodgood, Michael;Bozga, John;Brand, Javier;Bu rk et, Elise;Butcavage, Alexander;Carrion , Robert;Case, Michael;Chaudhary, Suresh;Clayton, Kelly;Collins, Brendan;Collins, Jay;Cumblidge, Stephen;Drake, James;Dykert, Jason;Farnholtz, Thomas;Floyd, Nikla s;Ga vula, James;Gray, Harold;Hills, David;His er, Allen;Ho l mberg, Mel;Honcharik, John;Jayr oe, Peter;Jones, William;Jose, Benny;Karwoski, K enneth; Kaufman, Paul;Kopriva, Ron;Lupold, Timothy;Makor, Shiattin;Meghani, Vijay;Mitchell, Matthew;Modes, Michael;Neurauter, Jame s;Nove, Carol;Poehler, J effrey;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, J ay;Wi lli ams, Robert;Young, Matt; Huang, John;Tsao, J ohn;Thomas, Brian; Dunn, Darrell;Tregoning, Robert;Jandovitz, John;Davis, Robert;Widrevitz, 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, J effrey;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, Jo e l;Dijamco, David;Cooper, Pau la

Subject:

September 2016 MECC Meeting Minutes

Dear MECC Call Participants ,

Ple ase see th e following meeting minut es 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 summa ry. 1. Wolf Creek Leakage through Threaded Connections (Isaac Anchondo & Greg Warner. RIV, DRS): Control room operators calcu lated unidentified reactor coolant syste m leakag e of 1.358 gallons per minute , in excess of the TS 3.4.13 limit of 1 gallon per m in ute. The licensee entered containment and discovered a non-pressure boundary leak from a canopy sea l weld on a reactor vessel head penetration th at serves one of the co re exit thermo coup le s. Th e lic ensee has opted to use a CROM Seal Clamp Assembly as a repair method. The licensee has just i fied the structural integrity of the threads with a Westinghouse calcu l ation show ing a maximum allowab l e l eak of 3.5 GPM. Region I V staff is still in the process of review i ng this issue with the help of NRR. During the call, RIV staff ind icated that the main concern is that actual thread degradation i s causing the leak. The licensee is trying to justify that there's no th read degradation, the l eak is due to the canopy seal weld , and that the repair u s ing the CROM Seal Clamp Assembly is adequate as a cor r ect iv e act ion. S taff emphasized the concern that the li censee is n't addressing the potential root c ause of thread degradation, nor adequately justifying why the threads are s till intact. The Westinghou se ca l culation is current ly being r ev i ewed , but ther e are conce rn s. EPNB sta ff proposed that i f it can be established that only a s mall leak rate could be expected for the sea l weld with undamaged threads, then the NRC would be in a good position to ask the licen see to provide a rigor ous eva luation of th e structu r a l in teg rit y of th e threads. Conversely, if undamaged threads can, in fact, l eak 1.358 GPM then the NRC may not necessarily be i n th e best pos ition to request a m o r e rigorous eva luation. Other points that were r aised on this pertained to boric acid i ss u es as a re s ult of the l ea kag e, whether th e r e's ot h e r industry OE with leakage resulting from thread damage, and design c rit eria.

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 Imp r ovement ," and associated staff SE. The outgoing NRR SE is publically available at ADAMS Access i on No. ML 16208A485 and the NRR/DE/EPNB non-concurrence document is publically availab l e at ML 16187A319. The NRR contacts for this are Stephen Cumblidge , Jay Col li ns, David A ll ey , 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 re l ated 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 i f they're not accurate).

Staff also discussed a Millstone Code Alternative to implement UT in li eu 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 us i ng UT in l ieu of RT for non-safety-related piping. 4. NRC Generic Letter 90-05 (Isaac Anchondo, RIV/DRS):

There was question regarding li censee Relief Request subm i ttals based on GL 90-05. Cooper had this issue with piping less than the minimum wall thickness. Per G L 90-05 , licensees must perform Code repairs or request NRC to grant relief for temporary non-Code repairs on a by-case basis regardless of pipe size for a ll 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 repa i rs. I'll send out the scheduler for the next MECC call on Wednesday, October 19. Thanks , Chris C hri st oph e r R. Sydnor Materi a l s Engine e r V esse ls a nd Int e rn a l s Int eg rit y Br a n c h NRR/Div. of E n g in ee rin g USN R C (3 01) 41 5-6 0 65 Office: 0-9Hl 4 C hr ist op he r.Sy dnor@nr c.g ov From: Sent: To: C c: Subj e ct: Drake, J a m es S Oct 2016 1 1:11:10 -OSOO Werner, Greg;Taylor, Nick Alley, David;Anchondo, Isaac RE: Relief request coming from Wolf Creek Code requires the seal weld for threaded connections on Class 1 component.

From: Werner, Greg S e nt: Wednesday, October OS, 2016 11:10 AM To: D r ake, James <J ames.Drake@nrc.gov>;

Taylor, Nick <Nick.Taylor@nrc.gov>

C c: Alley, David <David.Alley@nrc

.gov>; Anchondo, Isaac <lsaac.Anchondo@nrc.gov>

Subj e ct: 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 r equire a sea l weld? Don't remember?

From: Drake, James S e nt: Wednesday, October OS, 2016 11:07 AM To: Werne r , Greg <Greg.Werner@nrc.gov

>; Taylor, N ick <Nic k.Taylo r@nrc.gov> Cc: A ll ey, David <David.Alley@nrc.go v>; Anchondo, Isaac <lsaac.Anc h ondo@nrc.gov> Subj ec t: RE: Relief request coming from Wolf Creek Code requ i res the flaw be removed or restored to an acceptable condition as pa rt 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 F r om: Werner, Greg S e nt: Wednesday, October OS, 2016 11:04 AM To: Taylor, Nick <N i ck.T ay l or@nrc.gov> Cc: Alley, David <David.Alley@nrc.gov>; Drake, James <James.Drake@nrc.gov>; Anchondo, I saac <l saac.Anchondo@nrc

.gov> Subj e ct: RE: Relief request comi n g 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 speculat ion on my part. Greg From: Taylor, Nick S e nt: Wednesday, October OS, 2016 10:22 AM To: Singa l , Ba l want <Balwant.S i nga l@nrc.gov>; Werner, Greg <Greg.Werner@nrc

.go v>; Kopriva, Ron <Ron.Kopriva@nrc

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

.gov>; T homas, Fabian <Fabian.Thomas@nrc

.gov>; A l lley, David <Dav i d.Alley@nrc.gov> Cc: Proulx, David <David.Pro u lx@nrc.gov>; Janicki, Steve n <Steven.Jan i cki@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 H afenstine). She was calling to correct one thing they told us yesterday.

They have apparently decided to request rel i ef from performing the volumetric inspection on Penetrat i on 77 only (not a ll 12). She did not know the basis for the request, nor did she know when they wou ld be ready to submit the req ue st. Thi s 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: n i ck.t a l o r ov From: Sent: To: Subje ct: Att a chment s: Tsao, J ohn 5 Oct 2016 13:33:33 -0400 Collins, Jay FW: Pictures of WC penetration DSC00024.jpg, DSC00026.j pg Fy i wo l f creek C R O M boric acid depos i ts From: Drake, James S e nt: Wednesday, October OS, 2016 11:45 AM To: Alley, David <David.Alley@nrc.gov>;

Tsao, John <John.T sao@nrc.gov>;

Hoffman, Keith <K e ith.Hoffman@nr c.g ov> Subj e ct: Pictur es of WC p e netration These are pict ur es of Penetratio n 77. W ill t ry to ge t cop i es o f addit i onal p i ct ur es. Jim

'f". :Drafe J ames F. Drake Off i ce pho n e: 8 1 7-200-1558 Ce ll Phone: l{b){6) I

From: Sent: To:

Subject:

Greetings , Anchondo, Isaac 11 Oct 2016 15:01: 15 -0500 T sao, John;C ollin s, Jay WC Call -Item of Note I was just thinking, what happens if Nozzle 77 & 78 are included in the nozzles to be UT/Leakpath given that th ey are also requesting relief from the examina tion volume for those two penetrations?

I ju st wanted to point that ou t as food for thought since we didn't ask them on the call. Thanks, R eactor In spector U.S. Nuclear Regulatory Commission I Region JV Di vis ion of R eac tor Safety I Enginee1ing Bran ch 2 (817) 200-1152 From: Sent: To: Subje ct: Att a chment s: DSCF3786.jpg Yes. Coll i ns, Jay 11 Oct 2016 20:49:54 +0000 Anchondo, Isaac RE: WC Call -Item of Note DSCF3798.jpg, DSCF3797.jpg, DSCF3792.jpg, DSCF3789.j pg, DSCF3788.jpg, Hey could you confirm that these are pictures from Wolf Creek this outage? Jay From: Anchondo, Isaac S e nt: Tue s day, 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*Žt seem to be a way for Cooper to clean it without having a*cerelevant indicationsa*

L left in place. But isna*Žt this the reason they are performing the volumetric examinations?

From: Co ll ins, Jay Sent: T u esday, October 11, 2016 3: 35 PM To: Anchondo, Isaac <lsaac.Anchondo@nrc.gov

> Subj ec t: 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 S e nt: Tuesday, October 11, 2016 4: 20 PM To: Collin s , Jay <Jay.Collins@nrc

.gov>

Subject:

RE: WC Call -Item of Note la*Žm not the inspector on-site. Would you like me to ask Ron Kopriva to give you a call sometime tomorrow?

Isaac From: Collins, Jay Se nt: Tuesday, October 11, 2016 3:16 PM To: Anchondo , Isaac <l saac.Anchondo@nrc

.gov>; Tsao, John <John.T sao@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 apprec i ate 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: Tu es day, October 11, 2016 4:01 PM To: Tsao, John <John.Tsao@n r c.gov>; Collins, Jay <Jay.Collins@nrc.gov

> Subj e ct: WC Call -Item of Note Greetings, I was just thinking, what happens if Noz z le 77 & 78 are included in the nozzles to be UT/Leakpath given that they are also requesting relief from th e examination volume fo r those two penetrations?

I just wanted to point that out as food for thought since we didna*Žt ask them on the call. Thanks, Reactor I nspector U.S. Nuclear Regulatory Commission I Region IV Division of Reactor Safety I Engineering Branch 2 (817) 200-1152 From: Sent: To:

Subject:

Remova l Att a chment s: DSCF3786.jpg Coll i ns, Jay 11 Oct 2016 16:41: 29 -0400 Tsao, John;A ll ey, Dav i d; Hoffman, Keith FW: W olf Creek RX Vesse l Head P ictu r es -As-Found during M irr or I nsu l at i on DSCF3798Jpg, DSCF3797Jp&

DSCF3792Jp&

DSCF3789Jp&

DSCF3788Jp&

These were reported to be from Wo l f Creek, and a l so part of my concern. I do not know if you guys have seen these. J ay From: Sent: To:

Subject:

Coll i ns, Jay 11 Oct 2016 14:00:33 +0000 Singal , Balwant Accepted:

Wo l f Creek Relief Request -Outage Support From: Singal, Ba l want Sent: 11 Oct 2016 11:37:22 -0400 To: A ll e y , David;Coll i n s, J ay; T sa o, John; Mui l e nbur g Will i am T C c: Li ng am, S i v a;Smith Stephen L;Reasoner Cleve O;Taylor, Nic k;Hafenst i ne Cynthia R;Werner, Greg;Anchondo, lsaac;Drake, James; Kopriva, Ron S u bj ec t: W o lf Creek Relief R e qu es t -Outage Support Att a chm e nt s: [E xt e rna l_Sender] Wolf Creek R e l ie f Requ es t Anticipated, [External_Sender] Wo l f Creek Rel i e f Requests 14R-03 & 04 Revi se d to attac h e-mail w i th copy o f the r e li efrequest.

No re s pon se n eeded. Thanks. Pl ease see the attached e-ma il from WCNOC requesting a phone ca ll i n support of th e re li efreq uest in s upport of the curre nt refueling outage. The forma l relief request is to be s ubm itted by noon today. I will pass on th e copy of the re li ef r equest to everyo n e after r eceip t. P l ease use t h e following Bridge No. for thi s ca ll. 866-624-3402 Passcode j (b)(B) j# Thanks.

From: Sent: To: Subje ct: Ba l want, Muilenburg Wil l iam T 7 Oct 2016 21:17:37 +0000 Singal, Balwant [External_

Sender] Wolf Creek Re l ief Request Anticipated 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 (ML 12353A241 provided NRC Safety Assessment of the request), and second, we will be asking to perform a n alternate exam v. that specified in code case N-729. Can you he l p us assemb l e the right peop l e to have a phone call regarding this request on Monday morning? I wi l l be in Saturday and Sunday if there are any questions I cain help answer. Thanks, Bill Muilenburg 620-364-4186 From: Sent: To: Cc:

Subject:

Att a chment s: Ba l want, Muilenburg William T 11 Oct 2016 15:23:50 +0000 Singal, Balwant 'Nicho las.Taylor@NRC.Gov';Dodson, Do u glas [External_Sender)

Wolf Creek Re l ief Requests 14R-03 & 04 pr222_.PDF See l i sting of Records A l ready Ava i lable to the Pub l i c for attachment.

H ere is the relief requests for discussion this afternoon.

Thanks for arrangi n g today's call, Bill Muilenburg From: Sent: To: C c: Nick;Groom, Jeremy Subj e ct: Att a chm e nt s: lingam , Siva 12 Oct 2016 10:59:57 -0400 jaknust@WCNOC.com;wimuile@WCNOC.com Pascare ll i, Robert;Singal, Balwant;Alley, David;Tsao, John;Coll i ns, Jay;T aylor, Wolf Creek--DRAFT RA l s for Relief Reque s t 14R-04 CROM nozzl e e xamination Wolf Creek RAl .. docx Attached p l ease find t he draft RA l s f or t h e subject relief request. Ba l want is l (b)(6) this morning , and wi ll be wo rk ing t h is af t ernoon. Ba l wa n t wi ll forwa r d th e official RA l s later. Siva P. lingam U.S. Nuclear Regulatory Commis s i on Project Manager (NRR/DORL/LPL4

-1) Palo V er de N u clear G e n era ting Station Locat io n: 08-05; Mail Stop: 08-B3 Te l ephone: 30 1-415-1 564; Fax: 301-41 5-1222 E-ma i l address: si v a.lingam@n r c.gov REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST 14R-04 ALTERNATE EXAMINATION OF CONTROL ROD DRIVE MECHANISM NOZZLE PENETRATIONS WOLF CREEK GENERA TING 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 i ts rev i ew , the Nuclear Regulatory Commission (NRC) requests the following additional information.

1. (a) Discu s s whether CROM nozzle numbers 77 and 78 w i ll be ultrasonically examined during the current fall 2016 refue l ing outage as part of the supplement examinations i n accordance with Re l ief 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 examinat i on distance will be reduced further for nozzle numbers 77 and 78 in the futu r e inspections. If yes , the li c ensee needs to propose an examinat i on 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 ML 12193A559) with supplement dated October 15 , 2012 (ADAMS Accession No. ML 12341A228), the licensee proposed alternate examination distance s for CROM no z zle 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 obta i ned during the 2006 inspection. By letter dated January 4 , 2013 (ADAMS Accession No , ML 12353A241

), 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 w e re 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 show s the threaded region; however, Figure 1 is not clear regarding the examination d i stance with respect to the location of the J-groove weld. Provide sketches (hand ske t ches are acceptable) of the region below the J-groove weld for CROM nozzle numbers 77 and 78. The sketch should be similar to Figu re 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 we l d to the bottom of the CROM nozzle numbers 77 and 78. (b) Identify the threaded region which is approximately 1.19 inches as s tated in the relief request, (c) Identified the required "a" distan c e 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 pred i ction. (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: Sent: To: C c: Subj e ct: All, Singal , Balwa n t 12 Oct 2016 15:56:23 -0400 Alley, David;T sao, John;Collins, Jay Pascare ll i, Robert;Lingam, Siva Wolf Creek Relief Request -Status Su r prisingly, I never heard from anyone from Wolf Creek about the r e lief re que st need e d in s upport of the current refueling outage since the discussions yeste r day and was unable to reach anyone in their licensing department. I just spoke to their Reg. Affairs Manager, Cindy Hafens t ine and she indicated that they are st ill wo r king on revising the relief request. We are expected to get it either by end of the day today or ear l y tomorrow morning. I ind icated to her that it w i ll be very difficult to process it and provide the verbal by F r i day. Thanks fo r everyone's patience.

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

From: Sent: To:

Subject:

Isaac, Tsao, John 12 Oct 2016 10:32:3 0 +0000 Collins, Jay;Anchondo , I saa c RE: WC Call -Item of Note I p lan to ask the licensee whether they will inspect nozzles 77 and 78 in th i s outage. From: Collins , Jay Sent: Tuesday, October 11, 2016 4:16 PM To: Anchondo , Isaac <l saac.Anchondo@nrc.gov

>; Tsao, Jo hn <John.Tsao@nrc.gov>

Subject:

RE: WC Call -Item of Note They are ones that they wo u ld have to perform the inspe c tion on. The volumetric leak path is performed on the nozzle above the weld. The limitation to inspection coverage i s b e low the weld. Therefore , not a spec i fic 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, I saac Sent: Tuesday , October 11 , 2016 4:01 PM To: T sao, John <John.Tsao@n r c.gov>; Collins, Jay <Jay.Colli n s@nrc.gov>

Subject:

WC Call -Item of Note Greetings , I was just thinking , what happens if Nozzle 77 & 78 are inc l uded in the no zzles to be UT/Leakpath given that th ey are also requ esting relief from th e exam ination volume for thos e tw o penetrations?

I just wanted to point that out as food for thought since we d i dn't ask them on th e call. Thanks , R e a c t o r In s p ec t o r U.S. N u clear Reg ul ato r y Commission I R eg i on I V Di v i s i on of R eac t or Safe t y I E n g in ee ri ng Branc h 2 (8 1 7) 200-11 52 From: Sent: To: Cc:

Subject:

Attachments:

Balwant, Tsao, John 12 Oct 2016 09:59:02 -0400 Singal , Balwant Lingam , Siva;Alley, David;Collins, Jay Wolf Creek--RAI for Relief Request 14R-04 CROM nozzle examination Wolf Creek RAl .. docx Attached are my draft RAI questions regarding the subject relief request. Dave Alley has not reviewed my draft RAI ques ti ons. I am forward the m to you because the urgency of the review. Please forward my RAI questions to the licensee as "d raft" Thanks John REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST 14R-04 ALTERNATE EXAMINATION OF CONTROL ROD DRIVE MECHANISM NOZZLE PENETRATIONS WOLF CREEK GENERA TING 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 i ts rev i ew , the Nuclear Regulatory Commission (NRC) requests the following additional information.

1. (a) Discu s s whether CROM nozzle numbers 77 and 78 w i ll be ultrasonically examined during the current fall 2016 refue l ing outage as part of the supplement examinations i n accordance with Re l ief 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 examinat i on distance will be reduced further for nozzle numbers 77 and 78 in the futu r e inspections. If yes , the li c ensee needs to propose an examinat i on 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 ML 12193A559) with supplement dated October 15 , 2012 (ADAMS Accession No. ML 12341A228), the licensee proposed alternate examination distance s for CROM no z zle 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 obta i ned during the 2006 inspection. By letter dated January 4 , 2013 (ADAMS Accession No , ML 12353A241

), 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 w e re 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 show s the threaded region; however, Figure 1 is not clear regarding the examination d i stance with respect to the location of the J-groove weld. Provide sketches (hand ske t ches are acceptable) of the region below the J-groove weld for CROM nozzle numbers 77 and 78. The sketch should be similar to Figu re 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 we l d to the bottom of the CROM nozzle numbers 77 and 78. (b) Identify the threaded region which is approximately 1.19 inches as s tated in the relief request, (c) Identified the required "a" distan c e 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 pred i ction. (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, J ohn Sent: 12 Oct 2016 06:38:00 -0400 To: Collins, Jay;Alley, David; Hoffman, Keith Subje ct: RE: Wolf Creek RX Vesse l Head Pictures -As-Found during Mirror Insu l at i on Remova l The wolf creek's RPV head l ooks like the corroded RPV h ead at Dav i s Besse. Do we want to ask wo l f 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 satisfact i on. Fr o m: Co ll in s, Jay S e nt: Tuesday, October 11, 2016 4:41 PM To: Tsao, John <John.Tsao@n r c.gov>; A ll ey, David <David.Alley@nrc.gov>; Hoffman, Ke i t h <K e ith.Hoffman@nrc.gov

> Subj e ct: FW: Wolf Creek RX Vessel Head Picture s -As-Found du ri ng Mirror I nsulation Removal These were reported to be from Wo l f Creek, and a l so part of my concern. I do not know if yo u guys have seen these. Jay From: Sent: To: Subje ct: Alley, David 14 Oct 2016 00:32:46 +0000 Davidson, Evan RE: Input to the DE Per i od i c Meeting [)ue 10/12 If you are in tomorrow (which will hopefully be today if you are reading t his on Friday), lets try to have the meeting we tried to schedule for Wednesday. Dave From: Davidson, Evan S e nt: Wednesday, October 12, 2016 1:54 PM To: Alley, David <David.Alley@nrc.gov>

Subject:

RE: I nput to the DE Periodic M eeting Due 10/12 Wh at a bo ut y o u r S OI? Any pro g r ess or d ecisio ns a f ter re vi e w i n g th e a ppli ca nt s? F r om: Alley, David Se nt: Tuesday, October 11, 2016 8:48 AM To: Ross-Lee, M aryJane <MaryJane.Ross

-L ee@nrc.gov>; NRR_DE_DO Distr i bution <NRRDEDODistr i bution@nrc

.gov>; A l varado, Rossnyev <Ro ss nyev.A l varado@n r c.gov> Subj ec t: 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 i s quite at the level of the item s 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 like ly 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

<NRRDEDODistribut i on@nrc.gov>; A l varado, Rossnyev <Ros s nyev.A l var a do@nrc.gov>

Subject:

FW: I nput to the DE Periodic Meet in g D u e 10/12 Import a nce: High Reminder that input is due. We meet with Brian on Thursday. Mary Jan e Ross-L e e (MJ) Deputy Director , Division of Engine e ring Off i ce of Nuc l ear Reactor Regulation OWFN 9H1 US Nuclear Regulatory Commission ir Office: 301-415-3298 e-ma i l: mary j ane.r oss-l ee@n r c.gov From: Davidson, Evan Sent: Thursday, October 06, 2016 11: 23 AM To: NRR_DE_DO Distribut ion <NRRDEDODistribution@nrc

.gov>; Sacko, Fanta <Fanta.Sacko@nrc

.gov>

Subject:

In put to the DE Periodic M eeting Due 10/12 The next DE periodic briefing with Brian McDermott is sc heduled for October 13. Please take a look at the attached b ri efing sheet from last mon th and provide updates for your branch a l ong with suggestions for a ddit io n s or de l etions. If you can get it to me by 10/12 that would b e grea t. I can also s top by to go over the list in person if that's fas ter. (Today or Wednesday)

Thanks, Evan From: Sent: To:

Subject:

Att a chment s: Coll i ns, Jay 13 Oct 2016 15:39:56 -0400 Tsao, John;Kalikian, Roger FW: Leak Location leak l ocation.pdf 1 s-p a ge attachment withheld in full under ex4. Some potential additional nozzles to discuss .... Regional discussion?

From: Anchondo, Isaac S e nt: Thursday, October 13, 2016 2:41 PM To: Collins, Jay <J ay.Co ll ins@nrc.gov>

C c: Werner, Greg <Greg.Werner@nrc.gov>;

Proulx, David <David.Proulx@nrc.gov>;

Kopriva, Ron <Ron.Kop r iva@nrc.gov>;

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

Subject:

Leak L ocation J ay, I made an attempt to identify t h e l ocation of the nozzles to be inspec t ed from some of the p i ctures that R o n too k t h is pas t Saturday.

U nfortunate l y, i t's ha r d to see t h e v i c i ni t y of some o f t he nozz l es th at you q u estion during th e ca ll (i.e., 52, 34 , 26). One i te m of n o t e , t h e view n ext t o the vent lin e i s a b i t co n fus in g whe r e it's h ard to figu re o u t w h ich s i de yo u a r e l ooking at i n relation to t h e draw i ng provided.

H ope this h e l ps. Re actor Insp ector 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 Hoffman, Keith To:

Subject:

Re: NRC Report for code meetings I wi ll l eave that one to Dave. I doubt we are l ook in g at removin g 25% sa mple fo r all MSIP. I think if any actions are taken it i s go in g to be more focused. In a week or two we w ill hold anot h er call with Calvert/Westinghouse to discuss the RES we ld residual stress analysis.

From: Hoffman , Keith Sent: Thursday , October 13, 2016 9:48:06 AM To: Co llin s, Jay

Subject:

RE: NRC Report for code m ee tin gs What about for Calvert Cliffs? This is what I put in the last report. Calvert Cliffs Un it 1 Pre ssurizer Safety Relief N ozz le to Safe-end Weld LERN o. 3172016002, ADAMS ML 16106A304)

U l trasonic (UT) examinations pe rformed at Calvert Cliffs Nuclear Power Plant, Unit 1 id entif i ed a change from previous examinations in an axial fl aw in a pressurizer safety relief nozzle to safe-end weld that was mitigated by the Mechanical Stress Im provemen t Proces s (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 thic kn ess. UT exam in atio n s prior to the applica ti on of MSIP identified an ax i al flaw in the same l ocation as the 2016 flaw but a depth of 8% through-wall.

UT following MSIP confirmed th e flaw was still present at a depth of 8% through-wall.

The ISi examinations in 2010 reported essentially no change in the through-w all depth of the indication.

G iven this informa tion , the NRG is conside r i ng rulemaking action to e liminat e the allowance of a 25% sample of welds mit igated by MSIP. The NRG is also considering r equ iring 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 r ecommendations on how I s hould update this? Keith M. Hoffman Materials Engineer NRR/DE/EPNB (301)415-1294 From: Co ll ins, Jay Sent: Thursday, October 13, 2016 9:43 AM To: Hoffman, Keith

Subject:

Re: NRC Report for code meetings Greetings, The Wolf Creek item s hould pr obably include details r egard in g while this was not a pressure boundary l eak, th e boric acid leak above the head caused difficulti es in performing the head visual in spec tion. It might be good to wait until after the relief requests a r e completed t o decide h ow to write that up.

On the second one, the acceptance criter i a is based on the genera l surface exam acceptance cr i teria as i t is a repair and not necessar i ly a req u irement of N-729. Just t h e comment. Jay Fr o m: Hoffman, Keith S e n t: Thur sda y, Octob e r 13, 2016 8:54:12 AM To: Coll i ns, Jay Subj ec t: NRC Report for code meetings Jay I wa s pl a nning on discussing the two events sho wn below i n the NRC Report do have a ny comme nt s 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 Whi l e operati n g i n MODE 1 at 100 pe r ce n t r ated t h e r ma l powe r and placi n g Excess L etdown i n service for Reactor Coolant System (RCS) lea k detection, RCS operational l eakage exceeded 1 gpm [ga ll on per minute] unidentified l eakage as i de n tified by perform in g RCS Water Inventory Ba l a n ce us i ng t h e Nuclear P l a n t Informat i on System Co m pute r. Th i s requ i red t he Unit to be p l aced i nto Mode 3 i n 6 h ours. Trending of containment sump l evel ind i cated the l eakage was ins i de containment with the exact location within containment unknown. The l icensee made a conta i nme n t ent r y and eve n t u a ll y found t h e source of the un i de n tified l eakage. W h ile looking down on the vesse l head the licensee identified signs of a bor i c acid l eak over a m i rrored insulation panel. After removing the pane l and using a camera the lice n see saw a p l ume in the area of severa l penetrat i ons. T he li censee was ab l e to determine that the l eak was on a core ex i t thermocoup l e nozzle threaded connection.

The l icensee also determined that this was not pressure bounda r y leakage. In additio n, t h e l icensee ident i fied t h at excess letdown m ade t h e l eak rate seem worse than the actua l va l ue. The leak rate was eve n tua ll y q u ant i fied at around 0.6 gpm. Wit h out being pressure boundary leakage and since the l eak ra t e was less t h an 1 gpm, t h e l i censee was ab l e to exit the L CO. T he licensee h as dec i ded to go into their p l anned refuel i ng o u tage and wi ll perfo r m some pre-outage survei ll ances before cooling down to MODE 5. The l eak wi ll be repaired dur i ng the r efue lin g outage w hi le the h ead is on t h e sta n d. This one I was going to put under RV Head Penetration In spections BRAIDWOOD 1 (EN 52275)-LIQUID PENETRATION EXAMINATION R E SULT S IN INDI C ATIONS ON REAC T OR V ESS EL H EAD PENET RA T I ON During the Br aidwoo d Station U nit 1 Refu e ling outage (A1R19), an inservice Liquid P enetration examination w as performed on the previously repa i red control rod drive mechan ism (CROM) penetration

69. Dur ing the exa mination on th e w eld build up for CROM p e n e tration 69, two indi cat ion s were discovered. A 7 /32 inch rounded indicatio n was discovered located at 359 degrees on the reactor head portion of the weld buildup, and it i s 4 inche s from the tran sition of the he ad to penetration.

A 1/4 inch rounded indication wa s a l so discovered located at 200 degrees at the trans i t i on 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 Insp ections? With a description of what was found @ WC and the difficulties they are having performing the exams and whatever rel ief they are requesting.

Keith M. Hoffman Materials Engineer NRR/DE/EPNB (301)415-1294 From: Sent: To: Cc:

Subject:

Singal , Balwa n t 13 Oct 2016 16:09:26 -0400 Tsao, John;Collins, Jay Pa s carelli, Robert;Alley , David;lingam, Siva FW: WCNOC RV pictures I GUESS T H ERE W IL L BE ANOTHER E-MA I L PROVIDING ACCESS. I W I LL REQUEST THEM TO PROVIDE ACCESS TO SIVA AND J AY AS A M I NIMUM. T HANKS. Balwant K. Singal Senior Project Manager (Diablo Canyon and Wolf Creek) Nuclear Regulatory Commission Division of Operating Reactor Licensing Ba l want.Singal@nrc.gov Te l: (30 1) 415-3016 Fax: (301) 415-1222 ... From: Good N icole R [m ailto:ni l yon@WCNOC.com)

Sent: Thursday, October 13, 2016 4:05 PM To: lingam, S i va

Cc: Singal, Balwant <Balwant.Singal@nrc.gov>

Subject:

[External_Send e r] WC N OC RV pi c ture s I was told you would like pictures of t h e penetrations with labels of the penetration number. I have on l y 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 helpfu l (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.

T hank you, Nicole Good Licensing nil yo n@w c no c.co m (620) 364-8831 x 4557 Wolf Creek , Nu cleor Operohng Corporol 1on From: Sent: To: Cc:

Subject:

0052 Attachments:

Dave, Singal, Balwa n t 13 Oct 2016 07:32:46 -0400 Alley, David Lingam , Siva; Pascarelli, Robert;Tsao , John;Collins, Jay FW: Wolf Creek -Draft revision of Relief Request Document Number WO 16-W016-0052R5dt.pdf 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 Ba l want.Singa l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 .. ... " ... .,;;,.,, ... ) .... . ... .. From: Haf ens tine Cynthia R [1]

Sent: W ednesda y, October 12, 2016 6:11 PM To: Singal, Balwant <Ba lw ant.S i nga l@nrc.gov>;

's i va.l i ngman@nrc.gov

' <siva.lingman@nrc.gov>

Cc: Muilenburg William T <wim uil e@WCNOC.com>;

Tou ga w Dennis E <detouga@WCNOC.c om>; Barraclough Richard M <r ib arra@WCNOC.com>

Subject:

[External_Sender] Wolf Creek -Draft revision of Relief Request Document Number WO 16-0052 Attached is ou r current draft revision of t h e relief request. We have not yet incorporated the questions listed in the dra f t RA I that you provided.

We would lik e to have a call on Thursd a y at 1:00 pm Eastern Time/ Noon Central Time. Please l et me know if that will work for you. We appreciate your su pport in getting this document revised to s upport our request. Thanks, Cindy H afenst ine Office 620-364-42 0 4

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

Reference:

Subject:

Gentlemen:

were examine (WCGS) I s hutdo reactor vessel head penetration no zz l es 29-1 and both Wolf Creek Generating Station nd Bor ic 1d Program s. A canopy seal weld leak led to the boric acid accumulation from the canopy seal weld leak he subject of the head inspection.

The boric acid te visu inspection of 12 nozz l es because th e canopy seal weld y bo on from a nozzle leak. Wolf Creek Nu c lear Operating dent that the observed deposits were the result of the canopy to no zz le int e rfa ce areas w e re obscured s u c h that adequate vi s ual in s pection s s ible on the top side of the head. Be cause of this , WCNOC will be performing a supple examination of the obscured nozzles from the unders i de of the head in accordance with C e Case N-729-1. WCNOC is reque s ting relief from the requir eme nt 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 reques t for the WCGS , lnservice Inspection (ISi) Program, fourth ten-year interval.

The a ttachment iden tifies t he affected components , a pplicable American Society of M echa ni ca l Engineers Bo iler and Pressure Vessel Code (ASME) Code re quirements , reason for request , proposed alternative, and bas is for pro posed a lternative. The alternatives are proposed to b e ap pli ed during I nt erva l 4, which b ega n September 3 , 2015 and e nd s 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 in s pection protocol for the remainder of ISi Int erva l 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 refue l ing outage. In add i tion, pu r suant to 10 CFR 50.55a(z)(2), Wolf Creek Nu clear Operating Corporation (WCNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of 10 CFR 50.55a Request Number 14R-04 for the Fourth Ten-Year Int e rv a l of WCNOC's ln se r vi c e Inspection (ISi) Program. The attached 1 O CFR 50.55a Request (14R-04) requests rel ief from certai n ASME Code Case N-729-1 requirements for examination of reactor vessel ead penetrat i on welds. (Attachment

2) Thi s request is similar to that requested in the Ten Year Int erva l of WCNOC's lnservi ce Inspect i on (ISi) that was accepted by ML 12 241. The Code of Fed e ra l Regulations 10 CFR 50.55a(g)(6)(ii)( the reactor vessel head be performed in accordance with conditions specified in paragraphs 10 CFR 50.55a chosen by WCNOC to perform these examinatio cove r age below the J-groove weld on two contr Both of these CROM penetrations are configur s required by N-729-1 cannot be met. Attachment 2 t 04 , documents the ultrasonic coverage limitations.

WCNOC had intended to reque s t this re but the circumstances described above h penetrations be performed at this time rin requests approval of the attached 1 O CFR support inspection and res om Refuelin November 14 , 2016. s t hat exam i i nations of e N-729-1 subject to (6). The vendor d examination etrations. on distance Request 14R-tions in Refueling Outage 23

• ion of one of the sub j ect Outa 21. Therefore , WCNOC -04 by October 14, 2016, to is now schedu l ed to comp l ete questions concerning this matter , please e (620) 364-4204. COR/r l t Attachments

1) 1 0 CFR 50.55a Request Number 1 4R-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. Ta ylor (NRC), w/a Senior Resident Inspector (NRC), w/a Sincerely , Cleve l and Reasoner Attachment 1 to WO 16-005 2 Page 1 of 6 Wolf Creek Nuclear p 10 CFR 50. ance with (2)

Attachment 1 to WO 1 6-0052 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 C l osu r e Head Code Class: Class 1 Examination Cate o 8-P Code Item Number: ative Examinati on e l Upper Heads with ia l-Pene tration Des cri ption: Size: Material: 2. 3. Paragraph

-3 *1er and Pressure Vessel Code (ASME ddenda 50 .55a(g)(6)(i i)(D) (1) r e res that examinations of th e reactor vessel head be 't h A E Code Case N-729-1 subject to the condi tion s specified a(g)(6)(ii)(D)(2) through (6). The supp/em al examination performed to satisfy -3 142.2 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 w eld , 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.

Attachment 1 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 spray 7 deflection 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 noz zle 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-729-1 -3200(b). A third party vendor has been cont ract ed to perform the examinations.

The options for the surface examination of the Attachment 1 to WO 16-0052 Page 4 of 6 5. 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 re su lt 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. each pe verify there in the penetra 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 1eve s that the combination of the volumetric exams and ill pr ide an acceptable level of confidence in the condition of ecause , as shown in the figure below , the two examinations will ns in the nozzle tube and verify that there has been no leakage VCH interface.

Attachment 1 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 we l d. Performing the leak path asse groove weld will demonstrate that the head were a result oft opy seal we l ath assessments.

The table below li sts the Combo-2 OHS Combo-2 Combo-2 35.2 Combo-2 35.2 Combo-2 38.7 Combo-2 38.7 Combo-2 4 4.3 Combo-2 45.9 Combo-2 4 5.9 Combo-2 77 48.7 OHS eld measured in 2013 using the axially xia l coverage was at l east 2 inches above 2013 Coverage Obt ai ned Above Weld (Axial Shooting) in inches 3.22 3.64 3.00 3.44 3.04 2.76 2.80 3.12 2.60 2.96 3.20 3.32 Attachment 1 to WO 16-005 2 Page 6 of 6 6. BE VERY SPEC I F I C REGARDING EXAMS TO BE PERFORMED

-WHAT PROBES, Zero Degree Etc, and details of coverage -Follow-up c om me nt from our PM The Open Housing Scanner (OHC) uses Type PSC-24 TOFD 5 MHz transducers with a refrac ted angle of 55° for the circumferent i al shoot i ng and a refracted ang l e of 40° for the ax i a l shooting. The Combo-2 blade probes use Type PSC-23.5 TOFD 6.2 MHz transducers w i th a refracted ang l e of 57° for the circumferent i al shoot i ng and a refracted angle of 44° for the axia l shoo ti ng. In both the OHC and Combo-2 probes , the search units utiliz exam ination have a nominal frequency of 2.25 MHz. WCNOC has exami n ed t he RCV H prev i ously in 2006 n degradation of the RCVH. The data from the e compared to the data from the previous exams. the le ak path data. WCNOC believes that the estimated ad used), the added time penetration welds Assessment and REM depending on method ace examination of the partial fit over the proposed L eak Path artial penetration welds does not result in r safety. utilized during WCNOC Refueling Outage 21 on ly. nor inspection protocol for the remainder of ISi Interval 4 , 015 and ends on September 2 , 2025.

Attachment 1 to WO 16-0052 Page 7 of 6 7. Precedent

8. References

1. ASME Boiler and Pressure Ve ssel Code Case N-729-1 " Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzl es Having Retaining Partia l-Penetration WeldsSection XI , Division 1" 2. NUREG CR 7142 , "Ult r asonic Phased Array As sess ment of the Interferen c e Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Val i dation 3. WDl-T J-0-03-P , " Ultrasonic Testing of In terference Fit Sam Leak Path Detection (PWROG PA-MSC-0532)"

Attachment 2 to WO 16-0052 Page 1of13 Wolf Creek Nuclear Attachment 2 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 2. 3. Code C l ass:

Reference:

It em No.:

Description:

P aragra ph -2 drive mechanism (CROM) J-oove weld that attaches erside of the head for 008 Addenda, as augmented by ASME ative Examination Requirements for PWR Hea ozzles Having Pressure-Retaining Partial-1, Divi s ion ,"as amended by 10 CFR 50.55a(g)(6)(ii)(D).

(0)(1) requires that examinations of the reac tor vessel head be ce with ASME Code Case N-729-1 s ubj ect to the co ndit io n s s 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6). of Code Case N-729-1 states , in part: I f obstructions or lim i tation s prevent examination of the vo lum e or surface req uired by Figure 2 for one or more nozzle s, the analysis proc edu re of Appendix I shall be used to demonstrate the adequacy of the examination volume or surface for each such nozzle. If Appendix I is u sed, the evaluation s hall be subm itted to the regu l atory authority having jur isd i ction at the p l ant site.

Attachment 2 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 i sta n ce below the toe of the J-groove weld (i.e. dimension "a") be 1.5 inches for incidence ang l e , 8 , less than or equal to 30 degrees; 1 inch for incidence ang l e , 8, greater than 30 degrees; or to the end of the tube, w h ichever i s l ess. Th ese cove ra ge requiremen t s are applicable to Wo l f Creek Generating Station (WCGS) reactor vessel head penetrations as shown in Table 1. Table 1: WCGS Reactor Vessel Head Penetration Coverage Requi re ments Penetration Numbers 1to29 30 to 78 4. Reason for Request *red Coverage , "a" inches 1.5 1.0 styles of ends, referred to gh 73 are T ype "Y" that are er diameter and inner diameter. meter and a n internal taper. t i on nozzles 7 4 through 78 , referred to as , approximately 1.19 inch i n length at the re l ocated at the 48.7 degree l ocation. T he a t th 1 1s such that the distance from the lowest point Id to the top of the threaded region could be l ess than the "a" shown in Figure 2 of ASME Code Case N-729-1. ired inspect i on coverage is sought for reactor vesse l , as the required coverage for these two penetrations A t tachme nt 2 to WO 1 6-0052 Page 4 of 1 3 5. T h e table be l ow lists t h e cove r age obta i ned on nozzles 7 4-76 du r ing the 2006/2013 exams p e rfo rm e d per NR C O r de r EA-03-009 (2006) and N-7 29-1 (20 1 3). N o t e: T he l ower measu r e m ent i n 2006 was performed us i ng circ u mfe r en ti a l shoot i ng T O FD t r a n s du ce r s w hil e the 20 1 3 m e a s u re m e nt s we re acc om plis he d us in g ax i a l shooti n g T OFD tra n sducers. While the ta bl e below shows d i fferen t coverage va l ues it i s noted i n the 2 01 3 exam report that t he "Lowe r ex t ent compar i so n us i ng Cha n nel 2 da t a s h o w s no c h a n ge from 200 6 t o 2013 meas ure ments." Penetration 0 (degrees)

No. 74 4 8.7 75 4 8.7 7 6 4 8.7 N-729-1 Requi r ed Exam Coverage (inches) 2013 Inspection Coverage Obtained inches 1.00 1.08 .00 design weld size o r contou r is

• side o f th e pe r i p he r a l net r at i ons 77 and 78 , as we l d, r esul ti ng in less of t he e thre ads) be i ng ava il a bl e f or sse l hea d penetratio n welds pelffo r me d in t w as previo u sly s ubmi tted fo r i nab ilit y to eferences 3 and 4 ). T his p r ev i ous r equest i n F or t h e exam i natio n s pe rf o r me d i n 2 01 3 in , as co n d i t i oned by 1 0 C F R 50.55a, ano t her si mil ar reques t fe nces 7 , 8, & 9). e vol u metr i c and surface examination coverage requireme n ts "a" in F igu re 2 of ASM E Code Case N-729-1 , WCGS proposes the ul trason i c exam i nat i on d i s t ances shown i n Ta b le 2. T he r eq ui red exami n at i on c e r age d i mens i o n for the ot h e r pe n et r a ti ons w ill be met or exceeded.

Attachment 2 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 of an a lt ernative examination area or volume to tha t N-729-1 if impediment s prevent examination of the ASME Code Case N-729-1 requires, for alter groove we l d, that analyses shall be performe is p r ocedure for eva lu ation (Section 1-2000) or the deterministic frac 3000) to demonstrate that the app li cable in Sec tion 1-2000 were validated in WCAP-1 i n WCAP-16589-P was reviewed.

Th e stress a ove r the ent i re region outside t 1 6589-P analysis was compa r ed t it was determined that the require analysis in WCAP-16589-P was a l so 3000. S i nce the Figure 2 of Code Case e. Section 1-1000 of ones be l ow the analysis method d Section described ltern ative 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 Sec tion 1-3000 was also validated 5.1 Stres s An a lysis i n Accord a nce with ASME Code Case N-729-1 Section 1-2000 Section 1-2000 of ASME Code Case N-7 29-1 requires th at plant-specific analys i s demonstrate th at th e ho op and axial st re sses remain b e low 20 kips per square in c h (ksi) (tensile) over the entire region o u tside the alternative examination zone but wit hi n the examination zone defined in Figure 2 of the Code Case.

Attachment 2 to WO 16-0052 Page 6of13 5.2 The distance below the J-groove weld that requires examination, as determined by the po in t at whic h the CROM penetrat i on hoop st r ess dist ri butio n for the operating stress l evels i s 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 ax i a l stresses.

The hoop stress distribut io n p l ots for penetrations 77 and 78 are provided in Figure 2 of this submitta l. T h e hoop stress dis tr ibution plots i n F igu r e 2 indicate that t he min i m u m achievab l e inspect i on coverage below the bottom of the J-gr weld insures stresses remain below 20 ksi tensile over the entire region outs i d zone but within the examinat i on zone defined in Fig u re The hoop stress distribution plots display the downhill stress distribution plots shown are for the ins id *sis more limiting.

Also, surface. T able 3 summarizes the distance f r om below the toe of t o both the inside and outside surface hoop stre Ciro and 78. Pen e t ra t i on Nozzl e No. D is t a nce Below Toe o f i Sid e J-Groov e W e ld ere Hoop Str ess used i since the Appendix techn i cal requ r < 2 0 k s i inch 0.30 ed and documented in Reference

2. The ta po ral crack i n the unexamined zone w ill not grow d prior to the examirnation frequency specified in T ab l e 1 of wa repared prior to approva l to use Code Case N-729-1. ced EPR I MRP-55 as the source for the crack g r owth formula t Appendix 0 as required by Code Case N-729-1. However , for la for crack growth rate is used in both EPR I MRP-55 and rs no tec h n i ca l d i fference , and WCAP-1 6589-P does meet the ents for l-3200(a).) (1) Th e fo ll owi n g tab l e provides the dimensions fo r nozz l es 77 and 78 fo r both the des i gned and as-bui l t configurations. The actua l we l d height was measured using the ultrasonic test data and is listed for the as-built dimension.

Attachment 2 to WO 16-0052 Page 7of13 Penetration Nozzle Number As-designed As-built (inches) (inche s) 77 1.4 6 1.98 78 1.46 2.04 Th e flaw evaluatio n i n WCA P-1 6589-P is based on the as-designed J-groove weld dimensions w h ich assumed a smaller weld throat than the as-built condition.

Often, the as-built fillet weld d i mension on the downhill side of the CRD ozzle is larger than the 'on. When t he weld tion nozzle due to a larger below the J-groove weld on similar CRDM ds have a reduced e below the weld we l d heights of s profiles input from the prev i ous l y own as f l aw to l erance require i nspection coverage.

T his of the penetration nozz l e n ot before the next inspection. The ented in F igure 3. tes that a postulated through-wa ll flaw at xamination zone will not grow to the toe of al of four refue l ing cycles. The crack growth n six e e fu ll power years (E F PY) of operation required t he toe of the we l d. Addit i ona ll y, the assumed i nit i a l upper h-wall flaws are conservative based on achievable e assumed upp.er crack extremities are l ocated wit h in of port' nozzle sign i ficantly below the J-groove we l d is not omena of concern, which include leakage through t h e J-groove ntial cracking in the nozzle above the J-groove weld. I n all cases, erage is adequate to allow WCGS to continue to operate p ri or to the hypothetical aws reachi n g the J-groove weld. I n 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 Surfac e E x amin at ion 10 CFR 50.55a(g)(6)(ii)(D)(3) states in part that "if a surface examination is being substituted for a vo l umetric examination on a portion of a penetration nozz l e that is Attachment 2 to WO 16-0052 Page 8of13 6. 7. 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 in spected using remotely operated volumetric examination equipment.

Although dye penetrant test ing of threaded surfaces is possible , it is not practical.

The threaded outside diameter (OD) makes a dye penetrant exam i nation on the l ower section of the penetration i mpractical because of excessive bleed out from the threads. Eddy current examination would simi l arly not be effective due to the threaded configuration. rrent known radiation levels under the reactor vessel head are 4.5 Rem/hr a t th o of 1 CROM nozzle. This could result in an exposure of approximately

1. per noz z le using 4500 mRem/hr and 20 minutes/nozzle. At this time our e f dose rates (based on recent measurements in the area) r ange 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. The alternati the current 2025. will be applied for the remaining duration of I) Int erva l which ends on Septembe r 2 , granted to the following plants: Safety dated December 22 , 2009 , for San Onofre Nuclear ting S ion , Units 2 and 3 , "Relief Request I S l-3-29 , Request fo r Relief e n Requirements of ASME Code Case N-729-1 fo r Control Element chanism Penetrations (TAC N os. ME0768 and ME0769)" <:i 1035)

• NR C Safety Evaluation dated March 3, 20 11, for Braidwood Station Un i ts 1 and 2 , and Byron Station Unit s 1 and 2, "Re l ief Request from ASME Code Case N-729-1 Requirements for Examination of Reactor Vessel Head Penetration Welds (TAC Nos. ME3510 , ME3511 , ME3512 and ME 3513)" (ML 110590921)

Attachment 2 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. 1 3R-07 for the Third 10-Year l nservice Inspection Program I nterval (TAC No. ME9078)

Attachment 2 to WO 16-0052 Page 10 o f 13 8. References

1. ASME Code Case N-729-1 , "Alternative

• Examination Requirements for PWR Reactor Vessel Upper Head s with Nozzle s Having Pre ss ure-Retaining Penetration Welds ,Section X I , Division 1 ," March 28, 2006. 2. WCAP-16589-P , Revision 0 , " Structural Integr i ty Evaluation of Rea cto r Vesse l 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 l etter ET 06-0048 from T. C , "Additional egarding e J-Groove ," I nformation Related to the F irs Requirements for Nondest ructive Ex November 1, 2006. 5. NRC letter from D. Ter Generating Station -Req Pressure Vessel Head Pe (TAC NO. MD3210)," Dece uench , WCNOC , " Wolf Creek i ve Examination of Reactor evised Order EA-03-009

6. 7. evised NRC Order (EA-03-009) equirements For Re acto r Pre ssu re Vesse l s," February 20 , 2004 . . 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 Additiona l In formation Regard i ng 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, " Wo l f Creek Gener ng Station -Request for Relief No. 13R-07 for the Third 10 Year l nseNice Inspection Program l nteNal (TAC NO. ME9078)," January 4, 2013.

Attachment 2 to WO 16-0052 Page 11of13 Figure 1 WCGS Reactor Vessel Head Penetration Ends 0Eii' Set

'J" . .. ','I ;,_ ;> I I : / '

/ *in' '""'/

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."

Attachment 2 to WO 16-0052 Page 12 o f 13 =-* -* ! U) a 0 0 x Figure 2 H oo p S tr ess Di s tributi o n Downhill Side (48.7° C RDM Pea etra tl o n Nozzle) 80 , 000 ------...------..-------.....------.....-----

--. I I 70 , 000 ************

J ********** I **'*-*** I '**---*****--* I I 60.000

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• • • • I I I I .C0 , 000 ****** J **********

• • • I I t I 30 , 000 ' 20 , 000 I I I ... , ..*........

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• • --****** I I I *20.00 0 J-------------L-----*-*****

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______________________

_.... ___ -1 00 0.5 1 0 1 s 20 25 D i stance from B o tt om of We l d (In) I-+-Ins i de -o-outside I Attachment 2 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 m j ** 1. I -:c;.:c..f.c*

__ : cf:.*cc.c*c*c*.cj*

• ___ ,_, __ _ ,,_, __ , __ -1 )" 0 0 0 --t i I . ' 9 I "

From: Sent: To: Cc:

Subject:

Att ach ment s: Singal, Balwant 13 Oct 2016 13:59:46 -0400 Collins, Jay;Tsao, John Alley, David;Lingam, Siva;Pascarelli, Robert FW: FW: Relief Reques t for Code Case N-729-1 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 Ba l wa nt.S i nga l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 .; .. From: Hafenstine Cynthia R [mailto:cyhafen@WCNOC.com)

Sent: Thur sday, October 13, 2016 1:0 1 PM To: Singal, Balwant <Ba lwant.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 re lief 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, Ba l want Sent: 13 Oct 2016 09: 33:18 -0400 To: C o llin s , Ja y;T s ao, Jo hn;Lin ga m , Siv a; P asca r e lli, Rob e rt;All e y , D a vid;T a ylor, Nick;Proulx , D a vid;Drake, J ames;We rn er, Greg;Anchondo, l saac;Kopriva, Ron Subj e ct: Wolf Creek Reli e f Reque s t Att ac hm e nt s: [Ext e rn a l_ S e nd e r] Wolf Cr ee k -Dr a ft revi s ion of R e li e f R e qu es t Document Number WO 16-0052 P l ease see t h e attac h ed e-ma il fro m WCNOC w i th a d raft vers i o n of the rev i sed re li ef r eq u est. WCNOC h as r eq u es t ed a ca ll t o d i sc u ss th e r e li e fr e qu es t with t h e N R C staff befo r e i ss uing t h e fo r m a l r e q ues t. T h e r e li ef i s n eede d in su p port of t he c ur re nt r e fu e ling o ut age. Brid ge No. fo r th e ca ll. 86 6-62 4-340 2 P assco d e: l (b)(G) I# T hanks.

From: Sent: To: C c: Subj e ct: Number WO 16-0052 Att a chm e nt s: Hafenstine Cynth i a R 12 Oct 2016 22:10:42 +0000 Singal , Balwant;'siva.lingman@nrc.gov' Mu i lenburg William T;Tougaw Dennis E;Barraclough R i chard M [External_Sender)

Wolf Creek -Draft revision of Relief Request Document W016-0052R5dt.pdf Attached is our current draft revision of the re l ief request. We have not yet incorporated the questions listed in the draft RAI that you provided.

We wou l d like to have a call on Thursday at 1:00 pm Eastern Time I Noon Central Time. Please let me know if that wi ll work for you. We appreciate your support in getting this document revised to support our request. Thanks, Cindy Hafenstine Offr 620-364-4204 Cell (b X6J I From: Sent: To: Subje ct: 0052 Tsao, J ohn 13 Oct 2016 08:14:02 -0400 Collins, Jay;Singal, Balwant;A ll ey, Dav i d RE: Wolf Creek -Draft revis i on of Relief Request Document Number WO 16-I am available at 1 Pm also From: Co ll ins, Jay S e nt: Thursday, October 13 , 2016 8: 07 AM To: Singal, Ba l want <Ba l want.Singa l@nrc.gov>;

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

C c: T sao, John <John.T sao@nrc.gov> Subj e ct: RE: Wo l f Cree k -Draft revi s io n of R e lief Requ es t Document Number WO 16-0052 I am available at lpm. I have a call at 2pm. From: Singal, Ba l want S e nt: Thursday, October 13, 2016 7:33 AM To: Alley, David <Dav i d.Alley@nrc.gov> Cc: Lingam, S i va <Siva.Lingam@nrc

.gov>; Pasc a re ll i, R obert <Robert.Pascare ll i@nrc.gov>; T sao, J ohn <J ohn.Tsao@n r c.gov>; Co ll ins, Jay <Jay.Coll i ns@nrc.gov

> Subje c t: FW: Wolf C r ee k -Draft revision of Re l ief Request Document Numb er 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 forma l. 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 Ba l want.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 ... .... ) ..... F r om: Hafen s tine Cynthia R [ma i l to:c yhafen@WCNOC.com

] S e nt: W ed n esda y, Octob er 12, 2016 6:11 PM T o: Singa l , B a l w ant <Balwant.S i nga l@nr c.gov>; 's i va.lin g man@nrc.go v' Cc: Muilenburg Wi ll iam T <wimuile@WCNOC.com

>; Tougaw Dennis E <detouga@WCNOC.com

>;

Barraclough Richard M <ri ba rr a@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 re l ief request. We have not yet incorpora t ed the questions listed in the draft RAI that you provided.

We wou l d like to have a call on Thursday at 1: 00 pm Eastern Time I Noon Central Time. Please let me know if th at 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: Sent: To:

Subject:

Tsao, Joh n 13 Oct 2016 14:40:23 +0000 Singa l , Ba l want Accepted: W olf Creek Rel i e f Request From: Sent: To: Cc:

Subject:

Att a chment s: Importance:

Burkhardt, Janet 14 Oct 2016 06:09:33 -0400 FRN_ReviewRequest M e ndiola, Doris;Lingam, S i va FW: WCGS LT FRN MF8168 MF816 8-FR N.docx Hi g h Also , can you please remove the red footnote when you return the clean copy? We cannot l istserv a copy to the licensee l ike that when it's signed and dated. Thank you. From: Singal, Ba l want Sent: Thursday, October 13, 2016 3:14 PM To: FRN_ReviewRequest

<FR N_ReviewReq uest@nrc.gov>

Cc: Burkhardt, Janet <Janet.B urkhardt@nrc.gov>;

Lingam, Siva <Siva.Lingam@n r c.gov>

Subject:

FW: WCGS LT FRN Mf8168 Importance:

High Attached is the copy of the FRN for early FRN review. Thanks. Balwant K. Singal Sen i or Project Manager (Diablo Canyon) Nuclear Regulatory Commi ss ion Division of Operating Reactor Licensing Ba l wa nt.Sing al@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 From: Burkhardt, Jan et Sent: Thursday, October 13, 2016 3:0 1 PM To: Singa l , Ba l want <Balwant.Singa l@nrc.gov> Cc: Lingam, S iv a <Siva.Lingam@nrc

.gov>

Subject:

WCGS LT FRN MF8168 Import a nce: High Ba lw ant, Please forward the at t ached to FRN ReviewRequest@nrc

.gov for ear l y F RN rev i ew (cc me an d Siva). I'll return the hard copy pack age to Siva tomorrow after I reprint i t. Please note: if you are not goi n g to be here to sign the FRN, the s i gnatu r e block on the FRN (only) will need to be changed to the person sig ning i t.

[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: N uclea r Regu l atory Commiss i on. ACTION: Application for i ndirec t transfer of li ce n se; opport u ni t y t o comment, r equest a hear i ng, and pet i t i on fo r leave to i ntervene.

SUMMARY

The U.S. N uclear Regu l atory Comm i ss ion (NR C) received and is cons i dering approva l o f an indirect l icense t ransfer appl i ca ti on fi led by Wolf Creek N uclear Operat i ng Company (WC N OC) on Ju ly 22, 2016. WC N OC is the li censed operator of Wo lf Creek Generat in g Station (WCGS). Kansas C i ty P ower a n d Li ght Company (KCP&L) and K ansas Gas and Electric Compa n y (KG&E) are two o f t h e three non-operat in g owner licensees, each hold i ng 47 perce n t undiv i ded i nterest i n WCGS and 47 pe r cent of the stock o f WCNOC. KCP&L is a s u bsidiary of Great P lains Energy Inco r porated (Great P l ains) a n d KG&E is a subsidiary of Westar Energy I ncorpora t ed (Westa r). Th e i ndirec t license transfer will resu l t f rom th e proposed merger of Great Plains and Wes t ar, wi t h Westar as w h o ll y-owned subsidiary of Great P l a i ns. Every a tt empt h as been made to make these t empla t es a ll-inclusive. However , every FR N may involve case-specific i ssues re l ated to it , requiring certain modifications of th i s template. I f you believe that a particula r part o f this temp l ate does no t app l y to your s p ec i fic rule please check wi t h the ADM/RAD S Reg ul at i ons Specialist assigned t o you r working gro u p or e-m ai l FRN ReviewReques t@nrc.gov.

DATES: Comments must be filed by [INSERT DATE 30 DAYS FROM DATE OF PUBLICATION IN THE FEDERAL REGISTER]. A request for a hea ri ng must be fi l ed 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 desc r i bes a different method for submitting comments on a specific subject):

• Federal Rulemaking Web Site: Go to h ttp://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.Gallaghe r@nrc.gov. For technical questions contact the individua l listed in the FOR FURTHER INFORMATION CONTAC T section of this document.

• E-mail comments to: Hear i ngdocket@nrc

.gov. If you do not receive an automatic e-mai l reply confirming rece i pt , then contact us at 301-415-1677.

• Fax comments to: Secretary, U.S. Nuclear Regu l atory Commission at 301-415-1101.

• Mail comments to: Secretary, U.S. Nuclear Regu l atory 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 add i tiona l direction on obtaining informat i on and submitting comments , see "Obtaining I nformat i on and Submitting Comments" in the SUPP LEM ENTARY INFORMAT I ON section of th i s document.

2 FOR FURTHER INFORMATION CONTACT: Balwant K. Singal, Office of Nuclear Reactor Regulation , telephone:

301-415-3016, e-mail: Balwant.Singa l@nrc.gov; U.S. Nuclear Regulatory Commission, Washington DC 20555-0001.

SUPPLEMENTARY INFORMATION:

I. Obtaining Information and Submitting Comments A. Obtaining In formation Please refer to Docket ID NRC-2016-XXXX when contac ti ng the NRC about the avai l abi li ty of information for this action. You may obta in publicly-available in fo r mation related to this action by any of the following methods:

• Federal rulemaking Web Site: Go to http://www.r egu l at i o n s.gov and search for Docket ID NRC-2016-XXXX.

• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-avai l able documents online in th e ADAMS Publi c Documents collection at h ttp://www.n r c.gov/readin g-rm/a d a m s.h tm l. To begin the search, select " ADAMS Publ i c Docu m ents" and then se l ect " B egin Web-based ADAMS Searc h." F or prob l ems with ADAMS , please contact the NRC's Pub li c Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by e-mail to p d r.resou r ce@n r c.gov. The application for i ndirect tran s fer of the license dated July 22, 2016, is available in ADAMS at Accession No. ML 16208A250.

• 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 Do c ket ID NRC-2016-XXXX in the your comment submission.

The NRC cautions you not to include identify i ng or contact i nformation that you do not want to be publicly disclosed in your comment submission.

The NRC will post all comment submissions at http://www.r egulations

.gov as we ll 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 infor m those persons not to include identifying or contact information that they do not want to be publ ic ly disclosed in their comment submission.

Your request should state that the NRC does not routinely edit comment subm i ssions to remove such information before making the comment submissions available to the pub li c or entering the comment into ADAMS. II. Introduction The NRC is considering the issuance of an order under§ 50.80 of tit l e 10 of the Code of Federal Regulations (1 O CFR) approving the indirect transfer of contro l of WCGS, F acil ity Operat i ng License No. NPF-42 , cur r ently held by WCNOC. WCNOC i s the licensed operator of WCGS. KCP&L and KG&E are two of the three non-operat in g owner l ic ensees , each holding 4 7 percent undivided interest in WCGS and 47 percent of the stock of WCNOC. Kansas Electric Power Cooperative, Incorporated (KEPCo) holds res t of the 6 percent undivided inter est in WCGS a nd 6 per ce nt of the stock of WC NOC. KCP&L i s a s ubsidiary of Great Plains Energy In corporated (Great Plains) and KG&E is a subsidiary of Westar Ene r gy In corporated (Westar). The indirec t lice n se transfer will result from the proposed merger of Great Pla i ns 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 the i r respective 47.0 percent interests in WCNOC and WCGS. KCP&L and KG&E w i ll 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. T he remaining 6.0 percen t 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 1 O CFR 50.80 state that no l icense , or any right thereunder , shall be transferred, directly or indirectly, through transfer of contro l of the license, unless the Commission gives its consent in writ i ng. The Commission wil l 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 qua l ifications of the licensee to hold the l icense, and that the transfer is otherwise consistent with applicab l e provisions of l aw, regulations, and orders issued by the Commission Ill. Opp ort un i ty t o Co mm e nt 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 s hould be submitted as described in the ADDRESSES sect i on of this document.

IV. Opportunity to Request a Hearing and Petition for Leave to Intervene Within 20 d ays after t he dat e of publication of this notice, any pe r sons (petitioner) whose interest may be affected by t his action may file a r equest for a hearing and a pet i tion to intervene (petition) w i t h respect to the action. Petitions shall be filed in accordance with the Comm is s i o n's " Agency Rule*s of Practice and Pro cedu re" in 1 0 CFR part 2. Interested persons shou l d consult a current copy o f 10 C FR 2.309 , wh i ch is ava i lable at the NR C's PDR , l ocated at One White Flint N o rth , Room 01-F 2 1 , 11555 Rockville Pike (first floo r), Rockville, M aryland 20852. The NR C's regu l ations are access ible electron i cally fro m the NR C Library on the NRC's Web site a t http://www.nrc.gov/r ead i ng-r m/doc-collect ions/cfr/. If a pe tit ion is filed within 20 days , the Commission o r a presid i ng office r des i gnated by the Comm i ssion or by the Chief Administrative Ju dge of the Atomic Safety and Li ce n sing B oa rd Panel, wi ll rule o n the pe t ition; and t he Secretary or the Chief Admin is trative Judge of the Atomic Safety and Lic ens ing Board will issue a notice of a hearing or an approp ri ate o r der. As r equired by 10 CFR 2.309, a petition shall set forth w ith particularity the i nterest of th e petitioner in the proceedi ng , a n d h ow th at int e r est may be affected by th e re su lt s of the p r oceed ing. Th e peti ti o n sho u ld specifica lly exp l ai n the r easons why intervention shou l d be permitted with part i cu l a r reference to the fo llow in g general r equireme nt s: (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 th e proceeding; (3) the nature and exten t of the pe tit io n er's p r ope rty , financ ia l, o r other i nterest in the proc eed ing; and (4) the pos s i b l e effect of any dec is i on 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 opin i on which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also prov i de 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 proceed i ng. The contention must be one which , if proven , would entitle the petitioner to rel i ef. 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 i ntervene must be filed no later than 20 days from the date of publication of this notice. Requests for hearing , pet i tions 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 sati s fying the three factors in 10 CFR 2.309(c)(1 )(i) through (i i i). 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 T he pet i tion s h ould sta t e the nature and exte nt of the pe tit ioner's i nterest i n the proceed in g. The pet i tion s h ould be submitted to the Comm i ssion by [I NSERT DATE 20 DAYS FROM THE DATE OF PUBLICATION IN THE FEDERAL REGISTER]. The petition must be filed in accorda n ce with the fil ing i nstruct i ons in t he "Electronic Submiss i ons (E-Fi li ng)" sectio n of t his docume n t, and s h ou l d m eet t h e r equireme n ts for pet i tions set forth i n this sect i on , excep t that under 10 CFR 2.309(h)(2) a State , loca l governmen t a l body, or Fede r a ll y-recognized I nd i an Tribe, o r agency thereof does not need t o address the standing requ i rements in 10 CFR 2.309(d) i f the faci li ty is l ocated w ithi n its bounda r ies. A State, l oca l governmental body, Federa ll y-recognized I ndian Tr ibe , or agency the r eof may also have the opportu ni ty to part i cipate u nder 10 CFR 2.315(c). I f a heari n g is granted, any person who does no t wish, or is no t qua li fied, t o b eco me a pa rt y to t h e proceeding may, in t h e discretion of the presid in g office r , be permitted to make a limited appearance pursuan t to the provisions of 1 O CFR 2.315(a).

A person mak i ng a li m i ted appearance m ay make an ora l or written statement of p osi t ion on t he issues , but may no t otherwise partic i pate in t he p r oceeding.

A limi t ed appearance may be made at any sessio n of the hea r ing or at any prehear i ng conference, s u bject to the l i m i ts and condit i ons as may be imposed by the p r esiding off i cer. Details regarding the opport u ni t y to make a lim i ted appea r ance w ill be provided by t he presiding officer i f such sessions are schedu l ed. V. El e ct r onic Submission s (E-Filing) A ll docu m ents f il ed i n NRC ad j udicatory p roceed i ngs, in c l uding a req u est for hea r ing, a petition for leave to intervene , any mot i on or other docu m en t filed in the proceed i ng prior t o the subm i ssion of a request for hearing or petition t o intervene (hereinafter "petition"), and documen t s filed by in t erested governmenta l e nt ities participa t ing u nder 1 0 CFR 2.3 1 5 (c), must be filed in accor d ance with t h e N RC's E-Fi l ing ru l e (72 FR 49 1 39; August 28, 2007, as a m ended 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 dig i tal identification (ID) certificate, which allows the participant (or i ts 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 tha t the participant will be submitting a petition (even in instances in which the participant, or its counsel or representative, already ho l ds an NRG-issued d i gital ID certificate).

Based upon this information, the Secretary will establish an electronic docket for the hearing in th is proceeding if the Secretary has not already established an electronic docket. I nformation about applying for a digital I D certificate is availab l e 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 h tt p://www.n r c.gov/site-h e l p/e-submi tt a l s/a d j u d i catorv-sub

.h tml. 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 E l ectronic Fi l ing He l p Desk will not be able to offer assistance in using unl i sted software.

Once a participant has obtained a digital I D 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 s ite at http://www

.nrc.gov/site-help/electronic

-sub-ref-mat.html. A fi l in g 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 re ce i pt 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 pa rtic ipants (or their counse l or repre sentative) must apply for and receive a dig i tal ID certificate before a hearing petition to intervene i s 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 "Contac t Us" link located on the NRC's publi c Web site at http:llwww

.nrc.gov/site-he/ple-submittals

.html , mail to MSHD.Resource@nrc

.gov , or by a toll-free call at 1-866-672-7640. Th e 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 the ir initial pa pe r filing stating why there is good cause for not filing electronically and reques ting author i zat ion to continue to subm it docume nt s in paper format. Such filings must be sub mitted by: (1) first c la ss mail addressed to the Office of the Secretary of the Comm i ssion, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:

Rulemaking and Adjudications Staff; or (2) courier, express mail , o r expedited delivery service t o the Office of the Secretary, Sixteenth Floor, One White Flint N orth, 11 555 Rockville Pike , Rockville, Maryland , 20852, Attention:

Rulemaking and Adjudications Staff. Participants filing a document in this manner are respons ib le for serving the document on all o ther 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 serv i ce. 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 infor mation 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 Fa i r Use application, part ic ipants are requested not to include copyrighted materials in the ir submission.

The Commission will is s ue a notice or order granting or denying a hearing req ue st 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 heari ng. 11 2016. For further details with respect to this application, see the app l ication dated July 22, Dated at Rockville, Maryland, this day of October 2016. For the Nuclear Regu l atory Commission.

Balwant K. Singal , Senior Project Manager Plant Licensing B ranch IV-1 Div ision of Operating Reactor Licensing Office of Nuclear Reactor Regulation 1 2 From: Sent: To:

Subject:

Greetings, Coll ins, Jay 14 Oct 2016 19:39:58 +0000 A ll ey, David Fw: RE: WCNOC RV pictures It looks l ike you will get access. The numbering is a bit confus in g. On ce you get connected, in th e first folder i s a li st of pictur es. Th e fil e titl ed, Pen 67 & 54 DSC00068, seems to include a picture of penetration nozzle s 67 a nd 5 4. They also appear to be labe l ed in the picture. I believe th is is a view that I saac gave us prev io usly , but not the same photo. Note that th e v e nt lin e in the picture between t he two penetration nozzles. Now, if you go to the h ead m ap imag e l isted as M-706-00009_REACTOR PEN in th e folder, you w i ll find that p e n etra tion n ozzles numb e r s 67 and 54 are no where ne ar nozz le 77 , the source of the spi l l. Instead, number 67 is at approximately 320 degrees and nozzle 54 is a t 290 degrees, nea r the periphery, in the south-w est qu ad r ant of th e h ea d. Not e a l so, th ey are not near the h ead vent line, which i s at a bout the 45 de g ree l ocation in the North-w es t quadrant of the head. I b el ieve nozzle 67 i s the nozzle 76 that I saac c ir c l ed with a question mark in the images he sent on Thu rsday. Either way, 67 or 7 6 , i t has remaining indications in the annulus region and is not li sted for v o lum e tr ic inspection. I will look over the photo s this weekend, but i think we w i ll perhaps need an interna l discussion on Monday for a bit. Jay From: Good N i co l e R Se n t: Frida y, Octob e r 14, 20 1 6 10:25 AM To: Singa l , Balwant; L ingam , Siva Cc: Li ngam, Siva; Co ll in s, Ja y; T sa o, J oh n; A ll ey, D avid; P ascarel li , Rob e rt S ubj ec t: [External_ Sender] RE: WCNOC RV pictures Access has been pro vid ed to: Siva Lingam Jay Collins John Tsao Acces s has been requested for: Balwant Singal Dav i d Alley Robert Pa sca relli Thank you , Nicole Good From: Singal, Ba l want l want.Singal@nrc.gov

Sent: Thur sda y, October 13, 2016 3:12 PM To: Good Nicole R; Lingam, Siva C c: Lingam, Siva; Collins, Jay; Tsao, John; Alley , David; Pascarel li , Robert S ub ject: RE: WCNOC RV picture s Nicole, Not clear if we already have the access or will be getting access late r. Please have access to following two persons as a minimum: Siva Lingam Jay Collins Balwant K. S i nga l Senior Project Manager (D i ablo Canyon and Wolf Creek) Nuclear Regu l atory Comm i ssion Divis i o n of Operating Reactor Licensing Balwant.S i nga l@nrc.gov T el: (301) 415-3016 Fax: (301) 41 5-1 222 F r om: Good Nico l e R [mai l to: ni l yon@WCNOC.com) S e nt: Thursday, October 13, 2016 4: 05 PM To: Lingam, Siva <Siva.Li ngam@nrc.gov

> C c: Singal, Balwant <Ba l wan t.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 I MS 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 pic t ures tomorrow.

Th ank you, Nico l e Good Licen s ing n i l yon@wc n oc.com (620) 364-8831x4557 W o If Cree k..u...*.,,"""* Nuclear Operof i ng Corporat i on From: Sent: To: C c: Subj e ct: lingam , Siva 14 Oct 2016 21:03:06 -0400 Tsao, John;C ollins, Jay Pascare ll i, Robert;Al l ey, David;Singal, Balwant RE: Wolf Creek Relief Requests The licensee wants our approval (obvious l y verbal) b y 10/17 /16 , as d i scussed during the phone call on Thursday.

From: lingam, Siva Sent: Friday, October 14, 2016 9:00 P M T o: Tsao, John <John.Tsao@nrc.gov>;

Collins, Jay <Jay.Collins@nrc.gov>

Cc: Pa scare lli, Robert <Robert.Pa scare lli@nrc.gov

>; Alley , David <David.Al l ey@nrc.gov>;

Singa l , Balwant <Balwant.Singal@nrc.gov>

Subj e ct: FW: Wolf Creek Relief Requests Attached please find the revised RRs from the l icensee for your review/commen t s/eva l uation. Thank you. From: Mu i lenburg William T [ma i lt o:wimuile@WCNOC.com

] Sent: Friday, October 14, 2016 6: 4 2 P M T o: Singal, Ba l want <Balwant.S i nga l@n r c.gov>; 'nick.taylol r@nrc.gov'

<nic k.tay l olr@nrc.gov>; Dod so n , Douglas <Douglas.Dodson@nrc.gov

>; Thomas, Fabian <F abia11.T homas@nrc.gov

>; Lingam, Siva <Siva. Li nga m@nre.gov>

Subject:

[External_

Sender) Wolf Creek Relief Requests Eve r yone, Attached are Wolf Creek Re l ief Requests 1 4R-03 and 04. Bill Muil enburg, 620-364-4186 From: Alley, David Sent: 17 Oct 2016 16:51: 25 +0000 To: T s ao, J ohn

Subject:

FW: Wolf Creek: Verbal Authorizat i on script for Relief Request 14R-03 Volumetric Leak Path Assessment John, These are the wo r ds Jay put i n T he licensee made this req u est in acco r dance w i t h the requi r e m ents o f 1 0 C F R 50.55a(z)(2), such that compl i ance w i th the specified req u iremen t s would res ul t i n hardship or unusual diff i culty without a compe n sating inc r ease in the l eve l of qual i ty and safety. From: Al l ey, David S e nt: Monday, October 17, 2016 11: 58 AM To: Collins, Jay <Jay.Co ll ins@nrc.gov>; Tsao, John <Jo h n.T sao@nrc.gov>;

Singal, Balwant <Ba lwant.Singal@nrc

.gov> Cc: Lingam, Siv a <Siva.Lingam@nrc.gov

>; Pa s care ll i, Robert <Rob e rt.Pascarelli@nrc

.gov> Subj e ct: RE: Wo l f Creek: Verbal Author i zat i on script for Relief Request 1 4 R-03 Volumetric Leak Path Assessment J ohn , Jay, Both ve r bals appea r we ll wr i tten. I do have one question. I n a norma l SE i n the fi r st or second paragraph we say tha t the licensee is proposing its a l ternative in acco r dance wit h 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 t o officially find out that these a r e (z)(2) r eq u ests. Do we norma ll y pu t the a u tho r i t y fo r th e proposed a l ternative up front? D ave From: Collins, Jay Sent: M onday, Octo b er 1 7, 2016 11:24 A M T o: T s ao, John <John.Tsao@nrc.gov

>; S i nga l , Balwant <Ba l want.Si ngal@nrc.gov

> Cc: A ll ey, David <David.Alley@n r c.gov>; Linga m , Siva ; Pascarelli, Robert <Robert.Pascare ll i@nrc.gov> Subj e ct: RE: Wo l f Creek: Verbal Authorization s cript for Reli e f Reque s t 1 4R-03 Volumetri c Leak Path A sse ssment Greetings , 1st D raft of th e scr i pt for 1 4 R-03. Please prov i de m e comments when you can. J ay From: Tsao, John S e nt: Thur s day, Octob e r 13, 2016 2: 09 PM T o: Singal, Ba l want <Balwant.S i nga l@n r c.gov>; Collins, Jay <Jay.Coll i ns@nrc.gov

> C c: A ll ey, David <Dav i d.Alley@nrc.gov>; Lingam, Siva ; Pascare ll i, Robert <Robert.P asca re ll i@nrc.gov>

Subject:

Wolf Creek: Verbal Authorization script for Relief R equest 14 R-04 Alternate CRDM nozzle exam i nations 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 s l ight l y 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 , 14 R-03.

From: Sent: To:

Subject:

Attachments:

Greetings , Collins, Jay 17 Oct 2016 12:01:5 5 +0000 Tsao, John;C umblidge, Stephen;Alley, David Wolf Creek -Potential additional nozzles require volumetric in spection wolfcreekHead2016.pdf One-page attachment withheld in fulll under ex4. Attached is a small s i zed sl i de package of some pictures of the Wolf Creek Head during the Fall 2016 refueling outage. A l eak from above provided significant boric acid depos it s on the head surface, which interfered with the li ce n seea*Žs ability to perform an effective bare metal visual examination. The lice nsee identified 12 nozzles that wou l d require further volumetric examination. Howev er, in reviewing the attached photos , I believe there are additional nozz l es that should be considered , or at least further explanation should be provided by the licensee , if a volumetric examina tion i s not planned to be performed this refueling outage. We are setting up an internal ca ll this morning with the Region to discuss the i ss ue. Jay From: Sent: To:

Subject:

Attachment s: Coll i ns, Jay 17 Oct 201617:57:04

+0000 Cumblidge, Stephen BMV for upper heads BMV for upper heads.p df I think there is a Rev 3 for th i s now .... http://www.epr i.com/ a bstracts/P a ges/Prod uctAbst ract.asp x?Pr o duct ld =000000000001007842.

Visual Examination for Leakage of PWR Reactor Head Penetrations Revision 2 of 1006296, Includes 2002 Inspection Results and MRP Inspection Guidance WARNING: Plea se read the Licen se A g reement on t h e back cover before r emov ing the Wrapping Material. EPf21 Technical Report From: Sent: To: Cc:

Subject:

Leak Path Assessment Attachments:

Greetings, Coll i ns, Jay 17 Oct 2016 15:23:59 +0000 Tsao, John;Singal, Balwant Alley, David;Lingam, Siva;Pascarell i , Robert RE: Wolf Creek: Ve rbal Authorization sc ri pt for Relief Request 1 4R-03 Vo lum etric Wolf Creek verba l auth 14R-03 10-17-20 16.docx 1st D raft of th e script for 1 4 R-03. Pl ease provide me co m men ts w h e n you can. J ay From: Tsao, John Sent: Thursday, October 13, 2016 2:09 PM To: Singa l , Balwant; Collins, Jay Cc: Alley, David; Lingam, Siva ; Pascarelli, Robert

Subject:

Wolf Creek: Verbal Authorization script for Relief Request 1 4R-04 A lt ernate CRD M nozzle exa mination s Attached is the draft verbal authorization scr i pt for Relief Request 14R-04 regarding alternate exa m inations of CROM nozzles at Wolf Creek. I may change the script slightly after the licensee submits t he final version of the relief request. Please review and prov i de changes. The attached verbal authorization doe s not include Rel i ef Reque s t 14R-03 whi c h Jay i s working on. Jay , if you want I can prepare the script for your relief request , 14R-03.

V ERB A L AU TH O RI ZA TI ON B Y THE O FFI CE O F N U CL E A R RE AC T O R RE G U LA TI O N R E LI EF R EQ U ES T 1 4R-03 A L TERNAT I VE TO U SE VOL U METR I C L EA K PA TH FOR SU P P L EME N TA L EXAMS WO LF C RE EK G ENE RA TI NG S T A TI O N WO L F CREEK NU C L EAR OPERA TI NG CORPORA T IO N DOCKE T N UMBE R 50-4 82 T e chnical Evaluation r e ad by David Alley , Chief of the Component Performance , Non-Destructive Examination , and Testing Branch , Office of Nuclear Reactor Regulation By l etter dated Octo b er 1 4 , 2016 , Wolf Creek N uclear Ope r at i ng Corporat i on (t he li censee) s u bmitted Re li e f Reques t 1 4R-03 for t he al t ernate exam i nation of con t ro l r od drive mechan i sm (CRO M) nozzle penetrat i on weld numbe r s 20 , 27, 35 , 40 , 46 , 47 , 58 , 59 , 63 , 70 , 71 and 77 a t the Wo l f Creek Gene r a ti ng Station. T he l i censee p r oposed (a) to perfo rm a volumetric leak path assessment of each penetra ti on nozzle i n li eu o f the surface l eak path assessment requ i red by P a r agraph -3200(b) o f ASME Code Case N-729-1 , and (b) if an u nacceptable i nd i cation by t he l eak path assessment or vo l ume t r i c exa m is ident i fied , the li censee w ill revert to t he r equiremen t s of Code Case N-729-1 and 10 C FR 5 0.55a(g)(6)(i i)(D). T he N RC staff fi n ds that wh il e the demonstrated vo lum etr i c l eak path is not eq u iva l e n t to a full y q u ali fi ed su rf ace lea k pat h assess m en t , th e li censee i d en ti fie d s u ff i cien t ope r at i o n a l exper i e n ce , technica l basis and rad i olog i ca l dose hardsh ip to s h ow t ha t r egulatory co m p l iance wou l d res u lt i n h ardship w i thou t a co m pe n sati n g increase in t he l e v e l of qua li ty and safety. F or ope r at i ng exper i ence , t he licensee s h owe d that t he r e has been n o pr evio u s i dent ifi ed cracking o r lea k age id ent i fied fro m the C RDM nozz l e penetrat i ons or welds of the upper head at Wolf Creek. T he N RC staff noted that while th i s fact does not prec l ude the possibil i ty of c r ac k ing to be found as t h e p l ant continues t o age , p l ants wh i ch have p revio u sly i d en ti fied c r ac king a r e more l ikely t o see subseq u en t a n d m ore s i gn i ficant crack i ng i n th e futu r e. Given the lack of the in i tial cra c king b e ing i dentifi e d i n t h e n ozz l e heats o f ma t e ri a l , a t the ope r a t i ng t empe r a tu res o f Wolf Creek , th e N RC found t ha t t h e poten ti a l f or s i gn i fica nt crack i n g this outage was l ess l ikely. For t ec hn ica l basis , th e li ce n see i d en ti fie d t h at t h e ir in spection wo ul d be in co m p li ance wit h t h e Wes d ye Tech ni cal Ju st i ficat i on Docume nt s h owing an effec ti ve demonstration of the vo l umetric leak path t e c h nique. The N RC has accepted th e use of a demons t rated vo l umetric l eak pa t h as part of t he up p er head i nspection p r ogra m un der 1 0 CFR 50.55a(g)(6)(ii)(D). The licensee also referenced N U REG/CR-7 1 42, Ul trason i c Ph ased Ar r ay Assess m ent o f th e I n t erfe r ence F it and L eak P a t h of the North Anna U n i t 2 Control Rod D r i ve Me c h an i sm Nozz l e 63 w i th D es t r u ctive Vali d at i on , w hi ch fo u nd, i n part, th e use of a proper l y foc u sed 0 degree probe c o u ld detect a l eakage path under l ow leakage rates du r ing ope r ation t hat l ed to minima l wastage of the uppe r head low a ll oy steel. Wh i le the N RC staff di d no t fi nd t hat t he v o lu me tr ic l ea k pa t h assess m ent was equiva l ent to a qual i fied surface leak path a s ses s ment, the i n fo r matio n does demons t rate th e effect i veness of t h e vo l umet r ic l ea k path exa mi nat i on t o d e t ect l ow l eakage r ates, as perfo r me d i n accordance w i th th e li censee's propose d al t e rn at i ve.

For hardship, the l icensee 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 1 O 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 operationa l 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 examinat io n , 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 l evel of qua li ty 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 Bran ch IV-1 , Office of Nuclear Reactor Regulation, I concur with the Component Performance, Non-Destructive Examination , and Test ing Branch's determinations.

The NRC staff concludes that the proposed alternative prov i des reasonable assurance of st ructural integrity of t h e CROM penetration nozzles numbers 20, 27 , 35, 40, 46, 47 , 58, 59, 63, 70, 7 1 and 77 s u ch that comp lyin g with the ASME Code requirement would re su lt in hardsh i p o r unusual difficulty without a compensating i ncrease in the l evel of qua lity and safety. Accordingly , the NRC staff concludes t h a t the li ce n see ha s adequately addressed all of the regulatory requirements set forth in 10 C FR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(ii)(D). Th erefo r e, the NRG staff authorizes the use o f re l ief request 14R-0 3 at the Wolf Creek Generating Station during the current refueling outage subject to the l icensee's proposed alternative that if an una ccep table indication by the le ak path assessment or volumetric exam i s identified, the l icensee will revert t o 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 Authori ze d N uc l ear I n-service In spector. Thi s verbal authorization does not preclude the NRG staff from asking additiona l c l a ri ficat i on quest i ons r egarding Relief Request 14R-03, while p r eparing the s ubsequent written safety evaluation.

From: Sent: To: Cc:

Subject:

Att a chment s: Greetings, Coll i ns, Jay 17 Oct 2016 07:54:18 -0400 Lingam, Siva Singal , Balwant RE: Wolf Creek Relief Requests wolfcreekHead20 1 6.pdf Slides for the internal meeting. Jay Collins From: Collins, Jay Sent: 17 Oct 2016 19:00:26 +0000 To: Barillas, Martha Subje ct: List of Plants with Cold Leg Temperature Upper Heads with indications of PWSCC in penetrat i on nozzles and/or we l ds Att a chm e nt s: Cold Head Cracking.jpg N'ead Te *mD 556.8 ... 57 j.... ...:S 557.0 557.0 56 1 0.Q 558.0 557.0 0 . 557.0 560.0 561.0 5 5 1.0 558.4 559.9 556.0 552.0 550.4 561.0 547.0 -4 ""f 0 :5 I .. 557.3 P l ant Calla.v.1 ay *Stu1m1er !'vfcGtrir 1 e I McC.Jttlr 1 e 2 I Wolf Ci r eek Cata"\\iJa 1 Sl1 e aro11 Cata"\;\rba 2 "\r og t l e 2 Co1na 1 1ch e P e ak 1 B yro1 1 1 Nfil l s to11e 3 Seabrook B raid\.v*ood I B raid\."\'ood 2 B yron2 Co1nancl1e Peak 2 Seqt1oyal1 1 Seqt1o)r ah 2 Watts B ar 1 Materi al s u U ll D I el' H (B) H s H H s (B) H H H H H (B) B (B) H s s s Ve ssel F1 lm catmr -a -CE C B I CE RD1\t1 CE CE RDI\t l CBI CE C E CE B W CE CE B T\J\T B\A T B"V\T CE RDM RD!'i l l RDrvi From: Sent: To:

Subject:

Attachments:

Lingam , Siva 17 Oct 201612:34:35

+0000 Cumblidge, Stephen FW: Wo lf Creek Revised R Rs -Internal D i sc u ss ion p r222_.pdf , wolfcreekHead2016.pdf

---

Or i ginal AppointmcntFrom: Lingam, Siva Se nt: M onday, October 17 , 2016 8:20 AM To: Li n ga m , Siva; Dod so n , Dou glas; Thoma s, F abia n; Proulx , D avid; D rake, Jam es; Wern e r , Greg; Ancho ndo, I saac; Kopriva, Ron; Collins, Jay; T sao, J ohn; Alley, David; Pascarelli, Robert S ubj ec t: W olf C r eek R ev i sed RR s -Internal Discussion When: Monday, October 1 7, 2016 9:00 AM-1 0:00 AM (U T C-05:00) Eastern Time (U S & Canada). Wh e r e: HQ-OWFN-1OB06

-I2p Pl ease note the following to discu ss the s ubject RRs ba se d o n the attac h ed revised RRs and the s lid es. Th e l ice n see wants NRC approval today to sup port the outage. Bridge No.: Pa ssco d e: Date: T im e: 877-935-1422 mJ fo llow ed by # October 1 7, 2016 (Monday) 9:00 AM (Eastern Time)

From: Sent: To: C c: Subj e ct: lingam, Siva 17 Oct 2016 19:14: 00 +0000 Karwoski, Kenneth Pascare ll i, Robert; Peralta, Juan;Singal, Balwant RE: Wolf Creek SG Follow-up FYI/review.

Thank you. From: Singal, Balwant S e nt: Monday, October 17, 2016 3:00 PM To: lingam, Siva <Siva.Lingam@nrc.gov>

Subj ec t: Fwd: Wolf Creek SG Follow-up P l ea s e forward it ken k arwo s ki. F r o m: "Knu s t Jason B" <j a knu st@WCNOC.co m> S u b j ect: [External'-Sender] Wolf Creek SG Follow-up D ate: 1 6 October 20 1 6 15:28 To: "Si n ga l , Ba l want" <Balw a nt.S ingal@nr c.gov> Cc: "Wagner Pat G" <pawagn e@WCNOC.co m>, "Mui l enb u rg W ill iam T" <w i muile@WCNOC.com> Ba l want, We wanted to follow up w i th you regard i ng the 2 c i rcumferentia l crack indications we notified you about last week. In addition, we have been working close l y with the N RC Reg i on I V I S i inspector, Ron Kopriva. One of the indications was in SG D (Row 19, Column 83). This ind i cation was characterized as PWSCC, and confirmed with the rotating Ghent probe. I t was located within a geometric anomaly, 0.77" below the top of the hot leg tubesheet and was sized at 34% T W max depth and 32 degrees c i rcumferentia l extent. Per the guidance i n the EPRI Steam Generator In Situ Pressu r e Test Guidelines , no further testing is required.

The other i ndication was in SG C (Row 41, Co l umn 70). This ind i cation was characterized as ODSCC, and was confirmed with the rotating Ghent probe. The indication is located within the top of the hot l eg tubesheet expansion transition region, 0.23" below the top of the hot leg tubesheet.

T he maximum depth of this flaw was sized at 86% TW with a circumferent i al extent of 112 degrees. We performed in situ pressure testing on this indication today. N o l eakage was detected up to Steam Line Break Pressure and Burst Testing is not required.

Condition Monitoring i s met for this indication.

Eddy cur r ent 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 l eg top of tube s heet program to perform 100% inspections in steam generators B/C/D. With the 100% inspection originally p l anned for SG A during RF21, we have now performed 100% of the hot l eg TIS in all 4 SGs. No additional crack-like indications have been identified.

Thank you, Ja son 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, I saac; Kopriva , Ron;Dodson, Doug l as;Thomas, Fabian;Prou l x, David C c: Pascare ll i, Robert;Al l ey, David;Collins, Jay;Tsao, John;Singa l , Balwa n t Subj e ct: FW: Wolf Creek Re l ief Request s Att a chm e nt s: pr222_.PDF, wolfcreekHead2016.pdf See listing of Records A lr eady Ava il abl e to the Pub li c for these attachments.

Attached please find the s l ides for the internal meeting to be setup in add i tion to the rev i sed RRs from the licensee. From: lingam, Siva S e nt: Monday , October 17, 2016 7:50 AM To: Dodson, Douglas <Doug l as.Dodson@nrc.gov>;

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

Prou l x, David <David.Proulx@nrc.gov>

C c: Taylor, Nick <Nick.Taylor@nrc.gov>;

Burkhardt, Janet <Janet.Burkhardt@nrc.gov>

Subject:

FW: Wolf Creek Relief Requests We will have an internal cal l based on the attached rev i sed RRs submitta l from the l i censee and the pictures placed in the Certrec. I will setup the call and the details wi ll follow. From: lingam, Siva Sent: Friday, October 14, 2016 9:00 P M To: Tsao, John <John.T sao@nrc.gov

>; Collins , Jay <Jay.Collins@

n rc.gov> Cc: Pa scare lli, Robert <Rob ert.Pascarelli@nrc

.gov>; Alley, David <David.Alley@nrc.gov>; Singa l , Balwant <Balwant.Singa l@n r c.gov> Subj e ct: FW: Wolf Creek Relief Requests Attached please find the revised RRs from the l icensee for you r review/commen t s/eva l uatio n. Thank you. From: Mu i lenburg William T [ma i lto:wimui l e@WCNOC.com

] S e nt: Friday , October 14, 2016 6:42 P M To: Singal, Ba l want <Balwant.S i nga l@nrc.gov>; 'nick.taylolr@nrc.gov'

<nick.taylolr@nrc

.gov>; Dod so n , Douglas <Douglas.Dodson@nrc.gov

>; Thomas, Fabian <F abian.T homas@nrc.gov

>; Li ngam, Siva <Siva. Li nga m@nre.gov>

Subject:

[External_

Send er) Wolf Creek Relief Requests Everyone, Attached are Wolf Creek Reli ef Requests 1 4R-03 and 04. Bill Muilenburg, 620-364-4186 From: Sent: To:

Subject:

Singal, Balwant 17 Oct 201615:00:25

-0400 Lingam, Siva Fwd: Wo lf Creek SG Foll o w-up Pl ease forward i t ken kewasji. From: "Knust Jason B" <jakn u st@WCNOC.com> S ubject: [Ex t ernal_ Sender] Wo l f Creek SG Follow-up Date: 1 6 October 2 01 6 15:2 8 To: " Singa l , Balw a nt" <Balwant.S in ga l@nr c.gov> Cc: "Wagner Pat G" <pawa gne@WCNOC.com>, "Muilenburg William T" <w i m uile@WCNOC.com> Bal want, W e wanted to follow up w i th you regar d ing the 2 circumfe r ential crack indication s we notifi ed you about last w ee k. In addition, we have been working closely with the NRC R eg ion IV I S i in specto r, Ron Kopriva. One of the i ndications was in SG D (Row 19, Column 83). This in d ication was c h aracterized as PWSCC , a nd co nfi rmed w ith the rotating Ghent probe. It w as l ocated within a geometric ano maly , 0.77" below the top of the hot leg tubesheet an d was sized at 34% TW max dep th and 32 degrees ci r cumferential extent. Per the guidance in the EPRI Stea m Generator I n Situ Pre ss ure Te st Guideline s, n o f urt h er test i ng i s required. The ot h er ind icat ion was in SG C (Row 41, Column 70). Th is ind ication w as c h arac t erized as ODSCC, and was confirmed with the rotati n g Ghent pr ob*e. The i ndication i s l ocated within the top of the hot leg tubesheet expansion transition region , 0.23" b elow the top of th e hot leg tubesheet. The maximum depth of this flaw was sized at 86% TW with a c i rcumferentia l exte nt o f 112 degrees. We performed in s i tu pres su re testing on thi s ind i cation today. No leakage wa s detected up to Steam Line Break P ressure and Burst Testing is not requ ir e d. Condition Monitoring is met for th is indicat ion. Edd y current in spec tion following the in situ test determined no change to the flaw characteristics.

Ron Kop riv a w i tn essed the in situ test. We w ill stabilize and plug both of these tubes. As prev iou sly com muni cated, we have expanded our hot l eg top of tu besheet program to perform 100% inspections i n s t eam ge n erat or s B/C/D. W i t h the 100% insp ectio n originally planned for SG A during RF21, w e h ave now performed 100% of the hot leg TIS in a ll 4 SGs. No add iti onal crack-like i ndicat i o n s have been ident ifi ed. Th ank you, Jason Knust Licensing Engi neer Wolf Creek Nucl ear Operating Corporat i on (620) 364-8831 x4424 From: Sent: To: C c: Subj e ct: Siva , Taylor , N i ck 1 7 Oct 2016 15:30:32 -0500 Lingam, Siva; Dodson, Dou glas;Thoma s, Fabian; Prou l x, D a vid Burkhardt, Janet RE: Wolf Creek Relief Requests Did this internal call occur today? I have a few questions I observations about the relief request, wi ll hold them for the internal call whenever that occurs. Thanks, Nick Tay l or 972-921-6398 From: Lingam, Siva S e nt: Mond ay, October 17, 2016 6:50 AM To: Dodson, Douglas <Doug l as.Dodson@nrc.gov>;

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

P r ou l x, David <David.Proulx@nrc

.gov> Cc: Taylor, N ick <Ni ck.Ta ylor@nrc.go v>; Burkhardt, Janet <Ja n et.B urkhardt@nrc.gov> Subj e ct: F W: Wolf Cr ee k Relief Requests We will have an internal call based on the attached rev i sed RRs submittal from the l icensee and the pictures placed in the Certrec. I will setup the call and the details will follow. From: Lingam, Siva S e nt: Friday , Octobe r 14, 2016 9:00 P M To: Tsao, John <J ohn.T sa o@nrc.gov>; Collins, Jay <Jay.Collins@n r c.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@n r c.gov>; Alley, David <David.Al l ey@n r c.gov>; S in ga l , Balwant <Balwant.Singa l@n r c.gov>

Subject:

FW: Wolf C r eek Relief Requests Attached please f ind the revised RRs from the l icensee for your review/comments/eva l uatio n. Thank you. From: Mu i l e nbur g Will ia m T [ma i lto:wimu i l e@WCNOC.com) S e nt: Friday, October 14, 2016 6:42 P M To: Singa l , Ba l want <Ba l want.S i nga l@nrc.gov>; 'n ick.taylolr@nrc

.gov' <nic k.taylol r@nrc.g o v>; Dodson , Dou g las <Douglas.Dodson@nrc.gov

>; Thomas, Fabian <Fabian.T homas@nrc.gov

>; Li ngam, S iv a <Siva. Li nga m@nre.gov>

Subject:

[External_

Send er] Wolf Creek Relief Requ es ts Eve r yon e, Attached are Wolf Creek Relief Requests 1 4R-03 and 04. Bil l Mu i l enburg, 620-364-4186 From: Sent: To: C c: Subj e ct: Leak Path A ssess ment Att a chm e nt s: Tsao, J ohn 17 Oct 2016 12:57: 13 -0400 Collins, Jay;Alley, David Lingam , Siva;Pascarelli, Robert;Singal, Ba l want RE: Wolf Creek: Verbal Authorization sc ri pt for Relief Request 14R-03 Vo l umetric Wolf Creek verba l auth 14R-04 10-1 7-2016.docx Per Dave's suggestion , I have revised the verbal authorization script for Relief Request 14R-04 by adding the followin g sentence to the second paragraph in the attached script. "The li censee made t h i s request i n accordance with t he requ ir ements of 1 0 CFR 50.55a(z)(2), such that co m pl i ance w i t h the specified req uir ements woul d res ul t i n hards h ip o r unusual d i ff i culty withou t a co m pe n sating inc r ease in the l eve l of quality and safety ... " From: Co ll ins, Jay Sent: Monday , October 17, 2016 12:50 PM To: Alley, David <David.Alley@nrc.gov>

C c: Lingam, S i v a ; Pa sca r e ll i, Rob e rt <Rob e r t.Pascarelli@nrc.gov>;

Singal, Balwant <Balwant.Singal@nrc.gov

>; T sao, John <John.Tsao@nrc

.gov>

Subject:

RE: Wo l f Creek: Verbal Author i zation script for Relief Request 1 4R-03 Volumetric Leak Path A ssess m e nt Well, since it is a verbal and there are no actual requirements for what we put in the script. ... But t o provide you "the quote" , so you can say it one more time. as you go through my l ong scr i pt, I have inc l ude the attached a new ve r s i on of the scr i pt. (Jus t the one l i ne is added , you guys don't have to l ook at it agai n, t h anks for the feedback btw) From: Alley, David S e nt: M onday, Octob e r 17, 20 1 6 11:58 A M To: Collins, Jay <Jay.Collins@nrc

.gov>; Tsao, John <Jo h n.Tsao@nrc.gov>; Singal, Balwant Cc: Lingam, S i va ; Pa sca r e ll i , Rober t <Robert.Pascare ll i@nrc.gov>

Subject:

RE: Wo l f Cree k: Verbal Aut h o ri zat i on scrip t for Relief Reques t 1 4R-03 Volume t ric Lea k Pat h A ssess ment J ohn, Jay, Both verbals appear well written. I do have one question. I n a norma l SE i n the fi r st or second paragraph we say that the licensee is proposing its a l ternative in accordance with 50.55a(z)(1) (or in this case (z)(2)) or wo r ds to that effect. In the scripts I need to get al l the way to the end to officially find o u t that these are (z)(2) requests.

Do we norma ll y put the authority for the proposed a lt ernative up front? Dave From: Collins, Jay Sent: Monday, Octo b er 17, 2016 1 1:24 A M To: Tsao, John <John.T sao@nrc.gov

>; Singa l , Balwant <Ba l want.Singa l@nrc.gov> Cc: A ll ey, David <David.Alley@nrc

.gov>; Linga m , Siva ; Pascare ll i , Robert <Robe r t.Pascare ll i@nrc.gov>

Subject:

RE: Wo l f Creek: Verbal Authorization script for Relief Request 1 4R-03 Volumetric Leak Path Assessment Greetings , 1 s t Draft of the scr i pt fo r 1 4R-03. Please prov i de m e comments when you can. J a y From: Tsao, Joh n S e nt: Thursday, October 13, 2016 2:09 PM To: Singa l , Ba l want <Ba l want.S i nga l@nrc.gov>; Collins, Jay <Jay.Coll i ns@nrc.gov> Cc: A ll e y , David <Dav i d.Alley@nrc.gov>; Linga m , Siva ; Pascare ll i, Robert <Robert.Pascarell i@nrc.gov>

Subject:

Wo lf Creek: Verbal Aut h orizat i on sc ri pt for Relie f Request 1 4R-04 A l ternate CRD M nozzle exam i nations Attached is t he draft verba l authorizat i on script for Re l i ef Request 14R-04 regard i ng a lt ernate examinations o f CROM nozzles at Wolf Creek. I may change the script s l ight l y afte r the licensee sub m its the fi nal ve r sion of t he r elief r eques t. Please review and provide changes. T he attached verba l authorizat i on does not i n c l ude Re l ief Request 14R-03 w hich Jay i s wo rk i n g on. J ay, if yo u wan t I can prepa r e the scr i p t for your re li e f reques t , 1 4R-03.

V ERB A L AU TH O RI ZA TI ON B Y THE O FFI CE O F N U CL E A R RE AC T O R RE G U LA TI O N R E LI EF R EQ U ES T 1 4R-04 A LT ER N ATE EXAM IN AT I ON OF CO NT RO L RO D D R I VE M EC H AN I S M N OZZ LE PENET RA TI O N S WO L F CREEK GENERA TIN G STA TI O N WOL F CREEK NU CLEA R OPE R A TI NG COR P O R A T IO N DOCK ET N U MB ER 50-4 82 Technic a l Evaluation read by David Alley , Chief of the Component Performance , Non-Destructive Examination , and Testing Branch , Office of Nuclear Reactor Regulation By l etter dated Octobe r 1 1, 2016, Wolf Creek Nuclea r Ope r at i ng Co rp ora t ion (t he li censee) s u bm i tted Re l ie f Reques t 1 4R-04 for the al t ernate examination of con tr o l rod drive m echan i sm (C ROM) nozzle penetra t i on n u m bers 77 and 78 at t he Wo lf Creek Generating Sta t ion. T he l i censee p r oposed (a) an a lt ernate exam i nat ion d i stance for CR OM nozz l e numbers 77 and 78 in li e u of the re q ui r ed exam i nat i o n d i stance per ASME Code Case N-729-1 as c on d itioned by 1 0 CFR 50.55a(g)(6)(ii)(D) and (b) no t to pe rf o r m the surface exa mi nat i on o f the po rti on of the CR OM nozz l e be l ow the J-groove weld as requ i red by 1 0 C F R 50.55a(g)(6)(ii)(D)(3). Th e l ice n see made th i s r e q uest i n accordance w it h the r eq u i r emen t s of 1 0 CFR 50.55a(z)(2), s u c h that co m pl i ance wi th the specified requi r ements wou l d re s ul t in hardsh ip o r un u s u a l difficu l ty w i thout a co m pensa ti ng in c r ease in t h e level o f qua l i t y and safe t y. T he N RC s taff fi n ds that the propo s ed exam i nat i on di s tance are a c ceptab l e fo r CRO M nozzle n u mb e r s 77 an d 78. T his is based o n the val id i t y of t he lice n see's stress analys i s and fr acture mec h an i cs calcu l at i on , de m ons t rating tha t wit h in four refueling cyc l es, a pote n t i al flaw t h at in i t i ates in the unexamined zone (below the J-groove we l d) of t he C ROM n ozz l e n u m bers 77 and 78 w ill no t p r opagate i n to th e J-groove we l d. A t the e n d of every fo u rth ref u e li ng cycle, t he l ice n see w ill perform an exam i nat i on to confirm t he structura l i ntegrity of CR O M nozz l es 77 and 78. Th e NR C sta ff finds t h e li ce n see's h a rd s hip ju stifiica ti on i s acce p ta bl e b eca u se o f th e co n siderab l e ra di at i o n dose a n d th e nozz l e co n figurat i o n th at are no t co n duc i ve fo r th e r eq uir ed exa m ina ti o n. The N RC s taff finds that the li censee's p r opo s ed a l ternative examination d i stan c e s for CR O M penetra ti on nozz l e nu m be r s 77 a nd 78 prov i des r easo n ab l e assu r ance of structura l integrity and leak tig h tness un til t he next schedu l ed examination , and that comp li ance with the s u rface exam i nat i on requ i remen t s of 1 0 CFR 50.55a(g)(6)(ii)(D)(3) wo ul d res u lt in har d s hip w ith out a co m pensat in g in crease in the l evel o f qual i ty and safe t y. Authorization read by Robert Pascarelli , Chief of the Plant Licensing Branch IV-1 , Offi ce of Nucle a r Reactor Regul a tion As Chief of the Plant Licensing Branch IV-1, Office of Nuclear Reactor Regulation , I conc ur with the Component Performance, Non-Destructive Examination , and Testing Branch's determinations.

The NRC staff conc lud es that the proposed alternative provides reasonable assurance of structural integrity of t he 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 incr ease in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regu latory requirements set forth in 10 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(i i)(D). Therefore , the NRC s taff authorizes th e use of rel ief reque st 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*a rty review by the Authorized Nucle a r In-service In spec t o r. This verbal authorization does not preclude the NRC staff from asking add i tional clar ifi cat ion questions regarding Relief Request 14R-04 , while preparing the subsequent written safety evaluation.

From: Sent: To:

Subject:

Tsao, J ohn 17 Oct 2016 12:29:30 +0000 Lingam , Siva Accepted:

Wo l f Creek Revised R R s -I nternal Discuss i on From: Alley, David Sent: 19 Oct 2016 01:11: 44 +0000 To: Rezai, Ali Subje ct: RE: RAI for WolfCreek-1 1 3R-13 MF8308 Coverage Att a chment s: WolfCreek-1 1 3R-13 RAI MF8308 Coverage PP DA.docx, Wo l fCreek-113R

-13 SE MF8308 Coverage Draft with HOLES DA.docx Ali A few thoughts on both the SE and the RAls D ave From: Rezai, Ali Sent: M onday, Octo b er 1 7, 20 1 6 10:48 A M To: Alley, David <David.Alley@nrc.gov

> Subj e ct: RAI fo r Wo l fCreek-113R

-13 MF8308 Coverage H i Dave , A t tached is R A I for your rev i ew/concu rr ence. D ue R A I: 11-2 7 Have a great day. Rega r ds, ali REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST 13R-13 REGARDING WELD EXAMINATION COVERAGE WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NUMBER 50-482 By letter dated August 23, 2016 (Accession Number ML 16243A039), Wolf Creek Nu clear Operating Corporat i on (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vesse l Code (ASME Code) speci fi cally related to ASME Code Case N-460 "Alternative Examination Coverage for Class 1 and Class 2 Welds ,Section XI , Division 1." Relief request 13R-13 pertains to the examination coverage of the Class 1 and 2 piping welds at the Wolf Creek Generating Stat i on (Wolf Creek). Co mm e nt IADI: I normally view MRP: 146 as an augmented inspec t i on f program. In that case it appears that ! the RAJ is unnecessary. Given the ! 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 f he gets back (vacation right now I ! think) and ask the PM to confirm that i the plant considers MRP-146 to be an / augmented inspection program , I.e., f these welds are not MRP 146 welds. ! Given t h at the request c l early states f that they are not in augmented To comp lete its review , the NRC staff requests the fol l owing additional information.

f programs and MRP-146 s hould be : cons ider ed an augmented program , I 1. ________ j

" 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 Materia l s Reliability Program (MRP)-146 " Management of Thermal Fatigue in Normally Stagnant Non-Jsolab le Reactor Coolant System Branch Lines" and/or the Electric Power Research Institute (EPRI) interim guidance MRP 2015-025 " EPR l-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 EPR I interim guidance MRP 2015-025.

b. Given the susceptibili ty 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 th at 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 interroga te 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; should be obvious. b. Prov i de 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 1 3R-13 REGARDING WELD EXAMINATION COVERAGE WOLF CREEK NUCLEAR OPERAT IN G CORPORATION WOLF CREEK GENERA TING STAT ION DOCKET NUMBER 50-482

1.0 INTRODUCTION

By l etter dated August 23 , 2016 (Accession Number ML 16243A039), 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 Eng i neers Boiler and Pressure Vesse l Code (ASME Code) specifica ll y related to ASME Code Case N-460 " A l ternative Examination Coverage for Class 1 and Class 2 Welds ,Section XI , Division 1." This rel ie f request, 13R-13 , pertains to the examination coverage of the Class 1 and 2 piping welds at the Wolf Creek Generat i ng 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 requirement s (if necessary), for in se rvice 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), lnservic e Inspection Standards Requirement for Operating Plants , throughout the service life of a boiling or pressurized water-cooled nuclear power facility , components (inc luding 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 ed i tions 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 prac ti cal within the limitations of design , geometry , and materials of construction o f the components.

Pursuant to 10 CFR 50.55a(g)(4)(ii), Applicable ISi Code: Successive 120-month Intervals , inservice examination o f 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 pa r agraph (a) of 50.55a 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases li sted i n NRC Regulatory Guide 1.14 7, Revis io n 17 , when using Section XI, that are incorporated by re ference in paragraphs (a)(3)(i i) of 50.55a), subject to the conditions listed in paragraph (b) of 50.55a. However, a licensee whose inservice inspection interv al commences during the 12 through 18-month period after July 21, 2011, may delay the update o f their Appendix V III program by up to 18 months afte r July 21 , 2011. Pursuant to 10 CFR 50.55a(g)(5)(iii), IS/ Program Upd a t e: Notification of Impra c tic a l /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 limit ations experienced when attempting to comply with the Code requirements during the inservice inspect io n inte rval for which the reque s t is being s ubmitted.

Requests for rel ief made in accordance w i th 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 evaluat i on, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the l icensee. 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 C l ass 2 piping welds (Examination Category C-F-1 and C-F-2) in the third 10-year I Si interval of Wolf Creek. The licensee developed the ISi program in accordance with the NRC approved methodology of the Electric Power Research I nstitute (EPRI) Topical Report (TR)-112657, Revision B-A, " Revised Risk-Informed l nservice Inspection Evaluation Procedure" (Accession Number ML0134 70102). 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 therma l 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 sta i n l ess 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 r elief 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).

I n 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 I Si 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. I n 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 lim its access to valve side of the weld, and a fitting that limits access to flange side of the we ld , 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 VI II of Section XI. HO L E In the xxxxxxxxx l etter, the licensee provided the percentage of coverage for the "Best Effort" examinations.

The licensee stated that for the welds i n this r e l ief request with sing l e sided access, it extended the beam path into the far s i de of t h e we l d ce n te rlin e t o examine to the extent practical the other side of we l d as a " Best Effort" examination.

However, no credit was c l a im ed for the "Best Effort" exam i nation because a UT procedure must be qua l ified with flaws on the in accessib l e side of the weld. Cur r ently , there are no qualified s in gle-side examination procedures and the ex i st in g UT technology i s not capable of re li ably detect ing o sizing flaws on the fa r side of an austenitic weld. No unacceptab l e indications were identified. 1 Co mm e n t (A D(: This paragraph is (r he l icensee state d tha t the re w ere 1 5 o th er austeni ti c stainless stee l pipe we l ds wi t h t he / potent i ally the most i mportant part of degradation mechanism of thermal fatigue that were examined during the th i rd 10-year I S i / their submission.

P l ease inc l ude this i nt e rv al. The operating condi t ions and environ m ent would be sim il a r to the welds in t his re l ie f / conceo1 in the NRC eva l u ation. request. N o i ndicat i on of degradation due to the r mal fatigue , or a n y other mechan i sm, has been / identified dur i ng the exa mi nat i ons l_ __________________________________

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________________________________

_/ HOLE T he l i censee stated that we l ds included in this request are not part of any augmented i nspect i on programs inc l uding Mate ri als Reliabi li ty Program (MRP)-146

" Management of [T hermal F a t igue in N ormally Stagnant N o n-lso l able Reactor Coo l ant System Branch Li nes," a nd/or the Electric Power Re s earch I nstitute (EPRI) interim guidance MRP 20 1 5-025 " EPR lMRP I n te r im Guida n ce for M anagement of Thermal Fat i gue" (Accession N umber ML 15189A100). H OLE In the xxxx, letter, t he l i censee provided the cumu l ative fatigue usage (CFU) fact Q!J calcu l ated fo r each w e ld ed l ocation i n this re li ef reques t. The licensee stated that the welds in this relief request have been sub j ected to system leakage testing and no sign of leakage has been i dentif i ed. 3.8 Proposed A l ternative In Tab l e 3 of Attac h ment 1 to this re li ef request, the licensee reported the percent coverage achieved for each weld exami n ed. This is summarized below. C l ass 1 welds: BB-02-F019 BB-02-FW30 1 BG-21-F013B EJ-04-F048A C l ass 2 w elds: EP-01-MW7152 EP-02-MW7162 EP-01-MW7 1 65 50 percent 50 percent 50 percent 50 percent 50 percent 50 perce n t 50 percent The licensee proposed t h e above a l ternat i ve coverage in lieu o f the ASME Code required essentially 100 percent coverage. 3.9 NRC Staff Evaluat i on 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 j ustification exists to support t h e determ i nation that the ASME Code requirement is impractical

(2) that i mpos i tion of the Code required inspections wou l d result in a burden to the licensee; and (3) that the licensee's proposed alterna t ive (accepting the reduced inspection coverage in th i s case) prov i des reasonable assurance of structural i ntegrity and l eak t i ghtness of the subject weld. The NRC staff finds that if these three criteria are met t h at the requirements of 10 CFR 50.55a(g)(6)(i), (i.e., granting the requested relief will not "e ndanger 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 re su lt if the requirements were imposed on t he facility") wi ll a l so 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 (sing l e-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 ac hieving 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 sign ific ant design modification to the welds and their associated components.

The NRC staff finds that replacing or reconfigur i ng the components of the subject welds i s the only reasonable means to achieve dual s ided coverage of these we l ds and that rep l acement or recon fi guration of the pipe , va l ve , fitting, and flange constitutes a burden on the licensee. Structural integrity and l eak tightness The NRC staff considered whether the licensee's proposed alternative provided reasonable assurance of structura l integrity and leak tightness of the subject weld based on: (1) the examination coverage achieved and (2) safety significance of unexamined vo l umes -unachievable coverage (e.g., the presence or absence of known ac t ive degradation mechanisms and essentially 100 coverage achieved for simi l ar welds in similar environments subject to simi l ar degradation mechanisms).

Examination Coverage Achieved I n evaluating the lic e n see's proposed alternative, the NRC staff assessed whether it appeared that the li censee obtained as much coverage as reasonably possible and the manner in which the licensee reported the coverage ach ieved. From review of s ubmitt al and th e sketches in Attachment 1 to 13R-13 , the NRC staff confirms that:

• The welds were examined using the appropriate equ ipm ent, u l trasonic modes of propagation, probe angles, frequencies , and scanning directions to obtain maximum coverage;

• The coverage was calcu l ated in a reasonable manner;

• Th e UT p r ocedu r es used were qua lifi ed as requi r ed 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 iden tified. 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 Unexam ined 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 submitta l and the sketches in Attachment 1 t o 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 ba se material near the ID surface of the joint) that are typically susceptible to higher stresses and , therefore , potential degradation.

• HOLE For the stainless stee l welds, the NRG staff notes that the coverage obtained was limited to the vo l ume up to the weld centerline (near-side), because claiming coverage for the vo l ume 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 qua li fication attempts to date. The far-side volume was inspec ted by the " Best Effort" examination, no ind ications were identified , and no credit was taken for the coverage achieved from the " Best Effort" examination.

• HOLE A t each of the welded l ocation in th i s rel ief request, the licensee's calculated cumulative fatigue usage (CFU) factor does not exceed the l i mit of Section Ill of the SME Code. The CFU factor was dete r mined based on the actua l p l ant operating cycles. Therefore , this provide reasonable assurance that the potential for initiation and growth of fatigue cracks is low and of l east concern at these welds and the i r 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 i s rea sonable to co nclude that if s ignificant serv ice 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, the se welds have received the requir ed 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 redu ced coverage of the r equired examination volume , the NRC staff finds that th is inspection will provide additional assurance that any pattern of degradation, if it were to occur , would be detected and the licensee will take appropriate correc tion actions. Therefore , the NRG staff finds that the volumetric examinations performed to the extent possible provide a reasonable assurance of st ru c tural integrity and leak tightness of th e subjec t welds. Compliance with the ASME Code requirements for these welds would be a burden on the l icensee.

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 l aw and will not endanger life or property or the common defense and security, and is otherwise in the public interest giv i ng 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 app l icable, including the th i rd party review by the Authorized Nuclear In service Inspector.

From: Sent: To: C c: Subj e ct: Greetings , Coll i ns, Jay 18 Oct 2016 07:17:35 -0400 Taylor, Nick;Lingam, Siva Tsao, John;Alley, Dav i d;Cumb l idge, Stephen RE: Call with Wolf Creek regard i ng head i nspection 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 S e nt: M onday, October 17, 2016 5:55 PM T o: Drak e, James <Jame s.Drak e@nrc.gov>; Pa sca relli, Rob e rt <Rob er t.Pascarelli@nrc.gov

>; Alley, D avid <David.Alley@nrc.gov>;

Vegel, Anton <Anton.Vege l@nrc.gov>;

Clark, Jeff <J eff.C l ark@nrc.gov>;

Pruett , Troy <Troy.Pruett@nrc

.gov>; L antz, Ryan <Ryan.Lantz@nrc.gov

> Cc: T sao, John <John.T sao@nrc.go v>; Collins, Jay <Ja y.Collin s@nrc.gov>; Dod so n, Dougla s <Douglas.Dodson@nrc.gov>;

T homas, Fabian <Fabian.Thomas@nrc.gov>;

Anchondo, Isaac <l saac.Anchondo@nrc.gov>;

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

Werner, Greg <Greg.W e rn er@nr c.go v>; 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 rel i ef request this morning. Has that call a l ready occurred?

I left a message with Siva Lingam (who is standing in for Balwant) to that affect as welt. .. Thanks, Nick Tay l or Chief, Projects Branch B 972-921-6398 From: Drake, James S e nt: Mond a y, October 17, 2016 4:51 P M To: P asca relli, Rob er t <R obert.Pascarelli@

n rc.gov>; A ll e y , David <Dav i d.A ll ey@nrc.gov

>; V ege l , Anton <Anton.Vegel@nrc

.gov>; C l ark, J eff <Jeff.Clark@nrc.gov>; Pru e tt, Troy ; L antz, Ryan <Ryan.L antz@nrc.gov > C c: Tsao, Jo h n <J ohn.Tsao@n r c.gov>; Co llin s, Jay <Jay.Co ll i ns@nrc.gov>; Taylor , Ni ck <N i ck.T ay l or@nrc.gov>; Dodson, Doug l as <Doug l as.Dod s on@nrc.gov>; Thoma s, 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 indicat ions seen in the pictures provided that were not discussed i n 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 ind icat ion s they had that are not addressed in the relief reque st. 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 ton ight or office phone tomorrow.

Jim f:. :Dra{e Jame s F. Drake Office phone: 817-200-1558 Cell Phone: l<b)(6) I From: Sent: To: Holston, W i lliam 18 Oct 2016 12:52:5 1 -0400 byka@asme.org;Gary Park;powersl@asme.org Cc: Mano l y, Kamal;Benson, Michae l;Holston, W i lliam;ASME Code Day Attendees;ASME Code Par t icipants Subj e ct: NRC ASME August 2016 Report Att a chm e nt s: 2016-11--NRC Report to ASME -St Lou i s MO.docx I have attached the November staff report to ASME. Bill Holston 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 NRG Report to ASME Nov ember 2016 NRC Report for ASME Code Me e tings Nov e mb e r 2016-St Lo u i s, MO Table of Contents Amendment to 10 CFR 50.55a -ASME Code Edit ion/A ddenda ASME Code Case Rulemaking/Regulatory Guides Operating Plant Issues a n d Material Degradat i on NRO DC I P Quality Assurance and Vendor Inspection Branch Activities New Reactor Licensing Activities Multinationa l Design Eva l uation Program (MDEP) Activities 10 CFR Part 21 Rulemaking Commercia l Calibration Services Status NRC Staff Review of EPR I 1025243 Guide l i ne for Commercial-Grade Design and Analysis Computer Programs NRC Staff Review of EPRI Guideline for Dedication of Commercia l-Grade Items for U se in Nuclear Safety-Related Appl i catio ns NRC Staff I nterface with Nuclear Utilit ies Procurement I ssues Committee (NUP I C) Rev e r se Engineering I nformation Notice 2016-01 License Renewal Activities New Generic Lette r s New I nformat ion Notices New Regulatory I ssue Summaries NRC Pub l ications of Potential I nterest to ASME Upcom i ng Public Meetings of Potentia l Interest to ASME 2 2 4 8 10 13 15 15 16 16 17 17 17 20 20 20 20 21

1. Amendment to 10 CFR 50.55a -ASME Code Edition/Addenda Current ASME Edition/Addenda The NRG has approved: NRG Report to ASME Nov e mber 2016

• 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 ).

• Th e Operation and Maintenance of Nuclear Power Plants (OM Code) through th e 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 20 10 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 wi ll 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 cond i tions. NRG Staff has addressed the public comments received a nd is in the process of preparing the fina l rule for pub l ication. T he final rule is expected to be published by D ecember 2016. I n preparation for the next ru l emaking action, a review of the 2015 Edition of the ASME BPV Code and OM Code for incorporation by reference in 1 0CFR50.55a has been completed.

The scope of the rulemak i ng action along with draft conditions a r e current l y under development with a proposed ru l e pub l ication following shortly after the 2009-2013 rule becomes fina l. The NRC discussed its p reli minary positions on the 2015 Editions dur ing a public meeting held on August 22 , 20 1 6. 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 in corpo r ates by reference the Regu l ato r y Guides (RGs) li sted be l ow: Supplements Addressed:

S u pplements 1 thro u gh 10 to the 2007 Edition Effective date for the RGs: December 5 , 2014

• RG 1.84, Revision 36 , "Des i gn, Fabrication, and Materials Code Case Acceptability , ASME Section Ill" (ADAMS Accession No. ML 13339A515).

• RG 1.147 , Revision 17, "lnservice In spection Code Case Acceptabi l ity, ASME Section XI, Division 1" (ADAMS Accession No. ML 13339A689).

NRG Report to ASME Nov ember 2016

• RG 1.192 , Revision 1 , " Operation and Ma inte nance Code Case Acceptability , ASME OM Code" (ADAMS Accession No. ML 13340A034

). 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 pub l i s hed in the Federal Reg i ster 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 Rulemak ing Code Case Supplements Addressed: Supplement 11 to the 2007 Edit io n 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 rulemak i ng package at the request of ASME that are not listed in aforementioned Supplements

• N-694-2 (Supp. 1 to the 2013 Edition) " Evaluation Procedur e and Acceptance Criteria for PWR Reactor Vessel Head Penetration Nozzles"Section X I

• 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 Bol ts and Studs"Section XI OM Code_Code Cases Addressed:

• Draft Revision 2 of RG 1.192 Status of Next RG Publication 2009 Edition through 2012 Edition 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 Ed i tion 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 tho se pub li shed by ASME , that are referenced in Regulatory Guides , in Inspect io n Manuals and Procedures, and in the LWR Standard Review Plan. The lists may be found at this website: http://www

.nrc.gov/abou t-nrc/regu l atory/standar d s-dev/consensus

.htm l.OM Code

3. Operating Plant Issues and Material Degradation Dissimilar Metal Butt Welds NRG Report to ASME Nov e mber 2016 CALVERT CLIFFS UNIT 1 (LER No. 3172016002 , ADAMS Accession No. ML 16106A304)

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 MSIP needs 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 transit i on 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 c oolant 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 i ntervals 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 i n accordance with 50.55a. Operational Leakage SUSQUEHANNA UNIT 1 (LER 3872016020ROO, ADAMS Accession No. ML 16216A378)

-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, Un i t 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 Notificat io n 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 t h ird] of the pipe circumference.

At the time of discovery, the un i t was i n Lowered Inventory with both Loops of Decay Heat in se rvice. Subsequently, one train of Decay Heat was secured to reduce the likel i hood of crack propagat io n. 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 w it h 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 i n MODE 1 at 100 percent rated thermal power and p lacing Excess Letdown in service for Reactor Coolant System (RCS) le ak detection, RCS operational leakage exceeded 1 gpm [gallon per minute] unidentified leakage as identified by performing RCS Water Inventory Balance u s ing the Nuclear Plant Informat ion System Computer. This NRG Report to ASME Nov e mber 2016 required the Unit to be placed into Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 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 l eak over a mirrored insulation panel. After removing the pane l 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 the r mocouple nozzle threaded connection. The licensee also determined that this was not pressure boundary leakage. I n addit i on, the l icensee identified that excess letdown made the l eak 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 p l anned refue l ing 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. ML 16201A232)

MAIN STEAM LINE FLEXIBLE HO SE INTERGRANULAR STRESS CORROS ION CRACKING IDENTIFIED DURING REFUELING OUTAGE While the p l ant 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 i s 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. F l exible hose 1 B21-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.

T he vacuum port protective jacket was in place. F lexible hose 1 B21-D372E

-located at the inner elbow on MSL 'C' had water dripping out slow l y, 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 s i de, and dripping off the bottom. A failure analysis of the flexible hose fai lur es identified the failure mechan i sm as lntergranular Stress Corrosion Cracking (IGSCC). Both leak i ng flexible hoses 1 B21-D372C and 1 B21-ID372E were replaced during the refueling outage and their re s pective high side flexible connections were also rep l a c ed. No additional leaks were found during an inspection of other flexible hoses connected to MSLs and the reactor recircu l ation system. An examination of monitored drywell points prior to plant shutdown for C1 R1 1 6 showed no change in temperature , pressure or airborne radiation levels. IGSCC resulted in the failed flexible hose discovered during the C1 R16 wa l kdown. The root cause evaluation for th i s event determined that the corrective actions to prevent recurrence of the condition identified June 18, 2007 (LER 2007-003) failed to eliminate or significant l y NRG Report to ASME Nov e mber 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. ML 16278A007)

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 rema i ned 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 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. ML 16271A408)

-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 Rep o rt to ASME N o v ember 2 01 6 i so l ated a n d was considered R CS pressure bo u ndary leakage. Th is was eva l uated as R CS pres s ure bounda r y leakage and North Anna Unit 2 entered TS 3.4.1 3.B (RCS Ope r at i onal Leak age) and commenced a s hutd own to mode 5. While in Mode 5, the con t rolled b l eed-off pip i ng associated w i th the RCP seal for 2-RC-P-1 C was replaced. The di r ect cause of the RCS u nid entified leakage was determined to be a la r ge mean stress placed on th e socket we ld due to the co ntr o ll e d b l eed-off lin e not being prope rl y aligned in the downstream pipe support, and therefore not allowing for the the rm al growth of t h e RCS. As a resu l t o f the l arge mean stress , a c r ack initi ated at a small defect (l ack of fusion) in the toe of the socket weld and propagated through the weld d u e to normal cyc li c v i bration from the Reactor Coolant Pump. 50.55a RULEMAKING NR C staff plans to incorporate by reference N onmandatory Appendix U , " Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Piping and Class 2 and 3 Vessels and Tanks ," of Sect i on XI with condit i ons in the Final Ru l e to incorpo r ate the 2009-20 1 3 Editions and Addenda of ASM E B&PV Code. ASME se n t the NRC a l etter describing the conclusion of their activities fo r add re ssing ope r atio n a l le a k age of pressure retain i ng components (A D AMS Access i on No. ML 1 5099A624 ). I n view of ASME's Pressure Bound a ry Leakage project team's c onc l u s ions , t h e NR.C sent a l ette r ba c k to ASME on July 14 , 2015. (ADA MS Accession No. ML 1 5188A057). 4. NRO DCIP Quality Assurance and Vendor Inspection Branch Activ i ties NRO Vendor I nspection The NRO vendor in spection prog ra m i s desc r ib ed in I n spect i on Manual Chapte r (IMC) 2507, " Vendor I nspections

." This I MC was la s t updated on October 3, 2013. This IMC is imp l emented by various In spection Procedures (IP s) inc l uding: IP 43002: Routine In spect i ons of Nu clea r Vendors; IP 43003: React i ve In spec ti ons of Nu clear Vendors; IP 43004: Inspection of Comme r cia l-Grade Dedicat i on Pr ograms; IP 43005: NR C Oversight of Th ird Party Organ i zat i ons Im p l eme n ti n g Quality Assurance Requirements

IP 36100

Inspection of 10 CFR Part 21, Progr a ms for Reporting Defe c ts a nd N oncompliance

IP 37805

Engineering Design Verification In spections; IM C 0617: Vendo r and Q u a li ty Assurance Impl ementat i on Inspection Reports; and IM C 2507: Vendor I nspections FY 16 Vendor Inspection Plans

• AP1000 modu l ar cons tr uction (structural and mechanical)

• AP1000 mecha n ica l and e l ectrical qua li ficat i on test prog r ams

• Digita l I ns tru men t ation and Co n tro l f or AP1000

• Valve and pump manufactur in g

• 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, Pinev i lle, LA -issued

• SPX, Copes-Vu l can , McKean, PA -issued

• Namco Controls Corporation , Elizabethtown , NC -issued

• Aecon Industrial, Cambridge, Canada -is s ued

• Westinghouse Electric Company , Warrendale, PA-completed

• Paxton & Vierling Steel Company, Carter Lake , IA , completed

• Mang iaro tti S.p.A, Monfalcone, Italy -completed NRG Report to ASME Nov ember 2016

• Westinghouse Electric Company , Cranberry Township, PA -completed

• Target Rock , Huntsvill e, AL -planned

• Enercon, Oklahoma City , OK -planned

• Curtiss-Wright EMO , Chesw i ck, PA-planned

• Creusot Forge , Le-Creusot , Fran ce -planned

• PElco, Memphis, TN -planned

• GE Hitachi , Wilmington , NC -planned

• WECTEC, Charlotte, NC -planned Previously issued NRC inspect ion and trip reports are located at: http://www.nrc.gov/reactors/new

-reacto rs/ove rs i ght/quality

-assurance/vendor

-insp/i nsprep o rts.h tm l New Vendor Inspection Quality Assurance Website Links The NRC has implemented website pages to make it easier to become familiar w i th and follow vendor inspection and QA related activities:

http://nrcweb.nrc.gov

400/reacto rs/ne w-reacto rs/ove r siqht/qua l ity-assu rance/v ei nsp.htm l. 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 repres e ntatives from 1 1 1 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 be l ow l ink.

NRG Report to ASME Nov e mber 2016 http://www. n re.gov /reactors/new-reactors/overs i ght/qua I ity-ass u ran 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 FAQ s, and Enforcement FAQ s. T he page provides quick link s to questions we have received in the past about the mentioned topics: http://w ww.n rc.gov/reac tors/new-reacto rs/ove rs i ght/quality-assurance/gua l-assu refags.h t m l 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 Inspe ction reports that have either a Notice of Non confo rmance (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 in spection report th at is released:

http://www. n re.gov /r eactors/new

-reactors/overs i ght/qua I ity-ass u ran ce/non conformancesviolation

s. htm I The web page link below describes the vendor inspe ction program (VIP). The VIP verifies that reactor applican ts and licensees are fu lfilling their regulatory obligations with respect to providing effective ove r s ight of the supp ly cha in. It is accomplished through a number of activities, including:

performing vendor inspec tions that will ver i fy the effective implementation of the vendor's quality assurance program , establishing a strategy for vendor identification and selection criteria , and; ensu ring vendor in specto rs obtain necessary knowledge and skills to perform inspections.

I n addition, the VIP addresses interactions with nuclear consensus standards organizations , industry and external stakeholders, and i nternational consti tuent s: htt p://www.n r c.gov/docs/ML 16 0 7 /M L 16075A461.pdf

5. New Reactor Licensing Activities As of January 28, 2016, the status of new reactor li ce nsing 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 pre ss urized-water reactor design from France) is suspended at the request of the app l icant in its l ette r dated February 25, 2015, until further notice. The NRC staffs review of the Mitsubish i 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 app li cant except for a few key areas. Th e NRC staff comp l eted its review of General Electric-Hitachi's ESBWR (first passive BWR) and issued its final safety evaluation report (FS ER) in March 2011. On M a rch 24, 2011, the NR C issued in the Federal R e gister a proposed ru le (76 FR 16549) for public NRG Report to ASME Nov e mber 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 standa rd p l ant design to the NRC on September 30 , 2013. The NRG staff cond uct ed an acceptance review of the application for completeness , technical adequacy, and acceptability for docketing.

In a lette r to KHNP dated December 19, 20 1 3, 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 b ecause the application was not ready in seve ral key areas. The NRC staff continued pre-application interactions with KHNP to support preparation of a comp l ete application by D ecember 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 appl ic at i on for docketing in its l etter dated March 4, 2015 , based on its determination that the application is sufficiently complete and technically adequate to allow the NRC staff to co nduct it s detailed technical review. In addition , the NRC staff is reviewing two applications for design certification renewal:

• ABWR GE-Hitachi (applicat i on submitted on December 7 , 2010)

• ABWR GE-Toshiba (Revision 1 to app li ca tio n subm itt ed on June 22 , 2012) Earlv Site Permits (ESPs) NRC has issued four ESPs (Clinton , Grand Gulf , North Anna, and Vogtle). Th e 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-w ork authorization (LWA). The NRC received an application for an ESP for the Victoria County Station submitted by Exelon on March 25, 20 10. The site is located in Victoria County, T exas , with no specific technology selected. On August 28 , 2012 , Exelon requested withdrawal of the Victoria County Stat i on ESP application from the docket. By lette r dated October 3, 2012 , NRC accepted the applicant's requ est, and the app li cation was withdrawn. Th e NRG received an ESP application for the PSEG s ite in N ew 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. Th e NRC recen tly issued the final Environmental Impact Statement (EIS) for the ESP. The NRG is currently preparing for an Atomic Safety Li ce n s ing Board mandatory hearing on the permit application. The hearing will determine whether the staffs environmental rev i ew , 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 cur r ently reviewing 9 CO L applications (14 new reactor units): 3 AP-1000: William S. Lee Stat i on 1 &2 , Shearon H arris 2&3*, Levy County 1 &2 , B e ll efonte 3&4*, and Tur key P oint 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 r eq uested to withdraw the Calvert Cliffs, Unit 3 co mbined license application.

On April 25, 2013, Dominion Virginia Power rev ised its technology selection from the 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 l icense 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 co nstruct and operate a nuclear power plant under the alternative licen s ing 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 Un i ted 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 li i cense for an application referencing the ESBWR design. On November 19 , 2015, the NRC Commissioners held a mandatory hearing on th e 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 lice n se 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 s ubmitted to the NRC in November or December of 2016.

NRG Report to ASME Nov e mber 2016

• B&W mPower (iPW R)-B&W is developing a modular, sca l ab l e 180 MWe iPWR. At this time, mPower has r educed its activ i ties in the mPower deve l opment , and have not p r ov i ded a submittal date for the applica tio n.

• T VA i s planning to subm it its early site pe rm it in t h e seco n d quarter of 2016 fo r its Clinch River s i t e near Oak ri dge, T ennessee.

• H oltec is developing th e Holtec In herent l y Safe Modular Underground Reactor SMR 160 design that has a 160 MWe e l ectrical power output. T hey plan t o pursue a Part 50 lic ensi ng process that r equires an ap pli ca n t to app l y for a construct i on permit and a subsequent operating l icense. They have not provided an app li cation subm it tal date.

• XEnergy has ind i cated it p l ans to sub mit a design ce rti ficat i on applica ti o n to the NR C within the next few years for its pebble-bed high temperature gas-cooled reacto r. The X E nergy reactor (Xe-100) is a helium-cooled reactor with a power rating of 125 M Wt.

• Adva n ced Reactor (non-l ight wate r reactors)

Guidance Development:

• NRC has received Idaho National L aborato r y (INL) gene r ated Department of Energy technica l report "Guidance for Developing P rincipal Design C r iteria for Advanced (Non-Lig h t Water) Reactors." The INL report is the culmination of phase one of a two-phase in iti at i ve by the DOE and the NRC to develop advanced reacto r sa fet y design criteria from which the principal des i gn cr it e r ia cou l d be derived for advanced reactor concep t s. The NRC will follow its norma l process for develop i ng and i ssuing regulatory guidance and anticipates complet i on of such guidance by the end of 2016. 6. Multinational Design Evaluation Program CMDEP) Activities MDEP is a multin at i onal i nitiat i ve to develop i nnova ti ve approaches to leverage the resources and knowledge of mature , experienced national regulatory authorities who are tasked with the regulatory design review of new r eac tor plant designs. Some of th e ispecific working groups estab li shed under t h e MD EP organizat i on tha t the NR C participates in are the Codes and Standa r ds Working Group (CSWG), whose goa l is to achieve ha rm oniza ti on of code requi r ements for pressure-bo und ary components, and the Vendo r I nspection Cooperation Working Group (VICWG), whose goa l i s to maxim i ze the use of the results o f inspections obtained from ot h er regulators' efforts in inspecting vendors. Vendor Inspection Cooperation Working Group (VICWGJ The MDEP VICWG was formed because component manufac t uring is curren tl y su bj ect to mu lti ple inspections and a u dits simi l ar i n scope and i n safety objectives , but cond u cted by di ff erent regulators to different criteria.

Th e primary goa l of th e V I CWG is to max imi ze the use of the resu l ts ob t ained from other regulators' efforts in inspecting vendors. The MDEP V I CWG contin u es to achieve its short-term goa l s and i s mak i ng progress towards achieving its long term goa l s. The V I CWG continues to focus on maximizing info rm a ti on sharing, joint in spect i ons (mu l t ip l e regulators inspec tin g to th e regulatory requirements of one country), and witnessing of other regulators

' inspections. The N RC participated in 6 witnessed and joint inspections.

NRG Report to ASME November 2016 The working group enhances the understanding of each regu lator's in spection 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 inspec tions consist of one regulator performing an i nspection to its c riteria, observed by representatives of other MDEP countries.

The benefits to the observing countr ies include additional information and added confidence in the inspect ion results. MDEP regulators are using the experience gained during conduct of VICWG witnessed inspect io ns in their inspection planning.

The MDEP VICWG held i ts 17 1 h meeting during the week of April 4 i n 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 Em i rates , India , Canada, Finland , and Turkey were not in attendance.

Because the meeting was hosted in Dijon , France it allowed mult iple 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 identifie d several inspection activities that could be conduc ted 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 Dre sse r and Siemens. The members are continuing to communicate by e-mail to plan and conduct inspections.

Codes and Standards Working Grou p (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 regu lato rs 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 co ncerning international harmonization efforts. In November 2015, the CSWG met w i th representatives from CORDEL and the SDOs. Representatives from CORDEL presented and discussed their harmonization effo rts 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 result s of their meeting in China with Chinese counterparts.

Th e representatives from the SDOs presented the status of the i r 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 Installat io ns), as many of the topics are growing beyond the MDEP mandate.

7. 10 CFR Part 21 Rulemaking NRG Report to ASME November 2016 Th e NRC staff is current l y reviewing the Revision 1 of NEI 14-09 , "G uidelines for I mplementation of 10 CIFR Part 21, Reporting of Defects and No ncompliance," dated F ebrua ry 2016, which the NRC plans to endo r se in a Regu lat ory G uid e. Th e N RC staff h as completed draft guide DG-1291 , "E valuating Deviations and Reporting Defect s and Noncompl ia n ce." Based on Pr oject AIM reco mm endations and r ebalancing of agency's work load , P art 21 rulemaking has b ee n ca n ce ll e d. The NRC will continue to work on issuing DG-1291 for pub li c comment l ater t hi s year and hopes to i ssue the fina l r egula tory gu i de in 2017. 8. Commercial Calibration Services Status B y letter dated Apri l 29, 2014 , the Nuclear Energy In st i tute (NE I) submitted Revision 0 of NEI 1 4-05, "Guide li nes for the U se of Accredita t ion i11 Lieu of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services ," to the U.S. Nuclear Regu l atory Commiss i on (NRC) for NRC staff review and endorsement.

NEI 14-05 provides an approac h for l icensees and suppliers of basic components for using l abo r atory accreditation by Accred it at i on Bodies (ABs) that are signatories to the I nternationa l Laboratory Accred i tation Cooperation (ILAC) Mutual Recognition Arrangement (MRA) (hereby after referred to as the ILA C accreditation process) in lie u of performing commercial-grade surveys for procurement of ca l ibration and testing services performed by domestic and international l aboratories accredited by ILAC signatories.

By letter dated February 9, 2015 (ADAMS Accession N o. ML 14 322A535), the NR C s t aff transmitted it s safety evaluation (SE) id en t ifying the guidelines contai ned in NEI 14-05, Revision 1 (ADAMS N o. ML 14245A391) as an acceptable approach for li ce n sees and supp li ers to m eet the r eq uir ements of 1 O CFR Part 50, Appendix B to use IL AC la bo r atory acc r ed it ation as part of the commerc i al-grade dedication process for proc u rement of calibra tion and testing services. NRC's endorsement of NEI 14-05 , Revision 1, expands the NRC's acceptance of the IL AC accreditat i on process first d oc um en t ed in SE on an Ar izona Public Service r e quest (ADAMS Accession No. ML052710224). NRC's ear li i er acceptance was limit ed to calibration l abo r atory services accredited by specific U.S. Accredit i ng Bodies (ABs). Th e SE (1) con firm s tha t NE I 14-05, R evis ion 1, ref le cts the IL AC accreditat i on process prev i ous l y approved; (2) provides an eva lu a ti on of the unique aspects of N E I 14-05, Revision 1; (3) constitutes forma l NR C endorseme n t. of the g u idelines in NEI 14-05, Rev i s ion 1, fo r using the ILA C accredi t a tion process in li e u of pe rf orming a commercialgrade survey; and (4) finds that the IL AC accred i tation process continues to satisfy the requirement s of Appendix B to 1 0 CFR P art 50 a nd , therefore , i s acceptab l e. On March 16 , 20 16 , t he NRC i ss u ed R egu l a to ry I ssue Summary (RI S), 20 1 6-01, " Nuclear Ene r gy I ns ti tute Guidance for the U se of Accreditation in Li e u of Commercial Grade Surveys for Procurement of Laboratory Calibration and Test Services." T he NRC staff is issuing this R I S to notify add r essees of o n e method found acceptable by the NRC staff for procurement of calibration and test i ng services performed by domestic and internationa l laboratories for u se in safety-re l ated app li ca ti ons. Both domestic and interna tional laboratories are required to be accred it ed by accred i tation bodies (ABs) t ha t are signato r ies to the Internationa l Laboratory Acc r editat i on Cooperatio n (IL AC) Mutual Recognition NRG Report to ASME Nov ember 2016 Agreement (MR A) (he r eafter r eferred to as the IL AC accreditatio n process) in order for licensees a n d suppl i ers o f basic compo n ents to u se these services in lieu of perfo r ming commercial-grade s urv eys. On April 1, 20 1 6 , th e NRC staff approved th e license amendmen t request submitted by Un ion Electric Company (dba, Ame r en M issouri , the licensee) to change the operat i ng quality assu r ance program rev is ion 31, for Callaway Pl ant , Unit 1, to adop t NEI 1 4-05. Revision 1. 9. NRC Staff Review of EPRI 1025243 Guideline for Commercial-Grade Design and Analysis Comput e r Programs By letter dated July 18, 2012 , the Nu c l ear Energy In st it ute (NE I) submitted Electric Power Research In stitute (EPR I) 1 025243 , Plant Engineering

Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Related Applications for staff review and approval.

EPRI 1025243 desc r ibes a dedication methodo l ogy f or commercia l-grade design a n d ana l ysis computer programs for u se i n meet in g r eg ul atory requirements.

EPRI 1025243 fo llo ws th e method provided in EPR I NP-5652, which the NRC co n d i tionally endorsed in Generic Lette r 89-02. On July 1 , 2015 , the NR C issued for public commen t DG-1305, "Accepta nce of Commercial-Grade Design and Analysis Computer P rogra ms for Nu clear Powe r Pl ants." The DG prov i des new guidance that describes acceptance methods that the NRC staff considers acceptable i n meet i ng regu l ato r y r equ ir ements for accep t ance and ded i cat i on of commercial-grade des i gn and analysis computer programs for nuclear power pl ants. The DG-1305 pub li c comme nt period is now closed. The NRC staff has received and evaluated 40 comments from NE I ,, other stakeho l ders and the pub l ic. Regulatory Gu i de 1.231, " Acceptance o f Comme r cia l Grade Design and Analysis Compute r Programs for N uclear Power P l ants," is in the fina l review process and sho uld be i ssued in the fall. Thi s will be the first software-specific Commercia l-Grade Dedi cation (CGD) guidance with a limi ted scope fo r safety-rela t ed use of comme rc ia l Design & Analysis Computer Programs.

10. NRC Staff Review of EPRI Guideline for Dedication of Commercial-Grade Items for Use in Nuclear Safety-Related Applications I n September , 20 14 , the EPRI issued the 2014 T echnical Report 3002002982 , "Plant Engineering:

Guideline for the Acceptance of Com m ercial-G r ade It ems in Nuclear SafetyRe l ated Applicatio n s" -Revision 1 to EPR I N P-5652 and TR-102260.

The NR C staff participated during many of t he E PRI technica l adviso r y group (T AG) meetings held at EPRl's off ic es i n Char lo tte, No rth Carolina.

This report describes a methodo l ogy that can be used to dedicate commercial-grade items for use in safety-re l ated applications.

T he scope of app li cations for which commercia l-grade item dedica t ion is used has evo l ved signi fi cantly since the EPR I publ is hed i t s reports Guidel in e for the U til i za t ion of Commercia l Grade I tems in Nuclear Safe t y Related Applications (N CIG-07) (NP-5652) and Supplementa l Gu i dance for the App li cation of EPR I Report NP-5652 on the Utilization of Com m ercial Grade It ems (T R-102260) i n 1 988 and 1994, respectively.

T he guidance in this final report reflec t s l essons l earned and addresses NRG Report to ASME Nov ember 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 , "De dication of Commercia l-Grade It ems for Use in Nuclear Power Plants ," was issue d on June , 30, 2016, for a 60 day public comment period. This new RG approves for use, in part, Revision 1 to the EPR I NP-5652 and TR-102260, with respect to acceptance of commercial-grade dedication of i tems 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 Genera l Membership Meeting and the Annual Vendo 1 r Meet ing 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 periodica ll y 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 indu stry. The NRC issues tr i p reports to document NRC observation of audits performed by NUPIC that are available at the below web-site link: h ttp://www.n r c.gov/r eac t ors/n ew-r eactors/ove r s i ghtlq u a l i t y-assu r ance/n up i c-in d us t rv.h tm l 12. Reverse Enineering lnfortmation Notice 2016-01 On July 15 , 2016, NRC In formation Notice 2016-09: Recent I ssues Identified when using Reverse Engineering Techniques i n the Procurement of safety-re l ated Components was issued. The NRC is iss uing 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 com ponent s were supp lied without first ver i fying the supplied components met all safety-related design requirements.

The NRC expects that recipient s will review the information for app l icability to the ir facilitie s and consider actions, as appropriate , to avoid similar prob l ems. 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 re l ated to license renewal: Current status of applications, staff reviews and approvals

• 83 units approved (81 operat i ng plants with renewed li censes) o 1 (2 units) i n hearings (Ind ia n Po i nt 2 & 3) -supplemental SER issued November 2014, hearings held November 2015 o 3 (4 units) comp l eted ACRS Full Committee meeting (Fermi, Grand Gu lf , LaSalle 1 & 2)

NRG Report to ASME Nov ember 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, o Nine Mile Po i nt 1 -August 22, 2009 o Ginna -September 18 , 2009 o Dresden 2 -De ce mber 22, 2009 o H.B. Robinson -July 31, 2010 o Monticello

-September 8, 2010 o Po i nt Beach 1 -October 5 , 201 O o Dresden 3 -January 12 , 2011 o Palisades

-March 24, 2011 o Vermont Yankee -March 21, 2012 o Surry 1 -May 25, 2012 o Pilgrim -June 8 , 2012 o Turkey Point 3 -July 19 , 2012 o Quad Cities 1 -December 14 , 2012 o Quad Cities 2-December 14, 2012 o Surry 2 -January 29, 2013 o Oconee 1 -February 6, 2013 o Point Beach 2-Mar c h 8, 2013 o Turkey Point 4-April 10 , 2013 o Peach Bottom 2 -August 8, 2013 o Fort Calhoun 1 -August 9, 2013 o Prairie Island 1 -August 9, 2013 2013 o Oconee 2 -October 6, 2013 o Browns Ferry 1 -December 20, 2013 o Cooper Nuclear Station -Jan. 18 , 2014 o Duane Arnold -February 21 , 2014 o Three Mile Island 1 -April 19 , 2014 o ANO 1 -May 20 , 2014 o Browns Ferry 2 -June 28, 2014 o Peach Bottom 3 -July 2, 2014 o Oconee 3-July 19, 2014 o Calvert Cliffs 1 -July 31 , 2014 o Hatch 1 -August 6 , 2014 o FitzPatrick

-October 17, 201 4 o DC Cook 1 -October 25 , 2014 o Pra i rie Island 2 -October 29, 2014 o Brunswick 2 -December 27, 2014 o Millstone 2 -July 31, 2015 o Indian Point 3 -December 12 , 2015 o Beaver Valley -January 29, 2016 o St. Lu c ie 2 -Mar ch 1 , 2016 o Browns Ferry 3 -July 2, 2016 o Calvert Cliffs 2-August 13, 2016 o Salem 1 -August 13, 2016 Technical Issues NRG Report to ASME Nov ember 2016 o Brunswick 1 -September 8, 2016

• Steam Generator Div ide r Plates and Tube-to-Tubesheet Welds -Steam Generator Task Force submitted a topical report to address necessity for these inspections

-LR-ISG-20 16-01 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-22 7-A program and a gap analysis to address expected aging differences between 60 and 80 years. Subsequent Licen se Renewal The NRC is sued 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. ML 15348A 111 and ML 15348A 153)

• 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. ML 15348A265)

The staff received public comments on February 29, 2016 and is c urrently evaluating the input. The SRP-SLR Report, the GALL-SLR Report, and a NUREG report on "Disposition of Public Comments and Technical Bases" are schedul l ed to be published in July 2017. More information on subsequent license renewal, including detailed informat ion on public meetings, can be found at: h ttp://www.n r c.gov/reac t ors/ope r at i n g/l i ce n sin g/renewal/su b seg u ent-l i cense-renewa l.h tm l Applications for subsequent license renewal are expected in the third quarter of 2018 (Peach Bottom Atomic Pow e r Station, Units 2 and 3) and by the end of the fir st 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 rene wal and aging management:

• N U REG/CR-4 513, Revision 2, " Est i mation of F r acture T oughness of Cast Sta i nless Steels during T herma l Ag i ng i n L WR Systems." ML 16 145A082 (Apr i l 20 1 6). 14. New Generic Letters S i nce the l as t Code Wee k , no Gene ri c L et t ers (G L) were issued. 15. N e w Information Notices S i nce t he l as t Code Wee k , the f ollow i ng In fo rm a ti o n N otice (I N) was i ssued: i. N o r ecently i ss u ed IN s of i nte r es t. 16. New Regulatory Issue Summaries S in ce the l as t Code W e e k , t h e follow in g R egu l a t ory I ss u e Summar i es (R IS) were issued:

• N o r ecently issued R ISs o f in t eres t. 17. NRC Publications of Potential I n terest to ASME S i nce t he la s t Code Wee k , t h e follow i ng pu blica ti on th at may b e of i nterest to AS ME was iss u ed: i. WAS H-1400 (N URE G/KM-0010), T he I n tr oduction of R isk Assess m ent to t he R egulat i o n of N u cl ear R eacto r s, Aug u s t 20 1 6

• Thi s i s an upda t e to WAS H-1 400 inco rp orating i n fo rm a ti on fro m a Nove mb er 20 1 5 p r ese nt at i o n , "WASH-1400 and the Origins of Probabilistic Risk Assessment in t he N uclear In dustry," i i. N UREG/B R-0292, Rev i sion 1, Safety of Spent F ue l T r anspo r tat i on , August 20 1 6

• Thi s 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. i ii. N U R EG-2201, P robab ili s ti c R isk Assessment and R egulatory D ecis i on M a k ing: So m e F reque n tly Aske d Questions , Septembe r 2016

• P robab i listic r i sk assessmen t (P RA) is an i mpo rt an t decis i on-support t ool at th e U.S. N uc l ear R egu l ato r y Commission.

T he avai l ab il ity of experie n tia l data for accide n ts , i ncl u d i ng th ose at th e Fuk u sh i ma D ai-i c hi n u clear power p l an t , r aises na tur a l q u es ti ons r ega r d i ng th e need fo r and u tilit y of PR A, wh i ch is, a t heart, a systems mode li ng-based analy t ical approach. T hi s report addresses t hese ques t io n s us in g th e fo rm at of f r equen tl y asked quest i o ns (F AQs). i v. D G-1 33 1 , Service l eve l I , II , Ill a n d I n-Scope Li cense R e n e w al P ro t ect i ve Coa tin gs Applied t o Nuc l ear Powe r P l an t s

• Th i s r egu l a t ory gu i de is th e proposed r evision 3 to Regu l atory Guide 1.54. The revisio n add r esses upda t ed r e f e r ences re l ated to coat i ngs and expands the g ui dance to i n clu de i n te rn a l coa tin gs o n i n-sc o pe (fo r li cen s e re n ewa l pu r poses) components. T hi s rev i s i o n a l so p r ovides g ui da n ce for n e w reactor d es i gns with the r ecognit i on t h at t h e licensee or app li cant may need t o ad j ust some featu r es based o n the p arti c ular p l a nt des i gn.

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 compu terized data screening software used for analyzing eddy current data obtained during the inspection of steam generator tubes. vi. Revision to Reg u latory 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), Revis ion 4 , issued in June 2010 , and i s not aware of any issues that would preclude it s use. Revision 4, of RG 1.28 extended the scope of the NRC's endorsement to include NQA-1, Part I I. Part II contains amplifying QA requirements for certain specific work activities that occur at various s tages 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 , in s tallation , in spec tion , 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 re cent ly iss ued guidance from the ASME (NQA-1 b-2011, NQA-1-2012 and NQA-1-2015). T he refore , the staff has developed draft RG-1326 with the intent to approve for use, with several regulatory positions, the guidance from ASME NQA-1 b-2011 , NQA-1-20 12 and NQA-1-2015.

• The NRC expects to is sue draft RG-1326 for public com m ent 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 s u bseq uent license renewal app licat io n r ev iew process. 11/10/16, 9: 00 AM -1 2:00 PM , Teleconference Refer to the NRC Public Meeting Web Page at http://meetings

.n r c.gov/pm n s/mtg for a list of all currently-scheduled public meetings and further details.

From: Sent: To: Subje ct: FYI only. From: Taylor, Nick lingam, Siva 18 Oct 2016 15:38:05 +0000 Pascarelli, Robert RE: Wolf Creek Rev i sed RRs -Interna l Discussion S e nt: Tuesday, October 18, 2016 11:35 AM T o: lingam, Siva <Siva.Lingam@nrc.gov>

Subject:

RE: Wo l f Cree k 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 watch i ng their emai l carefully.

From: Lingam, Siva S e nt: Tuesday, October 18, 20 1 6 10:32 AM To: Taylor, Nick <Nick.Taylor@nrc

.gov> Subj e ct: FW: Wolf Creek Revised RRs -I nternal Discussion For t he internal ca ll he l d yes t e r day at 9: 00 AM (Easte rn), I did in clude both t he r es i dent inspec t ors and David P ro ul x (see th e attac h ed schedu l er).

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

Subject:

RE: Wolf Creek--Acceptance Review for Relief Request 1 4R-04 Alternate examination of CROM nozzles (MF8456) T hank you. From: T sao, 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 lett er dated October 11 , 2016 , Wolf Creek Nu c lear Operating Corporation

{the licensee) submitted Re lief Request 14 R-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 Nu c l ear 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 r eview. 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 r equest does provide technical information in sufficient detail to enable the NRC s taff to proceed with its detailed technical review and make an independent assessment regarding the acceptability of the proposed rel ief request in terms of regulatory requirements and the protection of public health and safety and the environment.

If needed , the NRC staff may r equest for additional information to complete its techn i cal review.

From: Sent: To: Cc:

Subject:

Attachments:

Here it is. Stephen Cumblidge Materials Engineer Cumblidge, Stephen 19 Oct 2016 15:32:01 -0400 Sengupta, Abhijit Collins, Jay RE: sharepoint lin k Volumetric Leakage Path.pptx 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:

s harepoint link Stephen How are you. Could you please send me the s h arepoint link where presentations a r e sto r ed from today's call. Thanks, Abhijit

..c I ' ro 0.... Q) I ' C) c ro ID E ID en _J en h en Q) <:( E :::J 0 >

Ultrasonic inspections performed to see flaws in welds and penetration tubes need to scan above and below the weld, 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 the interference fit than leaking nozzles. Interference __ -. No Leak Leak 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. Very Littl e reflection

?% ( ( ( ( ( ) ) ) ) ) ) ) } ) )I) ) ) ) ) 0% Weld ))l)))))))))})))))))))))

100%

• _ _,,,,--Air Total reflection " Cladding (Stainless St Butter (Alloy 82/182)

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 ..... -.... .. ._ , .. _..._l_ ... ....,, .. u u-,.....,,...... --0 to 1 70 deg. Circu m fere n ce 0 ,..... 0 00 0 3 3 )> x l--t Q.) 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. vVe s tiu g h o u se D ata PNNL 2.25 MH z Dat a )' w* r ==" -'we tt ed Side Wetted S id e 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 I usions 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: Sent: To:

Subject:

Drake, J a m es 19 Oct 2016 14:34:44 -0500 Coll i ns, Jay RE: Fw: See Attached I think the response is adequate. Jim From: Co ll i n s, Jay Sent: Wednesday, October 19, 2016 2:34 PM To: Drake, James <J ames.Drake@nrc.gov >

Subject:

RE: F w: 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?" I s 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: Co llin s, Jay <J ay.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 li ns, Jay Sent: Wednesday, October 19, 2016 10:52 AM To: Burne ll , Scott <Scott.Burnell@nrc.gov > Cc: Dra k e, James <James.Drake@nrc.gov >

Subject:

F W: 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, "App licability 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/ML 1506/ML 15068A 1 31.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 connect i on we l d, 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 nozz l e butt welds fabricated with Alloy 82/182 weld filler metal. There are some dissimi l ar metal nickel a l loy branch connection butt welds in the ASME C l ass 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 volumetr i c 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. I f you have additional i nformation or concerns , please feel free to contact us , Jay Collins (301 )415-4038 From: Sam nuke [mailto:nukewatcher@hotmail.com ] 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 Co llin s , severa l B& W plants have outages and h ave not inspected their b ranch connections in whic h the hole in the pipe is greater than 2" with a volume metic method , if th e process p i ping is l ess than 2" and the branch we ld is g r eate r than 2" does that branch weld requir e a volumetr i c exam? If so that i s not bein g done From: Sam nuke Sent: Sunday, September 18, 2 01 6 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 Re l ief from Paragraph-3200(b) of ASME Code Case N-729-1 for Reactor Vessel Head Penetration Nozzle Welds and Relief Request 1 4R-04, Request for Relief from the Requirements of ASME Code Case N-729-1. FYI From: E-RIDS3 Resource S e nt: Wednesday, October 19, 2016 3:07 PM To: WolfCreekEIS Resource <WolfCreekEIS.Resource@nrc.gov >; Watford, Margaret <Margaret.Watford@nrc.gov>; Singa l , Balwant <Ba lw ant.Singal@nrc.gov>; RidsRgn4MailCenter Resource <RidsRgn4 Ma ilCenter.Resource@nrc.gov>; Rid s Re sDE Resource <RidsResDE.R esource@nrc.g ov>; RidsNrrPMWolfCreek Resource <RidsNrrPMWolfCreek .Re s ource@nrc.gov >; RidsNrrDorllpl4 -2 Resource <Rids Nrr DorlLp l 4-2.Resource@nrc.gov>; RidsManager Re source <RidsManager.Resource@

nr c.gov>; Regner, Lisa

; Pascarelli, Robert <Robert.Pascarelli @nrc.gov>; Lyon, Fred <Fred.Lyon@nrc .gov>; Lingam, S iv a ; !Burkhardt, Janet <Janet.Burkhardt@nr c.gov>

Subject:

Relief Request Number 14R-03, Request for Relief from Paragraph-3200(b) of ASME Code Case N-729-1 for Reactor Ve ssel Head Penetration Nozzle Weld s and Relief Request 14R-04, Request for R elie f from the Requirements of ASME Code Case N-729-1. ADAMS Distribution Notification A0 4 7 -OR S ubmit ta l: In serv i ce/Test in g/Relief from ASME Code; r e l ated correspo nd e n ce 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 H ead Penetration Nozzle Welds and Relief Request 14R-04, Request for Relief from the Requirements of ASME Code Case N-729-1.) V i ew ADAMS PS Properties ML16293A5S1 Access i on ML16 2 93A581 Number R e li ef R e qu est Nu mb e r l 4R-03 , Requ es t for R e li e f from Paragraph-3 2 00(b) of Title ASME Code Case N-729-1 for Reactor Ve ss el Head Penetratio n Nozzle Weld s and R e li ef R eq u est 14R-04 , R eq u es t for R e li e f from th e R e quir e m e nt s of ASME Code Ca s e N-729-1. Dock e t 050004 82 Number Document 10/14/2016 Date Autho r McCoy JH Na m e Author Wolf Cree k Nuclear Operating Corp Affiliation Addressee Name Add r essee NRC/Docume nt Co ntr o l De sk Affi liation NRC/NRR Document Letter Type Ava il abi l i ty Publi c ly Avai l ab l e Date to be 10/2712 016 R elease d Document Non-Sensitive Sens iti v it y Co mment Date Added 10/19/2016 DPC a utoadd Keyword gps l s tt From: Sent: To:

Subject:

Pascarell i , Robert 19 Oct 2016 07:43: 46 -0400 Lingam , Siva RE: Comments on Relief Request 14R-03 Thanks Siva. Please set up a call sometime this afternoon. Dave Alley is in training a ll day but we should be OK if Jay Collins and so m e of t he o t her reviewers can m ake i t. From: Lingam, Siva Sent: Wednesday, October 19, 2016 7:00 AM To: Pascarelli, Robert <Robert.Pascarelli@nrc.gov> Cc: Singal, Balw a nt <Balw a nt.Singal@nr c.g ov>

Subject:

RE: Comment s on Reli e f Reque s t 14R-03 From: Taylor, N i ck Sent: Tuesday, October 18, 201 6 11: 5 0 PM To: Collin s , Jay <Jay.Collins@nrc .gov>; Lingam, Siv a <Siva.Lingam@nrc.gov >; T s ao , John <J ohn.Tsao@nrc.gov>; Alley, David <Dav i d.Alley@nrc .gov> Cc: Drake, James <J ames.Dra ke@nrc.gov>; Proulx , David <David.Proulx@nrc .gov>

Subject:

Comm e nt s on R e lief !Requ es t 14R-03 Good evening everyone , I'm sorry th is has take n so long for me to send out a note with my thoughts on th e relief req u est. I was a busy day today onsite. I am at Wo l f Creek, and actually went and stood on the head today , as well as spend i ng a s ignificant chunk of time t a lk ing with our I S i a n d RP inspectors , and spent about 1.5 h ours talking with their VP of Engineering this afternoon (who signed the relief request). I've have a few thoughts to share , and wou ld like to prov id e them t o help inform your decision. I nstead of putting them all in an email and possibly creating a l o t of email buzz , I th i nk it would be best to get on the phone sometime Wednesday to share my thoughts. It may be that grant i ng relief i s th e appropr i ate act i on -I ju s t want to be sure you all understa nd some o f the things in the request for relief a fu ll view of th e actua l conditio n s at the p l ant. Please let me know i f there is a good time for a short call to discuss on Wednesday. My on l y " bad" times are between 0830-1100 central time. Thank s! Ni c k Taylor Chief , Projects Branch B Divi s ion of Reacto r Projects USNRC Region IV 0: (817) 200-1141 C: l (b)(6) I E: rnc k.tayio r@nrc.gov I From: Co llin s, Jay Sent: Tuesday, October 18 , 2016 6:18 AM To: Tay l or, Nick <N i ck.Taylor@nrc .gov>; Lingam, Siva <Siva.L ingam@nrc.g ov> Cc: Tsao, John <J ohn.Tsao@nrc.gov >; Alley, David <Dav i d.Alley@nrc.gov >; Cumblidge, Stephen <Stephen.Cumblidge@nrc .go v>

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RE: Call with Wolf Creek regarding head inspectio n 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 morn i ng 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 i n 0-886 for our branch meeting from 9 to 10am. From: Taylor, Nic k Sent: 19 Oct 2016 15:47:50 -0500 To: Collins, Jay;Lingam, Siva;Tsao, John;Drake, J ames;Do d son, Douglas;Thomas, Fabian;Proulx, David Cc: Pascarelli, Robert;Alley, David;Cumblidge, Stephen

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RE: Wolf Creek Relief Requests 14R-03 and 14R-04 (CAC No. M F8456) 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 24 1 h. It's not quite clear yet what "clean" means or how they wi ll 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 Tay l or From: Co llin s, J ay Sent: Wedne s day , October 19 , 2016 2:37 PM To: Lingam , Siva <Siva.Lingam @nrc.gov>; Tay lor , Nick <Nick.T ay lor@nrc.gov>; Tsao , John <John.Tsao@nrc.gov>; D rake, James <James.Drake @nr c.gov>; Dodson, Douglas <D ouglas.Dodson @nrc.gov>; Thomas , Fab i an <Fabian.Thoma s@nrc.gov>; Proulx , David <David.Proulx @nrc.gov> Cc: Pascar e lli , Robert <Robe r t.Pascar e lli@nrc.gov>; All e y , David <David.A ll ey@nrc.gov>; C um b lid g e, Stephen <Stephen.Cumblidge @nrc.gov> S ubj ect: R E: Wolf Creek ReliefRequests l4R-03 and l4R-04 (CAC No. MF8456) <<F ile: Volumetric Leakage Path.p ptx >> Attached is Stephen Cumblidgc's slides explaining the volumetr i c leak path assessment. I f you have any questions please let us know. Jay -----Or i g inal F rom: Lingam, Siva Se nt: Wednesday , Oc t ober 1 9 , 20 1 6 7:44 AM T o: Lingam, Siva; Taylor, Nick; Collins, J ay; Tsao, John; Drake, James; Dodson, Dou g l as; Thomas, Fabian; Proulx , David Cc: Pa scare lli , Robert; Alley , D avid S ubj ect: Wolf Creek Relief R e qu es t s 14R-03 and 1 4R-04 (CA C No. MF8456) Wh e n: Wedne s day , October 1 9 , 20 1 6 1 2:00 PM-1:00 PM (UTC-05:00) Ea s tern Time (US & Canada). W h ere: HQ-OWFN-10B06-12p Plea se note th e following to discu ss the s ubj e ct RR s at th e r e qu es t of Nick Tay l or: Bridge No.: Pas s code: Date: Time: 877-935-1422 by# Octo b er 1 9 , 2016 (Wednesday) 1 2:00 PM (Ea s t e rn Time) We are st ill waiting for the licensee's repo1t p rovid in g ju st ifi cation for not in specting th e nozzle penetrations other than 12 n ozz l es m entio n ed in the subject RR s. This i s what I gat h e r ed from my BC w h o participa te d in the co n ference call held on October 1 7, 20 1 6, at 5 :00 PM (Eastern). From: Sent: To: Subje ct: Tsao, J ohn 19 Oct 2016 11:46:04 +0000 Lingam, Siva Accepted: Wo l f Creek Relief Requests 14R-03 and 14R-04 (CAC No. MF8456) From: Sent: To: Subje ct: Att a chment s: See attached emails Dave Alley, David 20 Oct 2016 14:16:00 +0000 Ross-Lee, MaryJane RE: Wolf Creek, Shearon Ha rri s, P a l o Verde RE: Harris Vesse l Head F l aw -Update to EN 52297 From: Ross-Lee, MaryJane Se nt: Thursday, October 20, 2016 8:58 AM To: Alley, David <David.Alley@nrc.gov>

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RE: Wo l f Creek, S h earon Harris, Palo Verde Where is the Shearon Harris indication? What location? Mary Jone Ross-Lee (MJ) D e puty Dire c tor , Division of Engineering Office of Nuclear Reactor Regulation OWFN 9H1 US Nuclear Regulatory Commission Off i ce: 301-415-3298 e-ma i l: m ary j ane.r oss-l ee@n r c.gov From: Alley, David S e nt: Wednesday, October 19, 2016 7:54 PM To: Ross-Lee, Mary J ane <MarvJane.Ross-L ee@nrc.gov > Subj ec t: Wo l f Creek, Shearon Ha rr is, Palo Verde M J A bit tonight and then I will update in t h e morni n g. Wolf C r eek They are p l anning to complete head c l eaning (met h od yet undefined) by t he 24th. We st ill have not h ing from the license concern i ng nozzles o t her than the orig i nal 1 2. We have been expecting some sort of docume nt at i o n for the last 2 days. Supposed l y it w ill come tomo rr ow (T h u rsday). Licensee appears to be bett i ng on r elief r equest be i ng g r anted as reg i on says they are not making plans t o do the surface exa m. B iggest i ss ue a p pears t o be a fea r on t h e p art of the l icensee of false ca ll s i f t he s u rface exa m is cond u cted. Reg i onal i nspector w ill look at head t omo r row t o check the status of nozz l es o t her t han t he 12 and l ook for evidence of whether the corros ion on th e head and nozzle annul i i s m ore than s u rface co r ros i on. If more than surface corrosion observed, it could indicate that corrosion has been occurring for longer than the canopy sea l leak. This might cause u s to believe that the re could be a leak through a J groove weld. Und er those ci r cumsta n ces, we may not want to au thorize til e proposed alternative. Shearon Harris Right now this is sounding more like a rounded ind ica tion on the weld (some disagreement concern ing l ocation). M ay be at a location of an indi cation at the time the weld was made. If rounded and especially i f 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 r epair 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 ind i ca tion on a 14 inch diameter line associated with safety inject ion tank). Materials and exact lo ca t i on of the indi cation are unknown. Dave David Alley PhD. Chief, Component Performance NDE and Testing Branch US Nuclear Regulatory Commission 11555 Rockville Pike Rockville MD 20852 301-4 1 5-2178 From: Sent: To: Cc:

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Attachments: All, Galvin, Dennis 20 Oct 2016 09:43:19 -0400 Collins, Jay;Tsao , John;All ey, David Barillas, Ma rth a;Butcavage, Alexander RE: Ha 1 rris Vessel Head Flaw -Update to EN 52297 Drawi n g of flaw in w e ld of prev i ous repair 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 in spectors 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 Licen sing 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.Ga lv in@nrc.gov>; Alley, David <D avid.A ll e y@nrc.go v> Cc: Barillas , Martha <Martha.Barillas@nrc.gov >; Dion, Jeanne <J eanne.Dion@nr c.go v>; Butcavage, Alexander <Alexa nd er.Butcavage@ nrc.gov>

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R e: H arr i s Ve ss e l H ead F l a w -Upd a t e to EN 52297 G r ee tin gs, The indi cat ion s that h av e b ee n found o n th e W es tinghou s e e mbedded fl aw r e paired no zz l e s urfac es hav e be e n attributed du e to a combi n ation of difficulti es assoc i a t e d with th e manual g rindin g activ iti e s , manual PT s a nd then th e pres s uri za tio n/th e rma l cyc lin g. l b e li eve wi th th e ha l f n o zz l e r e p a ir , ot h e r than th e l a c k of m a nual g rindin g, w e co uld h a ve s imil a r i ss u e s h e r e. Th a t is , unl ess th ere is a lin ear as pect to th e indic at ion , that c ould be a co n cern. Th a t i s why l would just like to get that information , whe n ava il ab l e. I don't know that we need to have a ca ll with t h e li censee, but just the final r esults of their in spection prior to flaw remova l , that would be appreciated. T h oughts? Jay From: Tsao, John Sent: Thur sda y , October 20, 2016 8:26:29 AM To: Galvin, Denn is; Collins, Jay; Alley, David Cc: Barillas, Martha; Dion, Jeanne; Butcavage, Alexander

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RE: Harris Vessel He ad 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 p r imary water stress corrosion cracking From: Galvin, Denni s Sent: Thursday, October 20, 2016 8:22 AM To: Collins, Jay <Jay.Collins@n r c.gov>; Alley, David <Dav i d.Alley@n r c.gov>; Tsao , John <J ohn.Tsao@n r c.gov> Cc: Barillas, Martha <Martha.Bar i l l as@nrc.gov>; Dion, Jeanne <J eanne.D i on@nr c.gov>

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RE: Harris Vessel Head Fl aw -U pdate to EN 52297 Here is the update from the RI this morning: Indicat ions 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-nozz l e and the reactor vessel h ead. This wa s an existi ng flaw that wa s i de n tified at the time of the repair , but was considered acceptable due the dimens i ons of the crack. The crack has since grown from .177 i nches to 0.331 inch es. Flaw is in the weld mater i al not the nozz l e. Licensee submitted a supp l eme nt to EN 52297 as another example of a degraded condition on the RVH. This is the first time that a previous l y repaired nozzle will need to undergo repair. Licensee is sti ll developing a strategy for path forward on Nozzle 23. From: Galvin, Dennis Sent: Thursday, October 20, 2 0 16 8:20 AM To: Collins, Jay <J ay.Coll i ns@n r c.gov>; Alley, David <Dav i d.Alley@n r c.gov>; Tsao , John <J ohn.Tsao@n r c.gov> Cc: Barillas, Martha <Martha.Ba r i l l as@nrc.gov>; Dion, Jeanne <J eanne.Dion@n r c.gov>

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Harris V esse l Head Flaw -U pdate 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 l icensee 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 [mailto: John.Caves@duke-energy .com] Sent: Thur s day, October 20, 2016 8:17 AM To: Galvin , Dennis <De nni s.Galv i n@nrc.gov>

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[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; Hi ckerson, Lonnie; Waldrep, Benjamin C; Ham i lton, Tanya M; Griffith, Donald L; Mi ll er, D Bryan; O'Connor, Sean Thomas; Grantham, Mark A; Womack, F r ankie L Cc: Zaremba, Arthur H.; Tr eadway, Rya n I; Green, Mary Kathry n; Sipe, R i ta B; Jones-You n g, All i son D; Volk, Samuel Joseph; Miller, Kris I; Fletcher II, Cecil Alexa n der; Grzeck, L ee; N olan, Chris; Rob ertson, Jeffrey N; Wasik, C hri stopher J; Zimmerman, Tony

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Update to EN 5229 7 NR C was updated at 1756 EDT about Nozzle 23. In g rid No rdby, P.E. Sr. Licensing En gineer , R egulatory Affairs H arris Nuclear Plant 919-362-2326 ingrid.nordby@duke -ene r gy.com From: Sent: To: Cc:

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Att a chment s: S teve , Riches., Mark 20 Oct 2016 09:26: 58 -0400 Ro se, Steven Jackson, Donna;Dodson, Jim;Galvin, Dennis; Barillas, Martha Drawing of flaw in we ld of previous repair Locati o n of Flaw on Repair ed Nozz l e.pdf The attached drawing in dica t es the a r ea of concern fo r No zz l e 23. L et m e know if you h ave a n y questions. Thanks, Mark J. R i c hes US NR C I R es id e n t In spec tor S h earo n Har ris Statio n P h on e: 9 1 9 362-06 01 E-mail: Mark.Riches @nr c.gov \ 112*MAX SEE NOTE 1 .70MIN FULL THICKNES S FROM TRIPLE POINT SEE NOTE 1 SURFACE UITABLE 20*M1N SEE NOTE 1 P T DETAIL B STEP4 WPS WP3143/F43TBSC3 F ILLER: ERNICrFe-7A AFTER MACHINING PER STEPS AFTER MACHINING PERSTEP5 UT --I =t--1-t --t ----t---STEP4 WELDING o .. ( 1.e.. WHEN REQUIRED, HOUSING EXTEN WELDED TO THE* ORIGINAL HOUSIN WELD FOR IOTB WELDING AND Ra AND IS REMOVED AFTER REMEDIA 1 SEENOTE5 NOTES: From: Sent: To: Doug l as;Kopriva, Ron C c: Subj e ct: No. MF8456) All, Alley, David 20 Oct 2016 22:07:10 -0400 Collins, Jay;Lingam, Siva;Drake, Jame s.;Taylor, Ni ck;Anchondo, l saac;Dod so n , Pascare ll i, Robert;Tsao, John; Cumblidge, Stephen;Singa l , Ba l wan t RE: Wolf c r eek -WCNOC response to verbal RAI for re l ief r e qu est 14R-03 (CAC 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) Releva n t cond iti ons for t h e purposes o f the VE sha ll i nclude areas of cor r osion, b o r ic acid d epos i ts, disco l o r ation, an d ot h e r evide n ce o f nozz l e lea k age. 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 re l evant cond i tions indicative o f possib l e nozzle leakage ... And 3142.3 Acceptance by Cor r ective Measures or Repair/Replacement Activity (a) A component with r elevant cond it ions not i ndicative o f possib le 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: T h e N RG considers any relevant condi ti on i n the annu l us r eg i on b e t ween t h e nozz l e and h ead s u rface t h a t cannot be re m oved by lig h t clean i ng ac t iv i ties to be a re l evant cond i tion of possib l e nozz l e l eakage. (Jay w h ere is this w r i tt en 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 i s 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 Pen e trat i ons w i th re l evan t co n dit i o n s i dentified 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 i n 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: Thur s d a y, Octob e r 20, 201 6 3: 36 PM To: Collin s , Jay <Jay.Collin s@nrc.gov>; Lingam, Siva <Siva.Ling a m@n rc.gov> Cc: Pascarelli, Robert <Robert.Pascarelli@nrc.gov >; T sao, John <J ohn.Tsao@nrc.gov>

Cumblidge, St e ph e n ; Sin g al, Balwant <Balwant.Singal

@nr c.gov>

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RE: Wolf cre e k* WCNOC r es p o n s e to verbal RAI for r e li e f reque s t 14R-03 (CAC No. MF8456) Paragraph 3 of section "Eva l uation of Vessel Closure Head Visual Examination Results" says "T here 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 cond ition" as if there is a gap betwee n the a n n u lus and the "problem" in my mind the problem may not be ev i dence o f leakage and, therefore, not r elevant. Based on what they have said, I agree w ith J ay t hat at l east an i nternal ca ll is needed. Dave From: Co ll ins, Jay Sent: Thur s day, Octob e r 20, 2016 3:18 PM To: Lingam, Siva Cc: Pascare lli, Robert <R obert.Pascarelli@nrc .gov>; A ll ey, David <Dav i d.A l l ey@n r c.gov>; Tsao, Jo hn <J oh n.Tsao@n r c.gov>; Cumbl i dge, Step h en <Stephen.Cumblidge@nrc .gov>; Singa l , Balwant <Balwant.S i nga l@n r c.gov>

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RE: Wolf creek -WCNOC response to verbal RAI for relief request 14R-03 (CAC No. M F8456) G r eetings, I n my opinion, th i s is complete l y inadequate to address the q u estion. I f the annu l uses of these nozz l es are not c l ear of bo r i c acid o r corrosion produc t , r egardless o f how the inspector t hought it go t there, t h en t he noz z le h as a re l e v a nt c on dition o f possi bl e no zz l e le a kage. Per N-729-1, s u pp l eme n t al examinat i ons are re q uired uncle r -3200 (b). I req u es t at least an i nternal p h one call this afte rn oon. J a y -3 142.2 Acce pt ance by S uppl e mental E xa min at ion. A no zz l e with r e le v ant condition s indi ca ti ve o f po ss ibl e n o zz l e l ea k age s h a ll b e acce pt ab l e for co n t inu e d se rv i ce if th e r e sult s o f s uppl e m e ntal examin a ti o n s [-32 00 (b )] m ee t th e r eq uir e m e nt s of -3 1 3 0. -3.141 Ge n e ral (c) R e l eva nt co ndit io n s fo r th e pur po s es of th e V E s h a ll includ e a r eas of co rr o s i o n , b o ri c ac i d d epos it s, d i sco l ora ti o n , a nd o th er ev id e n c e of n ozz l e l ea ka ge. From: Lingam, Siva Sent: Thur s day, October 20, 2016 2: 5 5 PM To: C ollin s , Jay <J ay.Collins@n r c.gov> Cc: Pa s care lli , Robert <Robert.Pascarell i@nr c.gov>; Alley , David <Dav i d.A l l ey@nrc.gov>; T s ao, John <J ohn.Tsao@n r c.gov>; Cumbl i d ge, Step h en ; Proulx, Da vid <Dav i d.P rou l x@nrc.gov>; Taylor, Nick <N i c k.T aylo r@nrc.gov>; Dod s on , Dou g las <Doug l as.Dodson@n r c.gov>; T homas , Fabian <Fab i an.Thomas@n r c.gov>; Kopriva , Ron <Ron.K opr i va@nrc.gov >; Drake, James <James.Drake@nrc.gov>; Anchondo , I saac <l saac.Anchondo@n r c.gov>; S i ngal , Balwant <Ba l want.Singal@nrc .gov>

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Wo lf creek -WCNOC re s pon s e to verba l RAI for r eli e f r equest 14R-03 (CAC N o. MF 8456) A t tached p l ease find the licensee's response for nozz l es with boric acid (o t her than 12 nozzles) for your review/eva l uat i on fo r RR 14R-03 .. From: Stone Lucille M [ma i lto: l urocke@WCNOC.com) Sent: Thursday, October 20, 2016 2:33 PM To: Lingam, Siva Subj ect: [External_ Send er] FW: WCNOC response to verbal RAI for rel ief reque st 14R-03 From: Stone Lucille M Sent: Thursday, October 20, 2016 1:27 PM To: 'balw a nt.si ngal@nrc.gov'; 'nick.taylor@nrc.gov'; 'ron.kopr i va@nrc.gov'

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WCNOC respo ns e to verbal RAI for relief request 14R-03 All, Here is electronic copy. Hard copies in the mail. Lu Stone WCNOC Licensing From: Sent: To: S ubj e ct: C h eruvenki, Ga n esh 20 Oct 2016 15:02:14 -0400 Medoff, James;Co ll ins, Jay;Min, Seung K;Hi ser, Allen BM l--SLR; Wolf Creek upper head leakage From: Sent: To: C c: Subj e ct: FYI Jim Drake, J a m es 20 Oct 2016 13:51:32 -0500 Alley, David Coll i ns, Jay;Tsao, John;Cumblidge, Stephen;Hoffman, Ke i th FW: Wolf Creek Reactor Vesse l Head Nozzle Leakage and Corrosion From: Proulx, David Se nt: 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>

Vege l , Anton <Anto n.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 <Je r emy.Bowen@nrc.gov>; L yon, Fred <Fred.Lyon@nrc.gov>; Singal, Balwant <Balwant.Singal@nrc.gov>; Pascarelli, Robert <Robert.Pascare l li@nrc.gov>; Taylor, Nick <N i ck.Taylor@nrc.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>; Dodson, Doug l as <Douglas.Dodson@nrc.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov> Subj e ct: Wo l f Creek Reactor Vessel Head Nozzle Leakage and Corrosion Good a f ternoon , Please see the attached revised one-pager re l ated to the Wolf Creek reactor vessel head l e akage and corros i on. Fee l free to forward to all interested parties. We wi ll con ti n u e to update the in fo rm a ti on as the li censee completes the i r clean i ng and insp e c t ion act i v i ties. I f you have any add i t i ona l q u estions, p l ease contact me a t 8 17-200-1 56 1. Very respec t fu ll y, Davia:fi.oulx SP E-RPBB x 1 56 1 From: Sent: To: C c: Subj e ct: Att a chm e nt s: lingam, Siva 20 Oct 2016 15:04:4 8 -0400 Collins, Jay;Tsao, John Alley, David;Cumblidge, Stephen;Pascarelli, Robert FW: Wolf Creek Reactor Vesse l Head Nozzle Leakage and Corrosion 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 S e nt: Thursday, October 20, 2016 2:45 PM To: lingam, Siva

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FW: Wolf Creek Reactor Vesse l Head Nozzle Leakage and Corros i on FYI From: Proulx, David S e nt: Thursday, October 20, 2016 2:42 PM To: Kennedy, K r iss <Kriss.Kennedy@nrc .gov>; M orris, Scott <Scott.Morris@nrc .gov>; Pruett, Troy <Troy.Pruett@nrc.gov >; Lantz, Ryan <Ryan.Lantz@nrc.gov >; Vege l , Anton <Anton.Vegel@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov >; Drake, James <Jame s.Drake@nrc.gov >; Dricks , Victo ir <Victor.Dricks@nrc.gov >; Maier , Bi ll <Bil l.M aier@nrc.gov >; Moreno , Angel <Ange l.Mo r eno@n r c.gov>; Bowen, Jeremy <Jeremy.Bowen@nrc.gov>; L yon, Fred <Fred.Lyon@nrc.gov>; Singal, Balwant <Balwant.Singa l@nrc.gov>; Pascarelli, Robert <Robert.Pascarelli@nrc .gov>; Taylor, Nick <N i ck.T aylor@nrc.gov>; Th omas, Fabian <Fabian.Thomas@nrc .gov>; Dodson, Doug l as <Doug l as.Dodson@n r c.gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov > Subj ec t: Wo l f Creek Reactor Vessel Head Nozzle Leakage and Corrosion Good afternoon , P lease see the attached revised one-pager re l ated t o t he Wol f Creek reactor vessel head l eakage and cor r os i on. F ee l free to forward to a ll interested parties. We wi ll con ti n u e t o update t he infor m a ti on as t he li ce n s e e completes thei r clean i ng and i n spec ti on act i v i ties. I f you have any add i t i ona l q u estions, p l ease contact me a t 8 1 7-200-1 56 1. Very respectfu ll y , 1J avicf f/5,.ouf x SPE-R P BB x 1 56 1 Wolf Creek Reactor Vessel Head Nozzle Leakage and Corrosion Key Messages NOTE: Significant up dates annotated by date entere d. * (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 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 install e d an approved canopy s ea l clamp assembly (CSCA) on penetrat i on 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 res i dent in spectors mon i tored reacto r coo l ant system l eakage throughout th e operat i ng cyc l e. Data indi cated a steady very sma ll leak r ate (approximately 0.05 ga ll ons pe r m i nute), that s uddenly began to in crease on Au g ust 31, 2016. On Septembe r 2 , 2016 , Wolf Creek observed R CS unidentified leakage in excess of 1.35 ga ll ons pe r min ute (gpm), exceeding the TS l im it of 1.0 gpm. As a r esu lt , the l ice nsee in i ti ated a TS required shutdown on September 2 , 2016. Contact: N i ck Taylor, Chief, Reactor Projects Branch B (817) 200-1141 October 20, 2016
• 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 mechanica l j oint serving the core exit thermocouple nozzle assembly is not considered pressure boundary leakage.
• Following the shutdown, the licen see decided to commence t he ir 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 i nspected 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 (817) 200-1141 October 20, 2016 From: To: Cc:

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Date: Clark. The r esa Alley David; R ud!and David; Chernoff H aro l d Bowen I e r e my ND E O pE top i cs -potent i a l 1 1 /2 age n da i tem Frid a y, Oc t obe r 2 1 , 2 0 16 8:29:2 4 AM D ave a nd D a v e-I k now y ou r gro up s are foll owi n g seve r a l mate r ials/I S i i ss u es at t h e s i tes. We ha v en't bee n discussing the rela t ed event r epo r ts at our OEDO morning meeti n gs in much detai l as they come i n, but there was a request today to discuss t h em in ge n eral at t h e next qua r terly OpE briefing if there is a trend o r anyt hi ng i n terest in g to say from a co l lective perspective. O n es w e'v e see n rece n tl y incl u de W o l f C r eek n ozz l es, H a rr is n ozz l es, P alo Ve rd e S I p i ping. Agai n , n o cu r rent actio n , just potential top i c fo r t h e n ext meet i ng (N ovember 2 at lOam) if you thi n k it's app r opriate. T h a n ks! Theresa Valentine Clark Execu tiv e Technica l Assistant (Reactors) U.S. Nucl ea r Re g ul a tory Comm i ss ion T h eresa.Cla rk@orc.gov I 301-415-4048 I 0-4H10 From: Coll i ns, Jay Sent: 21 Oct 2016 12:36: 45 +0000 To: All e y, David;Lin g am, S iv a; Dr a k e , J a m es;Taylo r , Ni c k;An c hondo, I saa c; Dod s on, Douglas;Kopriva, Ron Cc: Pascare ll i, Robert;Tsao, John; Cumblidge, Stephen;Singa l , Ba l wan t

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RE: Wol f c r eek -WCNO C r es pon se to v e rb a l RAI fo r r e l ief r e qu est 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 genera l disclo s ure. V i s u al Exa mi n a t ion fo r Le aka g e o f PWR R eac to r H ea d P e n e t ra ti ons R e vi s ion 2 of I 006296, In clud e 2002 I n s p ec tion R es u l t s and MRP In s pect i on Guidan ce 6.2.6 Bone acid leakage from abo v e ma y leak through the insulation and co ll ect on the underside of the insulation. or run down the penetrations as s hown in Figure 3-19. This condition can greaU y affect the ability to characterize whether the leakage is occlUring from the annu l us of the penetration or from another s ource. A depo s it of thi s type can take s e v era l forms. It can tend to cove r portions of the head like a tight adhering coating (Figure 3-2 0) or take the form of loo s e granular material that ma y rest on the uphill s ide of the penetration (Figure 3-2 1). 6.2. 7 Compre ss ed air , in the range of 40-t>O p s i 1 4 kPa), or a v acuum directed at depo s it s ha s been used to distingui s h whether a depo s i t is l oo s e bui l dup of material s impl y re s ting against a penetration that i s ea s il y remo v ed or i s a tightl y adherin g depo s it. originating from the annulu s of a l eaking penetration. F i gure s 3-22 and 3-23 s how *1>efore and after" example s ofa penetration eva l uated in this fa s hion. Figure 3-2 3 s how s Penetration

1. 63 after air. at 6Qi p s i (41 4 kPa), wa s blown at the boron depo s it. Becau s e the leakage wa s from the atumlu s. th e force of the air bla s t did not r emove the boron depo s it. If the depo s ir s bad faJlen from anorher leak point. th e y would have awa y , l eaving a clean appearing penerrat i on. Afte r proper do c umentation.

it i s important to remo v e the s e depo s i ts before the next inspe c tion. 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-" r ess\n Air at O.posits (Shows That the O.posit Was by an Actual L..ak It Woukl Not Blow Away) (Sff 6.27 in Sedion 2.) From: All e y, David Sent: Thursday, October 20, 2016 10:07 PM To: Collin s , Jay; Lingam, Siva; Drake, Jame s; T a ylor, Nick; Anchondo, I s aac; Dodson, Dougla s; Kopriva , Ron Cc: Pasca r elli, Robert; T s ao, John ; Cumbl i dge, Stephen ; Singal, Balwant

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RE: Wolf creek -WCNOC respon s e to verbal RAI for relief reque s t 14R-03 (CA C No. MF8456) All, I read through the licenseea*Žs response to us and the code case info again. Now I dona*Žt think I was quite right in my email below. Not quite sure how I was looking at this this PM but it isna*Žt holding water tonight. Based on what Jay sent from the code case: 3140 INSERVICE VISUAL EXAMINATIONS (VE) -3141 General (c) Releva nt conditions for the purposes of the VE shall include areas of corrosion, boric acid depo s its , disco l oration, and other evidence of nozz l e leakage. T his defines relevant conditions. It does not define l ocation on the head or nozzles, i.e., a relevant condition can be at the annulus or between nozz l es 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 Cor re ctive Measures or Repair/Replacement Activity (a) A component with relevant conditions not ind ica tive 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 noz z le and head surface that cannot be removed by light cleaning activities to be a relevant condition of possible nozzle l eakage. (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. Hav i ng 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*Žs a*celight cleaning activitiesa* J. Based on all the above, statements of significance in the licenseea*Žs 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 i n the annulus area. (This statement is inconsistent w i th the ev i dence so far presented to the NRC) and Reactor vessel head insulation The remaining noz zles were a lso ca refully reviewed both i n person and by video footage. The nozzles with residue buildup were carefully examined to the po int 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 eva lu a ting 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*Ž s precedent statement. If they have boric acid in the annulus and they dona*Žt get it up by vacuuming, it doesna*Žt seem possible for them to reach a conclusion that it didna*Žt 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*Žs 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: Collin s, Jay <Jay.Collins@nrc.gov >; Lingam, S iva <Siva.Lingam@nrc.gov > Cc: Pa s carelli, Robe rt <Robert.Pascarelli@nrc.gov >; Tsao, John <John.T sao@nrc.gov >; Cumblidge, Steph e n <Stephen.Cumb l idge@nrc.gov >; Singal, Balwant <Ba l want.Singa l@nrc.gov>

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RE: Wolf creek* WCNOC response to verbal RAI for relief request 14R-03 (CAC No. MF8456) Paragraph 3 of section a*reEvaluation of Vesse l C l osure Head Visual Examination Resultsa*1 I says a*re T here were relevant conditions in close prox i m i ty t o many nozz l esa*rl. Based on defin i tions i n the code case t h i s pu t s all those nozz l es in need of inspect i on. Given that they say a*rein c l ose proxim i tya*CJ ind i cates to me that they may not understand a*rere l evant conditiona* 1 as ifthere is a gap between the annulus and the a*reproblema* CJ i n my mind the proble m may n ot be evide n ce of l eakage a n d, therefore, not r eleva n t. Based on what they have said, I agree w i th Jay t hat at l east an i nternal ca ll is needed. Dave From: Co ll ins , Jay Sent: Thursday, October 20 , 2016 3:18 PM To: Lingam, Siva Cc: Pascarelli, Robert <Robert.Pascarelli@nrc .gov>; Alley , David <David.Al l ey@nrc.gov >; Tsao, John <J ohn.Tsao@nrc.gov >; Cumbl i dge, Step h en <Stephen.Cumb l idge@nrc.gov >; Singa l, Balwant <Balwant.Singal@nrc.gov >

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RE: Wolf creek -WCNOC re s ponse to verbal RAI for relief request 14R-03 (CAC No. MF 8456) Greetings, I n my opinion , th i s is completely inadequate to address the question. I f the annuluses of these nozz l es are not clear of boric acid or corrosion product, regard l ess of how the inspector thought it got there, t hen t he nozzle has a re l evant condition of possible nozz l e leakage. Per N-729-1, supplemental examinations are required under -3200(b). I request at l east an internal phone call this afternoon.MsoNormal"> J ay -3142.2 Acce ptan ce by S uppl e m e ntal Exa min at ion. A n o z z le w ith r e l ev ant co nditi o n s indi cat i ve o f po s s ibl e no zz l e l e ak a ge s hall b e acce pt a bl e for c ontinu e d se r v i ce if th e r e s ul t s of s u ppl e m e n ta l exam in at i o n s (-3 2 0 0(b)] m ee t th e r e quirem e n t s of -3130. -3141 Ge n e ral (c) R e l eva nt co nditi o n s for th e purp oses o f th e V E s ha ll includ e areas of corros i o n , bori c a c i d d e po s it s , di sc ol o r a ti o n , a nd o th e r e vid e n ce o f n oz z l e l ea ka ge. From: Lingam, Siva Sent: Thursday, October 20, 2016 2:55 PM To: Collin s , Jay <Jay.Coll i ns@nrc.gov> Cc: Pascare lli , Ro b ert <Robert.Pascarelli@ nr c.gov>; Alley, David <David.Al l ey@nrc.gov>; Tsao, Jo hn <J ohn.Tsao@nrc.gov >; Cumbl i d ge, Step h en <Step h en.Cumblidge@nrc.gov >; Proulx, David <David.Proulx@nrc .gov>; Taylor, N ick <N i ck.Taylor@nrc .gov>; Dod s on , Dougla s <Douglas.Dodson@n r c.gov>; T homas, Fabian <Fab i a n.Thomas@n r c.gov>; Kopr i va, Ron <Ron.Kopriva@nrc.gov >; Drake, James <J ames.Dra k e@nrc.gov>; Anchondo , I saac <l saac.Anchondo@n r c.gov>; Singal, B a lwant <Balwant.Singal@nrc .gov>

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Wolf creek -WCNOC response to verba l RA I for rel i ef request 14R-03 (CAC N o. MF8456) Attached p l ease find the licenseea* T M s response for nozz l es with bo ri c acid (other than 12 nozz l es) fo r your rev i ew/evaluation for RR 14R-03 .. From: Stone Lucille M [mailto: l urocke@WCNOC.com ] Sent: T hur sday, October 20, 2016 2:33 PM To: Lingam, Siva

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[External_

Sender] FW: WCNOC response to v erba l RAI for relief request 14R-03 From: Stone Luc ill e M Sent: Thur sday, October 20, 2016 1: 27 PM To: 'ba l want.singa l@nrc.gov'; 'nick.tayl o r@nrc.gov'; 'ro n.k opr i va@nrc.gov'

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WCNOC response to verbal RAI for relief r equest 14R-03 All, H ere is electronic copy. Hard copies in the mail. Lu Stone WCNOC Licensing From: Sent: To:

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Coll i ns, Jay 21 Oct 2016 11:27: 56 +0000 Lingam , Siva Accepted: Wo l f Creek Relief Request 14R-03 (CAC No. MF8456) From: Sent: To: Cc:

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FYI From: Drake, James lingam, Siva 21 Oct 2016 07:38:02 -0400 Collins, Jay Pascarelli, Robert;Alley, David; Tsao, John;Cumblidge, Stephen RE: Internal call with NRR concerning Wolf Creek Head inspection rel i ef request 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

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Internal call with NRR concerning Wolf Creek Head in spection relief reque s t We had a call with Dave Alleya*Žs group to discuss the licenseea* Žs 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 i n 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 i nclude areas of corrosion, boric acid deposits, discoloration, and other evidence of nozzle leakage. The NRC considers any relevant condition in the annulus region between t he nozzle and head surface that cannot be removed by light cleaning activities to be a relevant cond i tion of poss ib le no zz l e leakage. The code states in part, a*ce(c) A nozzle whose VE indic.ates relevant conditions ind icative of possible nozzle leakage shall be una ccep table for continued service unless i t meets the requirements of-3142.2 or -3 142.3. -3142.2 Acceptance by Supplemental E xa mination. A nozz l e with relevant conditions i ndi cative of possible nozzle leakage shall be acceptable for continued service if the re s ults of supp l emen tal examina ti ons [-3200(b)] meet the requirements of-3130. 3130 i s the l nservice Volumetric And Surface Examination

s. In the s ubmittal the l icensee uses the term a*cerelevant co ndition , a*D then goes on the state in part , a*ceRough c l eaning was performed using a vacuum cleaner. The suc tion created by the vacuum cleaner was minimal and i ncapable of removing particulate from surfacesa*:a*

D Th e licensee made the following th e stateme nt s; a*ceThe logi c used in evaluating the pe ne t ration s 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.a*D a*ceThere were relevant conditions in close proximity to many nozzles as well as a large percentage of the vessel head surface not i ncluded in the examination areas adjacent to the nozzles. These encompassed various forms of relevant conditions, but none were/are indi cative of pressure boundary le akage from the vessel closure head.a*D Based on t his information , we a r e setting up a call with the l icensee for 0930 cen tr a l , 1030 eastern time to discuss the process t hey used in app l ying the code case and determine if they have jus t used the te r ms incorrectly but d i d properly apply the code case and disposi ti on the indications correct l y. Understanding this is necessary before considering approval of the re lief request. Siva Li ngam is ar r ang i ng the logistics for the conference call. Jim $m ie.r 'f'. :Drak e James F. Drake Office phone: 817-200-1558 Cell Phone: l (b)(6) I From: Lingam, Siva Sent: 21 Oct 2016 07:26:40 -0 400 To: Collins, Jay;Tsao, John;Alley, David;Cumblidge, Stephen;Taylor, Nick;Drake, James;Dodson, Douglas;Thomas, Fabian;Proulx , David;Kopr i va, Ron;Anc hondo, lsaac;Pick, Greg;Kalik i an, Roger;wimulie@WCNOC.com;cyhafen@wcnoc.com;jaknust@WCNOC.com Cc:

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Att a chment s: P ascarell i , Robert;Werner, Greg Wolf Creek Relief Req u est 14R-03 (CAC No. MF8456) ET16-0028.pdf Please note th e following to discuss the subject RR with the licensee based on the attached response from th e l icensee: Bridge No.: Pa ssco d c: Date: Tim e: 877-935-1422 by# October 2 1 , 20 1 6 (Friday) I 0:30 AM (Eastern Tim e) From: Sent: To:

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Thanks Alley, David 14 Oct 2016 19:42:14 +0000 Collins, Jay RE: RE: WCNOC RV pictures 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>

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Fw: RE: WCNOC RV p ic tures Greetings, It lo oks lik e you will get access. The numbering is a bit confusing. Once you get connected, in the first folder is a lis t 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 l abe l ed in the picture. I believe this is a view t h at I saac gave us prev i ous ly , but not the same photo. Note that the vent line in th e picture between the two penetration nozzles. Now , if you go to the head map image l isted as M-706-00009 _REACTOR PE N in the folder, you will find that penetration nozzles numbers 67 and 54 are no where near no zzle 77, the source of the spi ll. Instead, number 67 is at approximately 320 degrees an d nozzle 54 is at 290 degrees, near the periphery, i n the sout h-w est quadrant of the head. Note also, th ey are not near the head vent l ine, which is at about the 45 degree location in the North-wes t quadrant of the head. I believe nozzle 67 is the nozz l e 76 that I saac circled with a question mark in the i m ages he sent on Thur sday. Either way, 67 or 76, it has r emain in g indi cat ion s in the annu lu s r eg i on and is not li sted for volumetric inspect i on. I will look over the photos this weekend, but i think we will p erhaps n eed an internal discussion on Mond ay for a bit. Jay From: Good Nicole R <nilyon@WCNOC.com > Sent: Friday, October 1 4 , 2016 10:25 AM To: Singal, Balwant; Ungam, Siva Cc: Lingam, Siva; Co llin s, Jay; T sao, J ohn; Alley, David; Pascarelli, Robert

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[External_Sender) RE: WCNOC RV pictures Access has been provid ed to: Siva Lin gam Jay Collins J ohn T sao Access has been r e q uested for: Ba l want S i ngal David Alley Robert Pascarelli Thank you, Nico l e Good From: Singal, Balwant [ma i l to:Balwant.S i nqal@n rc.gov] Sent: Thursday, October 13, 20163:12 PM To: Good N i cole R; Lingam, Siva Cc: Lingam, S i va; Collin s , Jay; Tsao, John; Alley, David; Pascarelli, Robert 

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RE: WCNOC RV pictures Nicole, Not clear if we already have the access or will be getting access l ater. P l ease 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 Ba l want.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 . From: Good Nicol e R [mailto: ni l yon@WCNOC.com) Sent: Thur s day, October 13, 2016 4:05 PM To: Lingam, Siva Cc: Singal, Balwant <Balwant.S i ngal@nrc.gov >

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[Ext e rnal_Send e r] WCNOC RV pi c tur es I was told you would like pictures of the penetrations with labels of the penetrat i on number. I have only been ab l e to locate a few pictures , at this point. I have granted you access to the Certrec I MS Sept 2016 Forced Outage. I tem #14 has five pictures that may be he l pfu l (D CS00006, D CS00039, D CS00029, D CS000 1 9, and D CS000 1 8). 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 y ou , Nicole Good Licensing n i lyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek Nuclear Operating Corporation From: Sent: To: C c: Subj e ct: Singal , Balwa n t 24 Oct 2016 10:22: 25 -0400 T sao, John;Alley, David;Collin s, Jay;Ka likian , Roger Pascare ll i, Robert;Lingam, Siva FW: EB2 Acting BC October 24 -28 I just spoke with the l icensee. 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 i ll 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 Ba l wa nt.S i nga l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 From: Proulx, David Sent: Monday , October 24, 2016 9: 57 AM T o: Thomas, Fabian ; Singal, B a lwant Subj e ct: FW: EB2 Actin g BC O c tob e r 24 -28 From: Werner, Greg S e nt: Monday, October 24, 2016 7:20 AM To: Tay l or, Ni ck <N i ck.T ay l or@nrc.gov>; Prou l x , D a vid <Dav i d.Prou l x@nrc.gov>; M a t e y c hick , John <J ohn.Mateych i c k@nrc.gov> C c: Drake, James <James.Drake@nrc.gov > Subj e ct: RE: EB2 A cti n g B C Octob e r 24 -28 Just got through talking with Jim Drake, seems pretty clear what HQs is telling WC. I f WC has relevant indications on the head , besides the 12 nozzles, WC has to do the supplemental inspe ctions. WC cana*Žt just do visual from the top and make a c l aim 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, N ick S e nt: Monday , October 24, 2016 7:05 AM To: Proulx, D avi d <Dav i d.Prou l x@nrc.gov>; M ateychick, John <John.Mateychick@nrc .g ov> C c: Werner , Greg <Gr e g.W e rner@nrc.gov >; Drak e, Jam es <J ame s.Drake@nrc.gov> S ubj ec t: FW: EB2 A c tin g BC Octob e r 24 -28 David I J ohn , I recommend the two o f you get sy n c h ed up quick l y th i s we ek on t h e status of t h e Wollf Creek rel i ef req u est & r eques t for addi ti onal i nfo f r om headq u a rt ers and EB 2. T he Wo lf C r eek manage r s a r e not h appya*: T hanks, Nick From: Werner, Greg Sent: Mond ay, October 24, 2016 6:58 AM To: R4DRS-EB2 <R4DRS-EB21@nrc.gov>; Hay, M i chael <Michael.Hay@n r c.gov>; R4ACES <R4ACES@nrc .gov>; R4DRS-BC <R4DRS-BC1@nrc.gov>; R4DRP-BCandSPE <G-R4-DRPBCand S PE@nr c.gov>; Vegel, Anton <Anton.V e gel@nrc.gov>; C l ark, Jeff <Jeff.C l a rk@nr c.gov>; Pruett, Troy <Troy.Pruett@nrc .gov>; Lantz, Ryan <Ryan.Lantz@nrc.gov >

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EB2 Acting BC October 24 -28 John Mateychick wi ll be the acting EB2 branch chief from today, October 24 , thru Friday, Oc t ober 28. I wi ll becheck j n q e ro ails and m y ce ll voicemails period i cally throughout the week. My cell number I ------Greg Werner From: Sent: To: Cc:

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Attachments: Collins, Jay 25 Oct 2016 14:32:39 +0000 Cumblidge, Stephen;Singa l , B a lw ant Kalikian, Roger RE: Important details for Wolf Creek Wolf Creek verbal auth 14R-03 10-17-2016 R ev 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: Tu es day, Octob e r 25, 2016 10:2 0 AM To: Collins , Jay

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Importa nt details for Wolf Creek What are the main issues tha t you would l ike covered for the Wolf creek rel ie f? 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 a re ignoring relevant i ndications. Also , please send over your drat sc rip t. Stephe n Cumblidge Materials Engineer US Nuclear Regulatory Commiss i on Mail Stop OWFN/9 H6 Washington, DC 20555-0001 Telephone: (30 1) 415-2823 (Office) V ERB A L AU TH O RI ZA TI ON B Y THE O FFI CE O F N U CL E A R RE AC T O R RE G U LA TI O N R E LI EF R EQ U ES T 1 4R-03 A L TERNAT I VE TO U SE VOL U METR I C L EA K PA TH FOR SU P P L EME N TA L EXAMS WO LF C RE EK G ENE RA TI NG S T A TI O N WO L F CREEK NU C L EAR OPERA TI NG CORPORA T IO N DOCKE T N UMBE R 50-4 82 T e chnical Evaluation r e ad by David Alley , Chief of the Component Performance , Non-Destructive Examination , and Testing Branch , Office of Nuclear Reactor Regulation By l etter dated Octo b er 1 4 , 2016 , Wolf Creek N uclear Ope r at i ng Corporat i on (t he li censee) s u bmitted Re li e f Reques t 1 4R-03 for t he al t ernate exam i nation of con t ro l r od drive mechan i sm (CRO M) nozzle penetrat i on weld numbe r s 20 , 27, 35 , 40 , 46 , 47 , 58 , 59 , 63 , 70 , 71 and 77 a t the Wo l f Creek Gene r a ti ng Station. T he l i censee p r oposed (a) to perfo rm a volumetric leak path assessment of each penetra ti on nozzle i n li eu o f the surface l eak path assessment requ i red by P a r agraph -3200(b) o f ASME Code Case N-729-1 , and (b) if an u nacceptable i nd i cation by t he l eak path assessment or vo l ume t r i c exa m is ident i fied , the li censee w ill revert to t he r equiremen t s of Code Case N-729-1 and 10 C FR 5 0.55a(g)(6)(i i)(D). T he li ce n see made t his reques t i n acco r dance w it h the req u iremen t s of 10 CFR 50.55a(z)(2), such that complia n ce with the speci fi ed requi r ements wou l d resu lt in hard s hip o r unus u a l difficul t y wi t hout a co m pe n sat in g i ncrease in the level o f qu a li ty and safe t y. T he N RC staff fi n ds that wh il e the demonstrated vo l u m etr i c l eak pa t h is no t equiv a l ent to a fu ll y q u ali fi ed s u rface lea k pat h assessmen t , th e li censee i d en ti fie d s u fficien t ope r at i ona l exper i e n ce , techn i ca l basis and rad i olog i ca l dose hardsh i p to s h ow t h at r eg u latory co m p l iance wou l d res u lt i n h ards h ip w i thout a co m pensating inc r ease in t he l e v e l o f qua li ty and saf e ty. For ope r ating exper i ence, t he licensee s h owe d t h at there has been no p r evio u s id e n t i fied cracking or l ea k age i dentified fro m the C RDM n ozz l e penetrat i ons o r we l ds of the uppe r head a t Wolf Creek. The N RC s taff noted t hat while th i s fact doe s not precl ud e the possibil i ty of c r ac k ing to be fou n d as the p l ant co n tin u es t o age, p l ants wh i ch h ave p reviously i d en ti fied c r ac king a r e more l i kely t o see subseque n t and m ore sign i ficant crack i ng i n the future. Given the lack of the in i tial cra c king be i ng i dentifi e d i n t h e n ozz l e heats o f ma t e ri a l , a t the ope r a t i ng t empe r a tu res o f Wolf C r eek , t he N RC fo u nd t hat t h e poten ti a l for s i gn i fi can t crack i ng this o u tage was l ess l ikely. F or t ech n ica l basis , th e li ce n see iden ti fied t hat t he ir i nspection wo ul d be in co m p li ance wit h the We s dye Tech ni cal J u st i ficat i on Docume nt s h owing an effec ti ve demonstration of the vo l umet r ic leak path tec h nique. T he N RC has accepted t h e use of a demonstrated vo l umetric l eak path as part of the upper head i nspect i on progra m under 1 0 CFR 50.55a(g)(6)(ii)(D). T he lice n s ee also referenced NU REG/CR-7 1 42, Ul trason i c P hased Ar r ay Assessment of the I nterfe r ence F it and L eak P ath of the North Anna U nit 2 Control Ro d D ri ve Mechan i sm N ozz l e 63 w i th Des t r u ctive Validat i o n , wh i ch fo u n d , i n part, th e u se of a p r oper l y foc u sed 0 degree probe cou l d detect a l eakage path under l ow leakage ra t es du r ing ope r ation that l ed t o mi n i mal was t age of the upper h ea d low a ll oy steel. Wh i le th e N RC staff d i d no t fi nd t hat t he vo l u m e tr ic l ea k path assess m ent was equ i va l ent to a qual i fied s u rface leak pat h assessment, the info r matio n does demonstrate the effectiveness of the volumetric leak path exam i nation to detect l ow l ea k a g e r ates, as pe rf o rm e d i n ac c o r da n ce with t h e li ce n see's propo se d a ltern a ti ve. , C o mm e n t I C J): Removed due to :' q u estion of specifics by Regional staff. I n / the NRC SE write-up I would consider For h a rd s hi p, th e l i c en se e not e d th a t a qualified s urf ac e l e ak p a th assess m e nt c ou ld b e / adding a stateme n t sim i l ar to ... performed in two manners that would require both addi ti onal radiolog i ca l dose and time / v e r s u s the p e rformanc e of a volum e tri c lea k path a ssess ment. {l:A&-lisensee-estima.too i 3.4 Rem and 10 days for an eddy current surface examination and 18 REM to perform a / liquid dye penetrant examination of all of the 12 penetration welds. both of the s e c onditions to be of suffi c ient hardship g i ven the operationa l experienc e and techni c al adequacy of the li c en s ee's propo s ed a ltern a tive ver s us the regul a tory requ i r e ment. Th e r e f o r e , th e NRC s t a ff f i nd s th a t th e l i ce n see's pr o po se d a lt e rn a tive pr o vide s reasonab l e assurance of structura l integrity unti l the next scheduled examination , and that co mpli a n ce w ith the s urfa ce e x a mi nat i on r e qu i r e m e nt s of P a r a gr a p h -3 2 00 (b) o f ASME Code Case N-729-1 , for the subject welds , would resu lt i n hardsh i p without a c omp e n sa tin g in c r ease in th e l e v el o f qu a lity a n d sa f e ty. Authorization r ead by R o bert P asca relli , Chief of the Plant Li ce n si ng Br a nch IV-1 , Offic e of Nuclear Re a ctor Regulation As Chief of the Plant Licensing Br a nch I V-1 , Office o f Nuclea r R e actor Regulation , I c on c u r with th e Comp o n ent P e rf o r man ce, Non-D es tru ct iv e Ex a minat io n , and T es ting Br a n c h's d e t e rmin a tio ns. Th e NRC sta ff c o nclud es that th e propo se d a lt e rn a tive pr o vid es r e a s onabl e assu ran ce of structural i ntegr i ty of the CROM penetration n ozzles numbers 20, 27 , 35, 40 , 46 , 47, 58 , 59, 63, 70 , 71 a nd 77 s u c h th a t co m p lying with th e A S ME Cod e r e quir e ment w o uld result in hard shi p o r unusual diffi cu lty w i thout a co mpen sa ting in c rea se i n the level of qu a l ity a nd sa f e ty. A cco rd i n g ly , th e NR C s t a ff co n c lud es th a t th e l i ce n see h as ad e quate l y addressed all of the regu l a t ory requ i rements set forth in 1 0 CFR 50.55a(z)(2) a nd 10 CFR 5 0.55a(g)(6)(i i)(D). T h e r e for e , th e NRC s t a ff au t h o ri zes th e u se o f r e li ef request 14 R-03 at the Wolf Creek G e nerat i ng St a tion during the c urrent r e f ue ling o utag e s ub jec t to t h e li ce n see's pr opose d a lt e rn a tiv e th a t i f an u n accep t a b le in d ica tion by th e leak path assessment or volumetr i c exam i s i dentif i ed, the licensee w i ll revert to the r e qu i r e m e nt s of Cod e Case N-729-1 a nd 10 CF R 5 0.55a (g)(6)(i i)(D). A ll o th er r e q ui r e m e nt s of A S ME Co d e , Sect ion X I , f o r whi c h re li e f w as n o t s p ec ifi ca lly r e que s ted a nd a uthori ze d by the NRC s t a ff r e m a in ap pli ca ble , in c luding th e th i rd p a rty review by the Autho r i ze d N u c l ea r In-se rvi ce In spec t o r. This v e rb a l authorization do e s not pr e c l ude the NRC staff from ask i ng addit io na l c l a rifi ca ti o n q u est i o n s r e g a r d in g R e li ef R eq u est 14R-0 3 , wh i l e pr e p a ring th e su b se qu e nt written safety eva l uation. In cons i d era t ion of t he radiologica l dose est im a t es, the N RC notes t h at t hese are conservatively hi gh estimates. The N RC bases this evalua t ion on actual survey dose rates i n the area. Add i tionally, t h e NRC staf f notes, that w i t h prior planning, the effect of this i nspect ion on outage schedule cou l d be s i g n i ficantly r educed. However, the NRC staff did fin d that even a l owe r bound radiologica l dose est i mate was s t ill s u fficient har d ship bas e d on t h e compensat i ng l evel of quality of the vo l umetric examinatio n g i ven no previous cracking had been found previous l y and d u r i ng this outage. I d o ubt tha t detail is n ecessa ry in t h e verbal, but i s included I n this comment in case a q uestion is r ai s ed by the region a s part of o u r r ev i ew conside r ations for this re lief. From: Sent: To:

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Attachments

Coll i ns, Jay 25 Oct 2016 11:36:33 +0000 All e y, David New wolf creek script Wolf Creek verbal auth 14R-03 10-17-2016 Rev 3.docx Attached changes for r eg i onal concerns.

V ERB A L AU TH O RI ZA TI ON B Y THE O FFI CE O F N U CL E A R RE AC T O R RE G U LA TI O N R E LI EF R EQ U ES T 1 4R-03 A L TERNAT I VE TO U SE VOL U METR I C L EA K PA TH FOR SU P P L EME N TA L EXAMS WO LF C RE EK G ENE RA TI NG S T A TI O N WO L F CREEK NU C L EAR OPERA TI NG CORPORA T IO N DOCKE T N UMBE R 50-4 82 T e chnical Evaluation r e ad by David Alley , Chief of the Component Performance , Non-Destructive Examination , and Testing Branch , Office of Nuclear Reactor Regulation By l etter dated Octo b er 1 4 , 2016 , Wolf Creek N uclear Ope r at i ng Corporat i on (t he li censee) s u bmitted Re li e f Reques t 1 4R-03 for t he al t ernate exam i nation of con t ro l r od drive mechan i sm (CRO M) nozzle penetrat i on weld numbe r s 20 , 27, 35 , 40 , 46 , 47 , 58 , 59 , 63 , 70 , 71 and 77 a t the Wo l f Creek Gene r a ti ng Station. T he l i censee p r oposed (a) to perfo rm a volumetric leak path assessment of each penetra ti on nozzle i n li eu o f the surface l eak path assessment requ i red by P a r agraph -3200(b) o f ASME Code Case N-729-1 , and (b) if an u nacceptable i nd i cation by t he l eak path assessment or vo l ume t r i c exa m is ident i fied , the li censee w ill revert to t he r equiremen t s of Code Case N-729-1 and 10 C FR 5 0.55a(g)(6)(i i)(D). T he li ce n see made t his reques t i n acco r dance w it h the req u iremen t s of 10 CFR 50.55a(z)(2), such that complia n ce with the speci fi ed requi r ements wou l d resu lt in hard s hip o r unus u a l difficul t y wi t hout a co m pe n sat in g i ncrease in the level o f qu a li ty and safe t y. T he N RC staff fi n ds that wh il e the demonstrated vo l u m etr i c l eak pa t h is no t equiv a l ent to a fu ll y q u ali fi ed s u rface lea k pat h assessmen t , th e li censee i d en ti fie d s u fficien t ope r at i ona l exper i e n ce , techn i ca l basis and rad i olog i ca l dose hardsh i p to s h ow t h at r eg u latory co m p l iance wou l d res u lt i n h ards h ip w i thout a co m pensating inc r ease in t he l e v e l o f qua li ty and saf e ty. For ope r ating exper i ence, t he licensee s h owe d t h at there has been no p r evio u s id e n t i fied cracking or l ea k age i dentified fro m the C RDM n ozz l e penetrat i ons o r we l ds of the uppe r head a t Wolf Creek. The N RC s taff noted t hat while th i s fact doe s not precl ud e the possibil i ty of c r ac k ing to be fou n d as the p l ant co n tin u es t o age, p l ants wh i ch h ave p reviously i d en ti fied c r ac king a r e more l i kely t o see subseque n t and m ore sign i ficant crack i ng i n the future. Given the lack of the in i tial cra c king be i ng i dentifi e d i n t h e n ozz l e heats o f ma t e ri a l , a t the ope r a t i ng t empe r a tu res o f Wolf C r eek , t he N RC fo u nd t hat t h e poten ti a l for s i gn i fi can t crack i ng this o u tage was l ess l ikely. F or t ech n ica l basis , th e li ce n see iden ti fied t hat t he ir i nspection wo ul d be in co m p li ance wit h the We s dye Tech ni cal J u st i ficat i on Docume nt s h owing an effec ti ve demonstration of the vo l umet r ic leak path tec h nique. T he N RC has accepted t h e use of a demonstrated vo l umetric l eak path as part of the upper head i nspect i on progra m under 1 0 CFR 50.55a(g)(6)(ii)(D). T he lice n s ee also referenced NU REG/CR-7 1 42, Ul trason i c P hased Ar r ay Assessment of the I nterfe r ence F it and L eak P ath of the North Anna U nit 2 Control Ro d D ri ve Mechan i sm N ozz l e 63 w i th Des t r u ctive Validat i o n , wh i ch fo u n d , i n part, th e u se of a p r oper l y foc u sed 0 degree probe cou l d detect a l eakage path under l ow leakage ra t es du r ing ope r ation that l ed t o mi n i mal was t age of the upper h ea d low a ll oy steel. Wh i le th e N RC staff d i d no t fi nd t hat t he vo l u m e tr ic l ea k path assess m ent was equ i va l ent to a qual i fied s u rface leak pat h assessment, the info r matio n does demonstrate the effectiveness of the volumetric leak path exam i nation to detect l ow l ea k a g e r ates, as pe rf o rm e d i n ac c o r da n ce with t h e li ce n see's propo se d a ltern a ti ve. , C o mm e n t I C J): Removed due to :' q u estion of specifics by Regional staff. I n / the NRC SE write-up I would consider For h a rd s hi p, th e l i c en se e not e d th a t a qualified s urf ac e l e ak p a th assess m e nt c ou ld b e / adding a stateme n t sim i l ar to ... performed in two manners that would require both addi ti onal radiolog i ca l dose and time / v e r s u s the p e rformanc e of a volum e tri c lea k path a ssess ment. {l:A&-lisensee-estima.too i 3.4 Rem and 10 days for an eddy current surface examination and 18 REM to perform a / liquid dye penetrant examination of all of the 12 penetration welds. both of the s e c onditions to be of suffi c ient hardship g i ven the operationa l experienc e and techni c al adequacy of the li c en s ee's propo s ed a ltern a tive ver s us the regul a tory requ i r e ment. Th e r e f o r e , th e NRC s t a ff f i nd s th a t th e l i ce n see's pr o po se d a lt e rn a tive pr o vide s reasonab l e assurance of structura l integrity unti l the next scheduled examination , and that co mpli a n ce w ith the s urfa ce e x a mi nat i on r e qu i r e m e nt s of P a r a gr a p h -3 2 00 (b) o f ASME Code Case N-729-1 , for the subject welds , would resu lt i n hardsh i p without a c omp e n sa tin g in c r ease in th e l e v el o f qu a lity a n d sa f e ty. Authorization r ead by R o bert P asca relli , Chief of the Plant Li ce n si ng Br a nch IV-1 , Offic e of Nuclear Re a ctor Regulation As Chief of the Plant Licensing Br a nch I V-1 , Office o f Nuclea r R e actor Regulation , I c on c u r with th e Comp o n ent P e rf o r man ce, Non-D es tru ct iv e Ex a minat io n , and T es ting Br a n c h's d e t e rmin a tio ns. Th e NRC sta ff c o nclud es that th e propo se d a lt e rn a tive pr o vid es r e a s onabl e assu ran ce of structural i ntegr i ty of the CROM penetration n ozzles numbers 20, 27 , 35, 40 , 46 , 47, 58 , 59, 63, 70 , 71 a nd 77 s u c h th a t co m p lying with th e A S ME Cod e r e quir e ment w o uld result in hard shi p o r unusual diffi cu lty w i thout a co mpen sa ting in c rea se i n the level of qu a l ity a nd sa f e ty. A cco rd i n g ly , th e NR C s t a ff co n c lud es th a t th e l i ce n see h as ad e quate l y addressed all of the regu l a t ory requ i rements set forth in 1 0 CFR 50.55a(z)(2) a nd 10 CFR 5 0.55a(g)(6)(i i)(D). T h e r e for e , th e NRC s t a ff au t h o ri zes th e u se o f r e li ef request 14 R-03 at the Wolf Creek G e nerat i ng St a tion during the c urrent r e f ue ling o utag e s ub jec t to t h e li ce n see's pr opose d a lt e rn a tiv e th a t i f an u n accep t a b le in d ica tion by th e leak path assessment or volumetr i c exam i s i dentif i ed, the licensee w i ll revert to the r e qu i r e m e nt s of Cod e Case N-729-1 a nd 10 CF R 5 0.55a (g)(6)(i i)(D). A ll o th er r e q ui r e m e nt s of A S ME Co d e , Sect ion X I , f o r whi c h re li e f w as n o t s p ec ifi ca lly r e que s ted a nd a uthori ze d by the NRC s t a ff r e m a in ap pli ca ble , in c luding th e th i rd p a rty review by the Autho r i ze d N u c l ea r In-se rvi ce In spec t o r. This v e rb a l authorization do e s not pr e c l ude the NRC staff from ask i ng addit io na l c l a rifi ca ti o n q u est i o n s r e g a r d in g R e li ef R eq u est 14R-0 3 , wh i l e pr e p a ring th e su b se qu e nt written safety eva l uation. In cons i d era t ion of t he radiologica l dose est im a t es, the N RC notes t h at t hese are conservatively hi gh estimates. The N RC bases this evalua t ion on actual survey dose rates i n the area. Add i tionally, t h e NRC staf f notes, that w i t h prior planning, the effect of this i nspect ion on outage schedule cou l d be s i g n i ficantly r educed. However, the NRC staff did fin d that even a l owe r bound radiologica l dose est i mate was s t ill s u fficient har d ship bas e d on t h e compensat i ng l evel of quality of the vo l umetric examinatio n g i ven no previous cracking had been found previous l y and d u r i ng this outage. I d o ubt tha t detail is n ecessa ry in t h e verbal, but i s included I n this comment in case a q uestion is r ai s ed by the region a s part of o u r r ev i ew conside r ations for this re lief. From: Sent: To:

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From: Co llin s , Jay Collins, Jay 25 Oct 2016 14:53:19 +0000 K a 1 li kian, Roger FW: Wolf Creek -RR 14R-03 (MF8456) Se n t: Thursday, October 20, 20 1 6 4:14 P M To: Lingam , Siva <Siva.Lingam @nrc.gov>; Alley, David <David.Allcy @nrc.gov>; Tsao, John <John.T sao@nrc.gov>; Drak e, Jam es <James.Drake @nrc.gov>; Taylor , Nick <Nick.Taylor @nrc.gov>; Dodson, Dou g la s <Dougla s.Dod son@nrc.gov>; Thomas, Fabian <Fabian.Thomas @nrc.gov>; Proulx, David <David.Proulx@nrc.gov>; Kopriva , Ron <Ron.Kopriva@nrc.gov>; Anchondo, I saac <l saac.Anchondo@nrc.gov >; Cumblidge, Stephen <Stephen.Cumblidge @nrc.gov> Cc: Pa scare lli , Rob ert <Rob e r t.Pa scare lli@nrc.gov> S ubject: RE: Wolf Creek -RR 14R-03 (MF8456) -31 41 Ge n eral -3 1 42.I Acce pt a n ce b y VE (b) A component whose VE d!et ects a relevant conditio s hall be una cceptable for co ntinu e d serv ic e unti l the requirements of -3 14 2.1 (b)( l ), (b )(2), and ( c) below are met. (/) Components with relevant co11ditions require further evaluatio11. This evaluation shall include determination of the so urce of th e l ea ka ge and co rrection of the source of leakage in accorda n ce with -3142.3. (2) A ll r e l eva nt conditions s hall b e eva lu ated to d e t e rmin e the extent, if any , of degradation. The boric acid crysta l s and residue shall be removed to the exten t necessary to a llow adequate examinations and eva l uation of degradation, and a subsequen t VE of t h e previously obscured s urfa ces s h a ll b e p erfo rm ed, prior to return to se rvic e , and again in the s ub seq u e nt: refueling outage. Any degradation dete cte d shall be evaluated to determine if any co rro s ion ha s impacted th e s tru c nira l int egr it y o f the component. Corrosion that has reduced component wall thi ckness below design li mits s h a ll b e resolved through rep air /replacement activity in accordance with IWA-4000. (£.!A no zz l e whose VE indicat es e l evan t con dition s j ndicative of possible no zzle l eakage shall be unacceptable for co ntinued se rvi ce unless it meets the requirements of -3142.2 or -3142.3. -3 1 42.2 Acce p tance by SuJ!P l emcota l .Examina ti on. A no zz le with r elevan t co ndition s indicat i ve o f possible J10zzle l eak ag e shall b e accep t able for continued service if the re s ult s of s uppl e mental examinat i ons [-3200(b)] meet the requirements of -3130. -3 1 42.3 Acce p tance by C orrective Measures or R e pair/Re pl aceme nt Activity (a) A component with r e l eva nt co ndition s not indicativ e of possible no zz le l eakage i s acce pt ab l e for co ntinued service if the source of the re l ev ant cond i t i on is corrected by a repair /r e pl acement activity or by correct iv e m eas u res n ecessary to p r ec lud e d egra dation. (b) A co mpon e nt w i th r e l eva nt co nditi o ns indi cat iv e of pos s ible n ozz l e l eakage s h a ll b e acceptable for co ntinu ed serv ic e if a r ep air/r e plac e m e n t activity corrects th e d efec t in accorda n ce with IW A-4 000. -3 1 40 INSERVICE VISUAL EXAM I NATIONS (VE) -3141 Ge n eral co ndition s for th e purposes of the V E{ s hall inc l ude areas of corrosion, boric acid depo s i ts, disco l oration, and other ev idence of no zz le leakage. The NRC consider s any relevant condition in the annulus region between the no zzle and hea d s urfac e that cannot be r e mov e d by li ght clea n i n g act i vit i es to be a r elevant cond iti on of poss i b l e no zz l e l ea k age. -----O ri g in al F r om: Lingam, Siva Se n t: Thursday , Octobe r 20, 20 1 6 3:5 1 PM To: L in g am , Siva; Alley, David; Collins, Jay; T sao, John; Drake, J a m es; Taylor, Nick; Dod so n , Douglas; Thoma s, Fab i an; Proulx , David; Kop ri va, R o n; Anc h o nd o, I saac; Cumbl id ge, S t e ph en Cc: P ascare lli , Rob ert S ubj ect: Wolf C r eek -RR 14R-03 (MF8456) Wh e n: Thursday , October 20, 20 1 6 4:15PM-5:15 PM (UTC-05:00) Eastern T im e (US & Ca n ada). Where: HQ-OWFN-1 OB06-12p Please not e the following to discuss the s ubj ect RR based o n the a tt ac h ed response from the li censee: Brid ge No.: Passcode: Date: Time: 877-935-1422 l (b)(6) I fo ll owed by # October 20, 2016 (Thursda y) 4: 15 PM (Eas t ern T im e) <<File: E Tl6-0028.pdf >> From: Sent: To:

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Thanks Bob Dean, Bill 25 Oct 2016 09: 18:26 -0400 P as car e lli , Rob e rt Re: Wolf Creek nozzle leakage On: 25 O cto b e r 201 6 0 2: 17, "P asca r e lli , Rob e rt" <Rob e rt.P asca r e lli@nr c.g ov> wrot e: Bill/Michele/Brian, Anne Boland indicated that were in terested in the discussions between the program office , Region IV , and the licensee r egarding the vessel head r epa ir at Wolf Creek. Attached for your information is a one-pager that Reg ion IV updated la te last week that describes the technical specification required shutdown on September 2 , 2016 and subsequent repair activ i t i es. The source of the l eak was determined to be a l eak in g canopy sea l we ld on a core exit thermocouple penetration nozzle above the reactor vessel head. Th e large amount of bor i c acid obscured the visua l inspection of t welve vessel head nozzles as a result of the spray pattern from the lea king 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 conduct in g the su rfa ce 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 confe r ence c alls with the l i censee as well as Region IV. We have been fortunate to have some DRS inspectors and the DRP Branch Ch i ef on-site and they have shared information regarding the licensee's progress on cleaning the head and the ongoing visua l examinations. We a r e awaiting some add ition al info rm at ion from Wol f C r eek that they agreed to provide to the staff during a conference ca ll on Friday. P l ease let me know if you have any questions. Bob Pascarelli Bob Pascarelli, Chief Plant Licensing Branch IV-1 D i vision of Operat i ng Reactor Licensing Office of Nuclear Reactor Regulation From: Sent: To: Ba l want;Lingam, Siva Subj e ct: Alley, David 26 Oct 2016 16:14: 03 -0400 M ate ychick , John;T a ylor, Nick;Proul x, David;Thoma s, Fabian;Singal, 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 i t will take my folks some amount of time t o go through whatever is submitted. I don't re l ish the thought of this continuing to be delayed and then the l icensee wanting a snap decis i on. Whatever can be done to ensure that the l icensee understands this issue wou l d be appreciated. Dave From: Matey c hick, John S e nt: Wednesday, October 26, 2016 4: 11 PM To: Taylor, Nick <Nick.Taylor@nrc.gov>; Prou l x, David <David.Proulx@nrc.gov>; Thomas, F a bian <Fab ia n.Thoma s@nr c.go v>; Singal, Balwant <Ba lwant.Singal@nr c.gov>; Lingam, Siva <Siva.Li n ga m@nrc.g ov>; Alley, David <David.A l ley@nrc.gov >

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RE: Wo l f Cree k Relief Request? No t yet. They had promised to get i nformation to us on Thrusday. John M From: Taylor, Nick Se nt: W ednesday, O cto ber 26, 2016 3: 01 PM To: M a tey chick , John <John.M atey c h i ck@nrc.gov >; Prou l x, D a vid <D a v i d.Prou l x@nr c.g ov>; Thom as, Fabian <Fabian.Thomas@ n rc.gov>; Singal, Balwant <Balwant.S i ngal@n r c.gov>; Lingam, S i va <Siv a.Ling a m@nr c.gov>; Alley, David <D a vid.A ll e y@nr c.gov> Subj e ct: Wo l f Creek R e l ief R eq ue st? All , P retty q ui et o n t h e Wo l f Creek f ront w i th r egar d to t h e h ea d. H ave we hea r d anything f rom the li censee o n the photos , etc r eq u ested in support of the ir re li ef reque st? Ni c k Tay l o r 972-921-6398 From: Greene, Louis Sent: 26 Oct 2016 14:13: 00 -0400 To: WolfCreekEIS Re s ource;Watford, Margaret;S ingal, Balwant;RidsRgn4MailCenter Resource; RidsResDE Resource; RidsNrrPMWolfCreek Resource ;RidsNrrDorllpl4-2 Resource;RidsManager Resource;Regner, Lisa;Pascare l li , Robert;Lyon, Fred;Lingam, Siva;Burkhardt, Janet

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Wolf Creek Response to Request for Additional Infor mation re Relief Request Number 14R-03 for Reactor Vessel Head Penetration No zzle Welds. Attachments: 8project_ Tem pFil es_ eRids _ erid66695565805 767314 70. 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 630 0 A2 l 4 Numb e r Title Wolf Creek Re s pon se to R eq ue st for Additional I nformation re Relief R e que st Number I4R-03 for Reactor V esse l Head Penetration Nozz l e Weld s. Docket 050004 82 Number Docum e nt 10 1 20 1 20 1 6 Date Au thor McCoy J H Name Author Wolf Creek Nuclear Operating Corp A ffiliation Addressee Name Addressee NRC/Docum e nt Co nt ro l De sk Affiliat i on NRC/NRR Document Letter Type Response to Request for Additional In formation (RAI) Availabi li ty Pub licl y Avai l ab l e Dat e to b e 11/04/2016 Relea se d Document Non-Sens i tive Sensitivity Comment Date Added l 0/26/20 1 6 Keyword daw5 DPCautoadd

• rr3 ADAMS Distribution Sheet Priority:

Nor mal From : Greene, Louis Ass i2n e d Recipient s: Watford , Margaret 0 OK Singa l Balwant 0 OK Rid sNr rPMW o lfCrc ek R eso urc e 0 OK RidsNrrDorlLpl4 Resource 0 OK R eg n e r Li sa 0 OK P ascare lli , R obert 0 OK L yo n, Fred 0 OK Lin ga m , Siva 0 OK Burkhardt , Janet 0 OK Int e rnal Recipient s: zzzF TL E CEN TER 01 0 Not Found WolfCreek Resource 0 OK R id s R g n4M a i 1 Ce nt e r 0 OK RidsResDE Resource 0 OK Rid s M a nag er Resource 0 OK Total Co pies: 1 Distribution Co de s Use d I A047 I OR Submittal: In service!festing /Relie ffr om ASME Code; related corre s pondence Docum e nt Profile Acc ess ion Number ML l 630 0 A214 Ti tl e W o l f C re ek R espo n se to Request for Additional in forma ti o n re R el i ef R eq u es t Number J 4R-03 for R eac t o r Vessel He a d P e n e t rat i o n Nozzle W e ld s. Do c k et Number 05000482 D oc um ent D ate 10 120/2 01 6 A uth or Name M cCoy J H Author Affiliation W o l f C re e k Nucl ea r Operating Co rp Addr essee Name Addressee Affi li ation NRC/Do c um ent Con tro l D es k NRC/NRR Do c um e nt T y p e L etter Re s pon se to R e qu es t for Additional In formation (RA!) Availability Publi c l y Avai l ab l e D ate to b e R e lea se d 11/04/2016 D oc um en t Se n s iti v ity Non-Sensitive Comme nt Date Added 10/26/2016 Keyword dawS DPCautoadd jrr J From: To: Sub je ct: Date: Balwant , Al l ey, Dav i d S j oga l B a l wa ot Wolf creek Thu r sday, Octobe r 27, 2016 8:00:00 P M I fa ile d in my dut ies. The licensing guy from Wolf Creek left me a voice mail this PM. I managed to listen to it but not respond. I l eft his name and number in t he office. His quest i on was about whether we needed a ll nozz l es or only the nozzles other than t hose in cluded in the rel i ef request. My answer is that if they have all the noz zles it would be benefic i al for us to see them. Th e real need however , is for t h e other nozzles. The plant has already acknowledged that the 12 nozz l es in the relief req ue st require further in spect io n (and i t is my understanding that they have inspected as proposed in the relief request). The real need at this point i s to determine whether add i tio na l nozzles r equire further inspection and whether there is l ess r eason to accept the concept that all the boric acid came from above (w hi ch impacts whether the r el i ef is authorized). Th e plant needs to make s ure th at they are fully in comp l iance with th e code case and ultimately with other applicable programs such as the boric ac i d program. As a result after the f inal clean i ng , the region w i ll probably be i nterested in the condition of all the nozzles , not ju s t the " other" nozzles Dave Dav i d A l ley PhD. Chief, Component P e r forma n ce NDE and Test ing Bra n c h U S Nucl ear Regu l atory Commiss i on 11555 Roc kv il l e P ik e Roc kv il l e MD 20852 301-415-2178 From: To: Sub je ct: Date: Al l ey, Dav i d Farna n Mi cha el For tomorrow Thu r sday, Octobe r 27, 20 16 8:06:00 P M 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. Conceivab l y we could get as far as issu ing a verbal. If so , you can read my part. Stephen has the script. Stephen will be around a ll day tomorrow a nd Roger will be in in the morning. Bo th 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 Dav i d A l ley PhD. Chie f , Component Perfo r mance NOE and Testing Branc h US N uclear Regu l ato ry Commiss i on 11555 Rockvil l e Pik e Rockvil l e MD 20852 301-41 5-2178 From: Sent: To: C c: Subj e ct: Taylor , N i ck 27 Oct 2016 13:44:20 -0500 All e y, David; Drak e, Jam es;Melfi, Jim Proulx, David;Werner, Greg;Anchondo, lsaac;Co ll ins, Jay; Kopriva , Ron RE: Wolf Creek Head I think Davea*Žs question is a good one and should probably be answered before we decide to have a call. Nick From: Alley, David S e nt: Thursday, October 27, 2016 11:09 AM To: Drake, James; Melf i , Jim C c: Taylor, Ni ck; Proulx, David ; Werner, Greg; Anchondo , I s aac; Collins, Jay; Kopr i va, Ron

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RE: Wo l f Creek Head Jim, Does this mean that t h ey are going to submit info to us to review in sufficie n t time that we can review it before the call? No t sure wha t a ca ll will accomplish until we have looked at t h ei r findings. Dave From: Drake, James S e nt: Thur sda y, October 27, 2016 11:55 AM T o: M e lfi, Jim <J i m.Me l fi@nrc.gov > Cc: Taylor, N ick <Nick.Taylor@nrc .gov>; Proulx , David <Dav i d.Prou l x@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov >; Anchondo, I saac <l saac.Anchondo@nrc.gov >; Alley, D a vid <David.Alley@nrc .gov>; Collin s , Jay <Jay.Collins@nrc .gov>; Kopriva, Ron <Ron.Kopriva@nrc.gov >

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Wo l f Creek H ead Jim , Wo l f Creek j ust contac t ed me. Th ey wan t to have a call tomo r row at 1 000 t o d i scuss t he head rel i ef req u est. They are going to call S i va to set up a br i dge. Jim $m n e.r <.f. :Dra ke J ames F. D ra k e Off i ce phone: 8 1 7-2 00-1558 Ce ll P hone j<b)(6) I Fro m: To:

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Date: A ll ey , David Wl.aWix R E: FW: Post-c l ean i ng pictu r es of reactor head up l oaded t o C E R T REC Fri day, October 28, 20 1 6 9: 22:0 0 At.! Current plan is a call at 3PM eastern. Dave From: Collins , Jay Sent: Fri day, O c tob e r 28, 2016 8:00 A M To: Alley , Davi d <Dav i d.A lle y@nrc.gov>

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R E: FW: Pos t-clean i ng p i c tur es of reactor head up l oa d e d to CERTREC Greet i ngs, I looked th i s mornin g and do not see them , therefore I guess the 10am call i s not happen i ng. If you need me to l ook a t some t hi n g, gi v e me a cal l , bu t I wi ll b e in tra n s it fr o m -9am t o -11a m. J ay From: Dra ke, Ja me s Sent: T hur sda y, Oc t ober 27, 2016 3:25 PM To: T a ylo r , Nic k <Ni ck T ay l o r@nrc goy>; Pr oulx, Dav id <D ayj d pro u lx@nrc goy> Cc: Wern e r, Greg <Greg Werner@nrc gov>: Anc h ondo , I saac <Isaac Anchondo@nrc gov>; Kopr i va , Ron <Ron Kopr i va@nrc gov>; A ll e y, D a v id <Qayjd A l l ey@nrc gay>; Co ll ins , J a y <Jay Col l j ns@nrc goy>; Cumb li d ge, Stephen <Stephen Cumb l j dge@nrc eoy>; M e lfi , Jim <Ji m Melti@orc gov>

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R E: FW: Pos t-clea n i ng p i ctures of reactor h ead up l oa d ed to CERTREC Nick , The pictures that Wolf Creek has l oaded to Certrec now are of the f l ange area. They won't help w i th the relief request. Reece said he doesn't think those pictures will be available unt i l l ate tonight or tomorrow. The 1000 call for the re li ef request is contingent on them prov id ing the pictures with sufficient time for NRC personnel to review them. Jim From: Tay l or, Nick Sent: T hur sda y, Octobe r 27, 2016 1:45 PM To: Prou l x, D a v i d <Oayjd Proulx@nrc 1,:oy> Cc: Drake, James <James Orake@nrc goy>

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R E: FW: Pos t-clea n i ng p i ctures of reactor head up l oa d ed to CERTREC I will try to d i a l i n tomorrow morn i ng but right now I have a meet in g scheduled ons i te during that t i me. David , can you represent the branch on this cal l in case I am u n ab l e to join? Thanks , Nick From: Dra k e, J ames Se nt: Th u rsday, October 27, 2016 11: 16 AM To: Alley, David <D a vid Alley@orc gov>: Coll i ns, J ay <J ay Col!ins@nrc gov>: Cu mbl idge , Stephen <Stephen C u mbljdge@nrc gov> Cc: T aylo r , N i ck <N i ck Taylor@n r c gov>; An c hondo, I saac ; W erner, G r eg <Greg Werner@nrc goy>; Kop ri va, Ron <Ron Kopr j ya@orc goy>; Li ng a m, Siva ; Prou l x, Dav i d <Oay j d prou l x@or c gay>

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FW: F W: P ost-clean i ng p ict ure s of r ea cto r hea d upload e d to CERTREC I think I got everyone. The pictures have been uploaded to item 12 in the 402016 Integrated Inspection fo l der on CERTREC. If you do not have access , let me know and I wi ll try to get you added to the list. Jim Fr o m: Hobby Reece D [m ajl t o*reho bb y@W C N O C com] Sent: T h u rsday, October 27, 2016 11:1 4 AM T o: D r ake, Ja m es <J ames Prake@or c goy> Sub j ect: [External_Sende r] FW: Post-d eaning pictures of r eacto r head uploaded to CERTREC From: Hobby Reece D Se nt: Th ur s d ay , O c to b e r 27 , 20 1 6 7:33 AM T o: Th omas Fab i an D; Dodson Do u g l as E; 'KOP R I V A, R o n A' Cc: Vi cker y Bra d J; Barrac l o u gh R ic h a rd M; Goo d N i co l e R; Stone L u cille M; Mu i lenburg W ill i am T; H a f enst i ne Cy n t h ia R

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P os t-c l ean in g p ict ur es o f reacto r hea d up l oa d ed to C ERTREC Fab i an, Doug and Ron: The most-r ecent pictures ta k e n after the reactor hea d was cleaned have been u p loaded to i tem 12 i n the 4 02016 I n t egrated I nspect i on folder on CERTREC in accor d ance w i th y o ur reques t. F ina l clea n i ng of the reactor head is c u rren t l y sc h eduled for 16 00 on Octo b er 30, 2016 but that s c h edu le cou l d cha n ge based on t h e p rog r ess of work in t h e n ext few days. We w ill n o t ify t he residen t i n s pector s about the sche du le for t he fina l cle an i ng. Reece From: C u mblidge, Stephen Sent: 28 Oct 2016 13:42:47 -0400 To: A ll ey, David;Coll i n s, Jay;S inga l , Balwant;Kalikian, Rog e r;T sa o, Jo h n; Drake, James;Tay l or, N i ck; Proulx, Dav i d; Regner, Lisa; Werner, Greg;Anchondo, lsaac;Kopriva, Ron;T homas, Fabian Subje c t: R E: I nternal NRC C a ll to Di scuss Wolf Creek R e lief Request I agree. T h e lan g u age d oes n ot appear to b e consis t e nt with t h e co d e case , th e ord e r , o r E PRI g u i d ance. Stephen Cu mblid ge M ater i a ls E ngineer US Nuclear R eg ulatory Co mm iss ion M a il S t o p OWFN/9 H 6 Wa s hin g ton , DC 20555-000 1 Telephone: (3 01) 415-28 23 (Office) F rom: Alley , D avi d Se nt: F rid ay, O ctober 28, 20 1 6 1:36 PM To: Co llin s, Jay <Jay.Co llin s@nr c.gov>; Si n ga l , Ba l want <Ba l want.S in ga l@nr c.gov>; Kalikian , Roger <R oger.Ka lik ian@nrc.gov>; Tsao, John <John.Tsao@nrc.gov>; Drake , Jam es <James.Drake @nrc.gov>; Tay l or , N i ck <Nick.Tay l or@nrc.gov>; Proulx , David <Dav id.Pro ul x@nr c.gov>; C umblid ge , S t e ph e n ; R egner, L i sa <Lisa.R eg n er@n rc.go v>; W erne r , Greg <Gr eg.Wern e r@nrc.gov>; Anc h ondo, I sa ac <lsaa c.A n c hondo@nr c.gov>; Kopriva , Ron <R o n.K o pri va@nr c.gov>; T h o m as, Fab ian <Fab i a n.Thomas@n rc.gov> S ubj e ct: RE: Int e rn a l N R C C all to Discuss Wo l f C reek R e li ef Request Folks, I have been through the sp r eadsheet, the eva luation document and many but not all the photos. I h aven't seen any thin g in the photos that would convince me that on l y the 12 nozzles had relevant conditions of potential nozzle l eakage (or whatever the precise words arc). I also didn't sec anything that confirms that all the relevant conditions have been removed IA W gu idance on th e subjec t. J n the spreadsheet and the evaluation document I found th e following: Nothing emanat in g from the annu lus region was co nfir med. (fro m spreadsheet and p age 4 of 5 of eva lu a ti on document) Po r t i ons of th e annulus cou ld b e observed w i t h out remov in g a ll res idu e and it was d e t e rmin e d that this wa s th e ex t e n t n ecessa t y to a llo w adequate examination (p age 2 of 5 of eva lu atio n do c um ent) Becau s e an adequate exam in at i o n wa s performed s h ow in g that non e of the r ema inin g popu l a ti o n of 66 n ozzles had nozzle l eakage , i t was not necessa 1 y to com pl e t e l y r emove the acc u m u l ation p r ese n t on many n o zz l es t o s at i s fy t he o bjective of dete rminin g th e ab se nc e of n o zz l e leakag e from a crack in th e no zz l e or j groove we ld. (page 3 of 5 of t h e eva lu at i on document) These statements appear to me to indicate that the licensee's act i ons arc no t co n sistent with the code case -I believe that the s tand ard is possible leaka ge not confirmed leakage Thoughts? Dave From: Co lli n s, Jay Se nt: Friday, O ctober 28, 2016 11 :34 A M To: Singal , Balwant <I3alwant.S i nga l@nr c.gov>; All ey, David <Dav i d.A ll cy@nrc.gov>; K a liki a n , R oger <R o g c r.K a l ik i a n C dl nr c.g o v>; T sa o , John <Joh n.Tsao@n r c.gov>; Drak e, Jam es <J amcs.D ra k c@nr c.gov>; Taylor , Nick <N i c k.Taylor@nrc.gov>; Proulx, D avi d <Davi cl.P r ou l x@n r c.gov>; Cu mblid ge, Ste ph en <Stephen.C u mb li dge@nrc.g o v>; Re g ner, Lisa <Lisa.R eg n e r@n rc.go v>; Werner , Gre g <Greg.W emer@nrc.gov>; Anchondo, I saac ; K o priv a, Ron <R on.Ko pr iva@nr c.gov>; Thoma s, Fabian <F ab i an.Thoma s@nrc.gov> S ubject: RE: Internal NRC Ca ll to Discuss Wolf C reek Relie f Request ASME Code Ca s e N-72 9-1 Note I The VE s hall consist of the following: (a) A direct examination of the bare-metal su r face of the en t ire outer surface of t h e head, including essentially I 00% of the int ersection of each no zz l e with th e head. lf we l ded or bo lt ed obstructions are present (i.e., mirror in sulation, insulation support feet, shroud s upp o rt ring/lug), the examination sha ll include 2::95% of the area in the region of the n ozz l es as defined in Fig. I and the head su rface uphill and downhill of any suc h obstructions. The examination may be performed with insulation in place using remote equipment that provides resolution of the component metal surface equ i valent to a bare-metal direct exa m inat i on. (b) The examination may be perforn1cd with the system dcpressurizcd. (c) The examination shall be performed with an illum i nation level and a s ufficient distance to allow res o lution o f lo we r cas e charncters n ot greater than 0.105 in. (2. 7 mm) in height. -----Original F rom: Si n gal, Balwant Sent: Friday, O ctober 28, 2 016 10: 1 7 AM To: Co llin s, Jay; A ll e y , David; Kalikian , Roger; Tsao, J ohn; Drak e, J a m es; Tay l or, N i ck; Proul x, D av id; Cumb li dge , Stephen; R eg n e r , Lisa; Wern er, Gr eg; Anchondo , f saac; Kopriva , R on; Thom as, Fabian S ubj ect: I n t e rnal NRC Call t o Di s cuss Wolf Creek R e lief Reque s t When: Fr id ay, O c tob e r 28, 20 1 6 11 : 00 AM-12: 00 PM (UT C-05 :00) Eas t e rn Tim e (US & Ca n a da). Where: Dave Alley's Offic e Dav e A ll e y and m e rec e iv e d a ca ll from Wolf C re e k (C yndia a n d Jaimm e Mc Co y) at 9.30 thi s mornin g. An int e rn a l NRC s t aff m eeti n g i s r eq uir ed lo di sc u ss path forward ba se d on inform a t ion pr ovide d durin g the ca ll. Bridge No. Info. 866-6 2 4-3 402 Pa ssco d c: l (b)(6) Lisa: You will n ee d to u se P asscode F b)(6) I (as ini t i ator of th e ca ll). I was not bale to se ar c h fo r co n fe r ence r ooms from h ome. I wi ll be out-o f-office j (b)(6) t for abo ut 3 h o ur s and Li s da wi ll be suppor tin g thi s ca ll. l can b e con t acte d .... (b-)(-6) ___ _.I fo r any questions. Thank s. From: To: Cc:

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Date: Cymbljdge. Stephen Lubi n s k i !oho* Evans Miche l e* McDe r m ott Br i an A l l ey. Day j d Wo lf Creek P h one Call Friday, October 28 , 2 0 16 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 i f the l eakage 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 li censee 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 bor i c acid deposits from t he canopy seal weld leak on Nozzle 77 , all nozz l es were categorized as having relevant conditions." IAW Paragraph -3 142.1, this statement requires the licensee to perform a VE exam on all nozzles to determine if there is a relevant condition indicative of possib le 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 nozz l e w i th the head. The second critical statement from the li censee's report was , " Because an adequa t e exa mination 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 l eakage from a crack in the nozz l e or j groove weld." This statement is incongruent with a VE exam. After a lengthy discussion about the requirements of N-729-1 , the li censee appeared to understand the NRC's posit ion. 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 u lt rasonic examinations of the affected nozz l es or to submit a relief request to defer the in spections. The licensee requested a later phone ca ll with the NRC to discuss which path they will take to resolve this situation. Stephen Cumb l idge Materials Engineer US Nuclear Regulatory Commission Mail Stop OWFN/9 H 6 Washington , DC 20555-0001 Telephone: (301) 415-2823 (Office) From: Regner, Lisa Sent: 28 Oct 2016 13:49:14 -0400 Singal , Balwant To: 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 exist i ng RR is adequate.

Discussion from call today at 11 am ET:

• the region has seen i ndications of boric acid and wastage on nozzles other than the 12 nozzles in the RR
• the licensee s tated in the existing relief request that there are other ind i cations out s ide of the 12
• WC states that they have followed the Code Case N-729, but have not, and apparently can not , provided documentat i on showing this
• WC ha s retained the equipment necessary to do additional volumetri c leak path testing
• acceptance cr i teria i s to do a 360 degree visual examination unless there is physical limitat i on preventing noth i ng larger than aspirin-si z ed 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 u n til the head i s cle a ned and the licensee provides the results of visual examinat i ons
• the code requires enough removal of debris to be able to make a determinat i on of leakage; the staff does not bel i eve WC has done this
• region stated WC has scheduled to clean the head on Sunday Lisa R e gn e r Sr. PM NRR/DORL/LPL4

-1 301-41 5-1906 08 0 08 From: Sent: To: Cc:

Subject:

Penetrations All, Singal, Balwant 28 Oct 2016 23:48:17 +0000 Pascarelli, Robert; Boland, Anne;Evans, Michele; Benner, Eric;Lubinski, John Lingam, Siva;Regner, Lisa;Alley, David;Taylor, Nick Wolf Creek Relief Request -Control Rod Drive Mechanism (CROM) Nozzle There were multiple calls internally between th e 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 l icensee that based on the available information, the NRC staff does not believe that the licensee is in compliance with the Code Case N-72 9-1for66 nozzles (in addition to 12 a lr eady covered by the relief request). The licensee requested for another call l ater during the day after considering the staff input and eva lu ate a va ilable options. There was another call at 6.30 PM participated by severa l members of Wolf Creek Nuclear Operating Corporation, N ick T aylor from R egion IV, Dave Alley from DE , a nd Balwant S in ga l 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 hards h i p (dose, schedule impact, and cost) and provide justification based on what has already been done. The N RC staff (Dave Alley) informed the licen see this option is not likely to have sufficient justification and does not seem to have a successfu l path. OPTION 2 The licensee proposed to perform additiona l exam ina tio n from underneath the head similar to the one's proposed for previous 12 penetration s and us the results of the examinations to ju stify all rema i n in g 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 a l so in th e scope of the re lief 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 i s not suggesting any particular option and it is up to the licensee t o pick path forward based on th eir best judgement. The call ended with this discussion. The licensee will inform the NRC staff about their decision after internal discussion. It ap p ears that the licensee is conv inced 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 Ba l wa nt.Singal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 .... From: Sent: To: Subje ct: 1 pm CDT From: Taylor, Nick Werner, Greg 3 Oct 2016 14:32:45 -0500 Alley, David RE: Internal commun i cations at Wolf Creek re head corrosion I p l ans S e nt: Monday, October 03, 2016 2:32 PM T o: Werner, Greg <Greg.Werner@nrc.gov>; A ll ey, David <David.Alley@nrc.gov>

Subject:

RE: I nte rn al communications at Wolf Creek re head corrosion I plans All , We are a go for 1 :00 pm tomorrow. We are getting a co nf erence b r idge set up and wi ll send out a n appoi n tme n t notice. I as k ed Wolf C r ee k t o l oad up a n y ava i lab l e i mages , evaluatio n s , e t c before the ca ll. N o t s u re if we will get anything .... More to follow, N i c k From: Werner, Greg Sent: Monday, October 03, 2016 1:34 PM To: Alley, David <Dav i d.Alley@nrc .gov> Cc: Taylor, Nick <Nick.Tay lo r@n r c.gov> Subj ec t: RE: I nternal communications at Wolf Creek re head corrosion I plans Y es. I w i ll l et N i ck T aylor know to add you to the appointment and prov i de any deta il s we might get before then. Greg From: Alley, David Sent: Monday, October 03, 2016 1:18 PM To: Werner, Greg <Greg.Werner@nrc .gov> Subj ec t: FW: I nternal communications at Wo l f Creek re head corros i on I plans Greg We wou l d like to be on the call tomorrow. Dave From: Tsao, Joh n S e nt: Monday, October 03, 2016 1: 29 PM To: Alley, David <Dav i d.Alley@n r c.gov>

Subject:

RE: I nte rn al communications at Wolf Creek re head corrosion I plans Dave, Yes we sho u l d be on t he ca ll w i th Wolf Creek tomor r ow From: Alley, David Sent: Monday, October 03, 2016 1:21 PM To: T sao, John <John.Tsao@nrc.gov >

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FW: Internal communications at Wolf Creek re head corrosion I plan s John Please take a look at this Greg Just tried to call -no answe r. 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 <Dav i d.Alley@nrc.gov> Cc: Tay lo r, N ick <Nick.Taylor@nrc.gov >

Subject:

FW: I nternal communications at Wolf Creek r e head corrosion I plan s FY I. 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 interna l 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 , informat ion o n potential relief requests, and how they selected the other 4 penetrations for the clamps. Greg Werner From: Taylor, N ick Sent: Monday , October 0 3, 2016 11:50 AM To: W e rn e r , Greg <Greg.Werner@nrc .gov>; Kopriva, Ron <Ron.Kopriva@nrc .gov> Cc: Pruett, Troy <Troy.Pruett@nrc .gov>; V egel, Anton <Anton.Vegel@nrc.gov >; L antz, Ry an <Ryan.Lantz@nrc .gov>; Clark , Jeff <Jeff.C l ark@nrc.gov >; Prou lx , David <David.Prou l x@nrc.gov>; Janicki, Steven <Steven.Janicki@nrc .gov>

Subject:

I nterna l comm un ications at Wolf Creek re head corrosio n I plans All , I'm still working on se tting up a ca ll 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: To: S ubje c t: Date: Atta c hm e nt s: Good news ... Lisa R egner S r. PM Reimer. Lisa Pas car e l li R ob ert PN: Wo l f C ree k h ead i n s p ec t io n upd a te 10-3 1-201 6 M on da y, Oc t obe r 3 1 , 20 16 4:2 0:0 0 P M i mage00 1.o n g N RR/DOR L/LP L 4-1 301-415-1906 08D08 From: Tay l or, Ni c k Sent: Monday , O c t ob er 3 1 , 2 0 16 4:1 8 PM To: Prue t t, Troy <T roy.Pr u et t@nrc.go v>; La n tz, Ryan <Rya n.L a ntz@nr c.gov>; C l a rk , Jeff <J eff.Clark@

nr c.g o v>; V eg e l, A n ton <An t on.V e g e l@nr c.gov>; K e nn edy , K r i ss <K ri ss.Kenne d y@nrc.go v>; Mo rris , S cott <Scott.M or ris@nrc.g ov>; Alle y , D a v id <D av i d.A l l e y@nr c.gov>; W erne r , Gre g <G r e g.W ern e r@nr c.go v>; Si nga l , Ba l wa n t <Bal w an t.Singa l@n rc.go v>; Li ngam , S iv a <Siva.Li ngam@n rc.gov>; R e g n er, Li s a

Cc: Dod so n , Dou glas <D ou gl a s.D o dso n@n r c.g ov>; Thom as, F a bia n <Fabian.Tho m a s@nr c.g o v>; Proulx, Dav i d <D av id.P ro u l x@nrc.g ov>

Subject:

W o l f C r eek h ead inspec t i o n u pda t e 10-3 1-2 01 6 All , By way of updates ... I spoke a few minute s ago w i th Fabian Thomas (re si dent@ Wolf Creek), who observed that the licensee i s now making p l ans and mob i lizing equipment to perform ultrasonic inspect i ons of the remaining 66 nozzles on the reactor vessel head. Fabian reports that the l i censee'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 visua l inspection of the vessel head on November 5. We are still expecting the licensee to need rel i ef from the surface examinations of the groove w e ld s for th e 66 addition a l pen e trations. Th e lic e ns ee ha s not yet communicated the i r plans for requesting this add i tional relief (or anything else for that matter). Thanks , Nick Taylor Chief , Projects Bran c h B Division of R e actor Proj e cts USNRC Region I V 0: 817 200-1 1 41 C: (b)(6) E: ojck.t ay l or@nrc.gov 4 l'tHnt*-t fttt/14...J ,N/_,,,.,,.._.,, From: Sent: To: Cc:

Subject:

Pascarelli, Robert 1Nov2016 12:17:49 +0000 Wilson, George Boland, Anne 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.Co llin s@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 in spection 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.Cl a rk@nrc.gov >; Vegel, Anton <Anton.Vegel@nrc.gov >; Kennedy, Kriss <Kriss.Kennedy@nrc .gov>; Morris, Scott ; Alley, David <David.All e y@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 <Dougla s.Dod s on@nrc.gov>; Thomas, Fabian <Fabian.T homas@nrc.gov>; Prou lx , David <David.Proulx@nr c.gov>

Subject:

Wolf Creek head inspection u pdate 10-31-2016 All , By way of updates ... ! spoke a few minutes ago with Fabian Th omas (resident@ Wolf Creek), who observed that the licen see is now making plans and mobilizing equipment to perform ultrasonic insp ect ions of the remaining 66 nozzles on the reactor vesse l head. Fabian r epo rts that the licensee's schedu l e shows thi s activity co mm encing on November 4. We don't yet know h ow long the activity will take. F ab i an a l so reports that the licensee i s schedu l ed to begin their bare metal v i sua l inspect ion of th e vessel head on November 5. We are still expecting the licensee to need relief from the surface examinations o f the j-groove welds for the 66 additional penetrations. The licensee h as not yet communicated their plans for reque st ing this additional relief (or anything else for that matter). Thank s, Nick Tayl or Chief , Projects Branch B Division of Reactor Projects USNRC Region IV R From: Sent: To: Alley, David 3 N o v 2016 20:44:18 -0400 Pascarelli, Robert Cc: Singal , Balwant;Collins, Jay;Cumb lidge, Stephen;Caponiti, Kathleen

Subject:

Request for additional Information, Wolf Creek, 14R-03 and 14R-03, Nozzle exam Attachments: Wolf Creek RAI R ev 2.docx , Wolf Creek verbal a uth 14 R-03 11-03-2016.docx, Wolf Creek verb a l 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 ob j ection 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 act i ng and will do my part. Dave David Alley PhD. Chief, Component Performance NOE and Testing Branch US Nucl ea r Regulatory Commi ssion 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 GENERA TING STATION WOLF CREEK NUCLEAR OPERATING CORPORATION DOCKET NUMBER 50-482 By letter dated November 1, 2016, Wolf Creek Nuclear Operating Corporation (the licen see) 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 unacceptab l e indic ation by the leak pa th assessment or volumetric exam i s identified, the l icensee 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 requ i rements of 10 CFR 50.55a(z)(2), such that compliance with the specified r equ irement s would r esu lt in hardship or unusual difficulty without a compensating increase in the level of quality and safety. T o complete it s review, t h e Nuclear Regulatory Commission (NRC) requests the following additional information. 1. On page 4 of 8 of the request , the li censee states that th e radiological dose estimated for the eddy current surface examination of 66 of the penetration welds would be 500 mRem. What is the total es timat ed 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 GENERA TING 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 a t 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 exa m is identified, the li censee will revert to th e requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made th is request i n accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the speci fied requirements would result in hardship o r unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that while the demonstrated volumetric le ak path is not equivalent to a fully qualified surface leak path assessment, the licensee i dentified sufficient operational experience, technical basis and radiological dose hardship to s how that r eg ulatory compliance would result in hardship without a co mpensating 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 th at 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 l ikely to see subsequent and more sign ificant crack in g in the future. Given the lack of the initial cracking being id entified 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 les s likely. For technical basis , the licensee identified that their inspect ion would be in compliance with the Wesdye Technical Justification Document s howing an effec ti ve 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 progra m under 10 CFR 50.55a(g)(6)(ii)(D). Th e licensee also referenced NUREG/CR-7 142 , Ultrasonic Phas ed Array Assessment of the Interf e rence Fit and Leak Path of the North Anna Uni t 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation , which found, in part, th e u se of a properly focused 0 degree probe cou ld detect a le akage 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

the effectiveness of the volumetric leak path exam i nation to detect low leakage rates , as performed in accordance with the licensee's proposed alternative. , Co m me n t ICJI: Optional addition a l 'j t ec h nica l b asis fo r re l ie f. I ncl u de i f d es ir e d. For hardship, the licensee noted that a qualified surface leak path assessment cou l d be j performed in two manners that would require both addit i onal radiological dose and time : 'I versus the performance of a volumetric leak path assessment. 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 l*------------------------------------------------------------------------------------------- j Therefore , the NRC staff f i nds that the l i censee's proposed alternative provides reasonable assurance of structural integrity unti l the next scheduled examination, and that compliance with the surface examinat i on requ i rements of Paragraph -3200{b) of ASME Code Case N-729-1, for the subject welds, would resu l t in hardship without a compensating i ncrease in the level of qual i ty and safety. Authoriz a tion r ead by Ro bert P ascar elli , Ch ie f of th e Pl a nt Li c en s ing B ra n c h IV-1 , Offic e o f Nu c l ea r Re a ctor Regul a tion As Chief of the Plant Licensing Branch I V-1 , Off i ce of Nuclear Reactor Regulation, I concur with the Component Performance , Non-Destruct i ve Examinat i on , and Testing Branch's determinations. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CRO M 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 w i thout a compensating increase in the level of qua l ity and safety. Accordingly , the NRC staff concludes that the l i c ensee has adequately addressed all of the regu l a t ory requ i rements set forth in 1 0 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(i i)(D). Ther e fore, the NRC staff authorizes the use of relief request 14R-03 at the Wolf Creek Genera ti ng S t ation during the current refuel i ng outage subject to the l i censee's proposed alternative that i f an unacceptable indication by the leak path assessment or volumetric exam i s identified , t he licensee w i ll revert to the requirements of C o de Case N-729-1 and 10 CFR 50.55a(g)(6)(i i)(D). A l l other requirements of ASME Code , Section X I , for which relief was not specif i cally requested and authorized by the NRC staff remain applicable, including the third party review by the Author i zed Nuclear In-service Inspector. This verbal authorization does not prec l ude the NRC s t aff from asking additional c l arification quest i ons regarding Relief Request 14R-03 , while preparing the subsequent written safety eva l uation. V ERB A L AU TH O RI ZA TI ON B Y THE O FFI CE O F N U CL E A R RE AC T O R RE G U LA TI O N R E LI EF R EQ U ES T 1 4R-04 A LT ER N ATE EXAM IN AT I ON OF CO NT RO L RO D D R I VE M EC H AN I S M N OZZ LE PENET RA TI O N S WO L F CREEK GENERA TIN G STA TI O N WOL F CREEK NU CLEA R OPE R A TI NG COR P O R A T IO N DOCK ET N U MB ER 50-4 82 Technic a l Evaluation read by David Alley , Chief of the Component Performance , Non-Destructive Examination , and Testing Branch , Office of Nuclear Reactor Regulation By l etter dated Octobe r 1 1, 2016, Wolf Creek Nuclea r Ope r at i ng Co rp ora t ion (t he li censee) s u bm i tted Re l ie f Reques t 1 4R-04 for the al t ernate examination of con tr o l rod drive m echan i sm (C ROM) nozzle penetra t i on n u m bers 77 and 78 at t he Wo lf Creek Generating Sta t ion. T he l i censee p r oposed (a) an a lt ernate exam i nat ion d i stance for CR OM nozz l e numbers 77 a n d 78 in l ieu of the requi r ed examinat i o n distance p er AS M E Code Case N-729-1 as condit i oned by 1 0 CFR 50.55a(g)(6)(ii)(D) and (b) not to pe rf orm the surface exa mi nat i on o f the po rti on of the CR OM nozz l e be l ow t he J-groove weld as requ i red by 10 CFR 50.55a(g)(6)(i i)(D)(3). T he N RC staff fin ds that t he p r oposed exa min a ti on dis t ance a re accepta bl e fo r C ROM nozzle nu m bers 77 and 78. T his is based on the validi t y of the licensee's stress analys i s and fr acture mec h a ni cs calcu l a t ion , de m onstra ti ng tha t within four refueling cycles, a pote n tial fla w th at in iti a t es in t h e un examine d zo n e (b elow t h e J-groove we ld) of t he CROM nozz l e numbers 77 and 78 w ill not propagate i n to the J-g r oove we l d. A t th e end of every fo u rth ref u e li ng cycle , t he l ice n see w ill pe rf o r m an exam i na ti on to confirm th e str u c tu ra l i n teg r ity of C RO M nozz l es 77 a n d 78. Th e NR C staff fin ds th e li ce n see's h a rd s hi p j us ti fica ti on i s accep t ab l e beca u se o f th e co n s id e r a bl e r a di at i o n dose a n d th e n ozz l e co n fig ur at i o n th a t a r e n ot co nd uc i ve fo r t h e r eq uir ed exa m i n a ti o n. Th e NR C staff fi n ds that t h e li censee's proposed a l te rn ative exa m ination di s t a n ces for CR O M penetra t ion nozz l e numbers 77 a n d 78 prov i des r ea s onab l e ass u rance of structura l i nteg r ity and leak t ig h tness un t i l t he next sc h edu l ed exa m ination , and that comp li ance with the s u rface exa min a ti o n requ ir e m e n ts o f 1 0 CFR 50.55a(g)(6)(ii)(D)(3) wou l d result i n hardsh ip w it ho ut a c ompensat in g i ncrease i n the l evel o f q u al i ty and safety. Authorization read by Rob e rt Pasc a r e lli , Chief of the Plant Licensing Branch IV-1 , Office of Nuclear Reactor Regulation A s Chief of the Pl ant Li ce n s in g B ran c h I V-1 , Offi c e of N u c l ea r R ea c t or Regulation, I co n cu r w i t h the Compo n ent Perfo r mance, N on-Des tr uc ti ve E xa m ination , and Tes t i ng Br a n c h's d ete r mi n ations. The NRC staff concludes that the proposed alternative prov i des reasonable assurance of structura l integrity of t he CRDM penetrat i on nozzles numbe rs 77 and 78 and that complying with the ASME Code r equirement 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 a ll 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 rel i ef request 14R-04 at the Wolf Creek Generating Station for the remainder of the fourth 10-year ISi interval , whi c h ends on September 2, 2025. All other requirements of ASME Code, Section X I , fo r which rel ie f was not specifically requested and authorized by the NRC staff remain applicable, in clud ing the third p*arty review by the Authorized Nuclear I n-service In spector. Thi s verbal authorization does not preclude the NRC staff from asking additiona l clarification questions regard i ng Relief Request 14R-04 , while prepar i ng the subsequent written sa fety evaluation. From: Sent: To: C c: Subj e ct: exam Att a chm e nt s: Coll i ns, Jay 3 Nov 2016 22:10:48 -0400 Alley, David;Pascarelli, Robert Singal, Balwant;Cumblidge, Stephen;Caponiti, Kathleen RE: Request for additiona l Information, Wolf Creek, 14R-03 and 14R-03, Nozzle Wolf Creek verba l auth 14R-03 11-03-2016 Rev 2.docx My apologies on the 14R-03 script, but I d i d not change the number of nozzles affected in the DORL section. I have now fixed that issue in red line str i keout. Attached i s Rev 2 of that script. Jay From: Alley, David S e nt: Thursday, November 03, 2016 8:44 PM T o: Pa s carelli, Robert <Robert.Pasca r elli@nrc.gov > Cc: Singal, Balwant <Balwant.Singal@nrc.gov>; Co ll ins, Jay <J ay.Collins@nrc.gov>; Cumblidge, Stephen <Stephen.Cumblidge@nrc.gov>; Caponiti, Kathleen <Kathleen.Caponiti@nrc.gov> Subj ec t: Request for additional I nformation, Wolf Creek, 14R-03 and 1 4R-03, Nozzle exam 1. I conc u r wit h t h e attached RA I 2. I conc u r with both of the ver b a l scr i pts 3. Stephen , please draft a 665 fo r t he RA I a n d se nd it t o Ka th leen 4. If we get t he RA I sen t ou t an d a res p onse b ack , I have n o ob j ec t ion t o do ing th e verbals on F r i day. I f we don't get a r esponse back for t he RA I , I propose we do the verba l s on Mo n day. I a m o p en t o doing t hem ove r t h e wee k e nd i f n eeded. I f done t o m o r row, S t ep h en i s acting a n d w ill do my pa rt. D ave David Alley PhD. Chief, Component Performance NOE and Test i ng Branch US Nuclear Regulatory Commission 11555 Rockville Pike Rockville MD 20852 301-4 1 5-2178 VERBAL AUTHORIZATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 14R-03 ALTERNATIVE TO USE VOLUMETRIC LEAK PATH FOR SUPPLEMENTAL EXAMS WOLF CREEK GENERA TING 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 a t 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 exa m is identified, the li censee will revert to th e requirements of Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D). The licensee made th is request i n accordance with the requirements of 10 CFR 50.55a(z)(2), such that compliance with the speci fied requirements would result in hardship o r unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff finds that while the demonstrated volumetric le ak path is not equivalent to a fully qualified surface leak path assessment, the licensee i dentified sufficient operational experience, technical basis and radiological dose hardship to s how that r eg ulatory compliance would result in hardship without a co mpensating 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 th at 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 l ikely to see subsequent and more sign ificant crack in g in the future. Given the lack of the initial cracking being id entified 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 les s likely. For technical basis , the licensee identified that their inspect ion would be in compliance with the Wesdye Technical Justification Document s howing an effec ti ve 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 progra m under 10 CFR 50.55a(g)(6)(ii)(D). Th e licensee also referenced NUREG/CR-7 142 , Ultrasonic Phas ed Array Assessment of the Interf e rence Fit and Leak Path of the North Anna Uni t 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation , which found, in part, th e u se of a properly focused 0 degree probe cou ld detect a le akage 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

the effectiveness of the volumetric leak path exam i nation to detect l ow leakage rates , as performed in accordance with the licensee's proposed alternative. , Co mm e n t ICJI: Optional addition a l 'j t ec h nica l b asis fo r re l ie f. I ncl u de i f d es ir e d. For hardship, the licensee noted t hat a qualified surface leak path assessment cou l d be j performed in two manners t h at would require both addit i onal radiological dose and time : 'I versus the performance of a volumetric leak path assessment. The NRC staff found both of these conditions to be of suff i cien t hardship g i ven the operational experience and / technical adequacy of the licensee's proposed alternative versus the regulatory / requirement l*-------------------------------------------------------------------------------------------j Therefore , the NRC staff f i nds that the l i censee's proposed alternative provides reasonable assurance of structural integrity unti l the next scheduled examination, and that comp l iance with the surface examinat i on requ i rements of Paragraph -3200{b) of ASME Code Case N-729-1, for the subject welds, would resu l t in hardship without a compensating i ncrease in the level of qual i ty and safety. Authoriz a tion re ad b y Robert P as c a relli , Ch ie f of th e Pl a nt Li c en s ing B ra nch IV-1 , Office of Nu c le a r Re a ctor Regul a tion As Chief of t he Plant Licensing Branch I V-1 , Off i ce of Nuclear Reactor Regulation, I concur with the Component Perfo r mance , Non-Destruct i ve Examinat i on , and Testing Branch's determinations. The NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of li=le--a ll 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 wou l d result in hardship or unusua l difficulty without a compensating i ncrease in the level of qual i ty and safety. Accordingly , the NRC staff concludes that the l i censee has adequately addressed all o f the regu l a t ory requ i rements set forth in 1 0 CFR 50.55a(z)(2) and 10 CFR 50.55a(g)(6)(i i)(D). Ther e fore, t he NRC staff authorizes the use of relief request 14R-03 at the Wolf C r eek Genera ti ng S t ation during t he current refuel i ng outage subject to the l i censee's proposed alternative that i f an unacceptable indication by the leak path assessment or volumetric exam i s identified , t he licensee w i ll revert to the requirements of C o de Case N-729-1 and 10 CFR 50.55a(g)(6)(i i)(D). A l l other requirements of ASME Code , Section X I , for which relief was not specif i cally requested and authorized by the NRC staff remain applicable, including the third party review by the Author i zed Nuclear I n-service Inspector. This verbal authorization does not prec l ude the NRC s t aff from asking additional c l arification quest i ons regarding Relief Request 14R-03 , while preparing the subsequent written safety eva l uation. From: Sent: To: C c: Subj e ct: Att a chm e nt s: Singal, Balwa n t 4 Nov 2016 16:03:29 -0400 Collins, Jay;Alley, David;Cumb l idge, Stephen;Tsao, John Pascare ll i, Robert FW: RAI Response from Wolf Creek (Relief Request 1 4 R-03CAC No. MF8456) ET 16-0031.pdf See l i sting of Records A l ready Ava i lable 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 wou l d 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 Sen i or Project Manager (D i ablo Canyon and Wolf Creek) Nuclear R e gulatory Commi ss ion Division of Operating Reactor Licensing Ba l wa nt.S i nga l@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 ... From: Muilenburg William T [2] S e nt: Friday, November 04, 2016 3:40 PM To: Singal, Ba l want <Balwant.Singa l@nrc.gov> Subj e ct: [External_Sender] RA I Response from Wolf Creek Ba l want, H ere is our RAI response. I have verified that Monday morning is fine for a respon s e, we don't need it over the weekend. I f there are additiona l questions I am here Saturday and Sunday. Thanks , B i ll From: Sent: To:

Subject:

Attachments

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.gov 9 Nov 2016 10:56:56 -0500 A ll e y , D a vid wolf creek notes green book [Untitled].pdf Pl ease ope n the a t tac h e d docu m e nt. Thi s d oc um e nt was di g it a ll y se n t to you u s in g an HP Di g i ta l Se nd i n g d ev i ce. \jO\__f C!J -:::, r" , \Cl \&V) h\\tb .. (\ .1 - \ 1-r-r . \\\<:_-\\ W f\S TuA \---_ \ ----r:::::, <$, f\ r \,Lt \<_ ,-* \\l-m' \-lA'-'1:. 'QJ \<_J\ '\!:) \ o h 6 ..-'. C. f\L p::_c-f o-R..:S k \ '() .J.kJ2 , CN'- --- 1 t-$\\ "'2-:G { <\-& \::,,, flO\ cf\L.. L.. \ k-'--1--"" \ Y..: v\b o 0t Aq£ \0 V\ b u jq \ G l1 \f'r\fJ I o &:> \\JiJf \.<-IV\ G-"\ 'f 1\-f' \\{\:S

By letter dated October 11, 2016 , Wolf Creek Nuclear Operating Corporation (the licensee) submitted Rel i ef 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 env i ronment. If needed , the NRC staff may request for add i tional information to complete its techn i cal review. . Relevant P e n #! .... .... l QC L e vel Il l C om ments Br ie f E v a l u a t i on -Se e Eva l uation doc u ment for more deta il ! Yes ! N o ! ! *-----**--*-*r*****--***r*

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• • ----------*--*---------------------------*------------------*--------------1 X 1 1 !Leak so u rce -crack i n t he Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : : : j Le a k was repa i red us i ng a clamp assemb l y t o preclud e futu r e de g radatio n . .............. L. ...........

L. .......... .. .. ...................................................... .. .. .. .. .. ............................. --..... . : 1 l !Nothing emanating from the annulus region was con fi rmed. 2 l X l l j Lea k so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) l l l !Leak w as repa i red us ing a clamp assembl y to preclude future degradation.

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*------.l.--

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........... .. .. ... .. .. ...................................... . j : : : Nothing emanating from the annulus region was co nfi rmed. 3 ! X ! ! !Leak so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : 1 l !Lea k was repa i red us ing a c l amp assembl y to preclude future deg r adation . .............. l ............. .l .............. .. .. .. ........... .......................................................... ... .. .. .. .. ................................. --... . : : : jNothing ema n ating from the annulus region was confirmed. 4 1 X j j !Lea k so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) l l l l Lea k was repa i red us i ng a clamp assembly to preclude future degradation . .............. L. .......... .J. ............ .. .. .. .......... .......................................................... ... ... .. .. .. ...................................... . ! ! ! !Nothing emanat in g from the annulus re g ion was c o nfirmed. 5 1 X l 1 j Le a k so u rce -crack i n t he Canopy Sea l weld o f Nozzle 77. (3142.1 (1 ), and 3142.3 (a) l l l l Lea k was repaired us i ng a clamp assembly to preclude futu r e de g radatio n . .............. L. ......... .J ............ .. a.! .. ................................................ .............................................. ... .. !?.Y. .. .. .. ...................................... . 1 ! ! !Nothing emanat in g from the annulus region was confirmed. 6 l X j l j Le ak so u rce -crack in t h e Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : : : : Leak was repa i red us i n g a clamp assembl y to preclude future de g radatio n . .............. L ........... ) ............. . .. .. .. 2f .. ................................................................................................ ... .. .. .. .. ...................................... . ! j j ! Nothing e ma n at in g from the a nnulus re gi on was confirmed. 7 ! x ! ! !Leak so u rce -crack i n t he Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : 1 l !L eak w as repa i red us ing a clamp assembl y to preclude future degradation . .............. L. .......... .J ........... .. . .. ... ':!.t! .. . ..................................... ............ .................. ................ ... .. .. ...................................... . : j : !Nothing ema n at in g from the annulus re gion was co nfir med. 8 ! X ! ! !Leak so u rce -crack in t he Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) 1 ! 1 !Lea k was repaired us i ng a cla m p assembl y t o preclude future de g radation . .............. L. .......... .L ........... .. .. ?..t .. .................................................................................................. ....................... .. .. .. .. .. ...................................... . : l l !Nothing emanat in g from the annulus region was confirmed. 9 j X j j !Lea k so u rce -c r ack in the Canopy Sea l weld of Nozzle 77. (314 2.1 (1), and 314 2.3 (a) l l l !Lea k was repaired u s i ng a clamp assembl y to preclude fut ure degradation.

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*--*---_L ___________ _L ___________

.. !().<?.:?.e. .. .. !?.<?.r.()_n, __ ... .. .. .. ...................................... . ! ! ! l Nothing emanating from the annulus region was co nfir med. 10 1 X l 1 j Lea k so u rce -crack in t h e Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : ! : jLeak w as repa i red us i ng a clamp assembl y to preclude future degradation . .............. L. .......... .! ............. .. 1.CJ.CJ:?.e. .. .. Cl.n. .. .. .............................................................................. ....................... .. ... .. .. . .. ...................................... . l l !Nothing emanating from the annulus region was co nfi rmed. X j 1 1 Leak so ur ce -crack in t h e Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! l j Leak was repaired us i ng a c l amp assembly to preclude future degradation. l ! Mi n or dry loose particulate dusted around nozzle l Structural integrity NOT compromised by the s u rface rust fo r m of degradation. 11 jNothing emanating from the annulus region was confirmed. 12 x so ur ce -crack i n the Canopy Sea l 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.1?..ry .. .. <?.n. .. ................................................................................... ................................. .... ... .. !?.Y. .. .. ...................................... . ! ! ! !Nothing emanat in g from the annulus region was confirmed. 13 1 X 1 1 1 Leak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! ! ! l Leak was repaired us i ng a clamp assembly to preclude future de grada tion . ........... ... L. ........... L. ........... 1.1?..ry __ .. .. <?.n. .. Y.r.! .. .. o.f . ................................... ............................................... .. .. !?.Y. .. .. .. ...................................... . ! ! ! !Nothing emanating from the annulus region was confirmed. 14 1 X 1 1 jlea k so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! : l !Leak was repaired us ing a clamp assemb ly to preclude future degradation . .............. l .......... ... .! ............. .i.1?..r:Y. .. .. . .<?.n. .. Y.r.! .. .. 0.f . .............. ..................................... .............. ........... ..... .. .. !?.Y. .. .. .. ...................................... . : ! ! jNothing emanating from the annulus region was confirmed. 15 i X i i !Lea k so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! : l jLeak was repa ir ed using a clamp assembly to preclude future de grada tion . .............. l. ............ J ............. l.P..r:Y. .. .. ?..n. .. Y.r.! .. .. O.f .. ......................................................... ......................... .. N..9!. .. .. !?.Y. .. .. ...................................... . ! : ! !Nothing emanating from the annulus region was confirmed. 16 1 X 1 1 1Leak so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 )._and 3142.3 (a) ! i i :Lea k was repaired using a clamp assembly to preclude future degradation . ............. .l. ............ .J. ............ .l.1?..r:Y. .. ................................................................................................................................. .......... .. .. !?.Y. .. .. .. ...................................... . ! j ! !Nothing emanating from the annulus region was confirmed. 17 1 X 1 1 1Le ak source -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) i ; ; ; Leak was repaired using a clamp assembly to preclude future degradation . ............. .l. ............ .i. ............ .i.l3.<?.ro.11 .. . .o.r:i .. .. . ..................... ................. ..................... ......................... .. .. .. .. .. ..................................... . ! i ! lNothing emanating from the a nnulus region was confirmed. 18 l X l l jleak so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! : ! ! Leak was repaired using a clamp assembly to pr e clude future de grada tion . .............. L. ........... L. ........... l.1?..r:Y. .. .. <?.n. .. Y.r.! .. . .0.f .. ............... ............................... .................................................. .. !?.Y. .. .. .. ................. ..................... . : : : :Not hing emanating from the annulus region was confirmed. 19 ! X l ! \eak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! ! l ! Leak was repaired using a clamp assembly to preclude future degradation. . . .o.r:i ... .. . .9.f . .............. ..................................................................... .. ... .. .. ................. ..................... . i i j j Unable t o visua ll y confirm no nozz l e leakage (3 1 42.1 (b) and (c), 3142.2 , 3 1 30 , and 3 1 32.1 (a)) 20 ! X ! ! j Supplemental Vol u met ric Exami n ation performed i de n tifying n o ch ange i n structura l i i i !character i stics from previous Vo l umetr i c exams of all n ozzles performed in 2006 and 2013 . .............. l ............. .i. .......... ... l.l?..r:Y. .. .. <?.n. .. Y.r.! .. .. o.f . .. . .. .. ..................... .. .. .. .. .. .. ........................................ .......... . : : ! !Nothing emanat in g from the annulus region was confirmed. 21 i X i i !Lea k so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) ! ! ! !Lea k was repaired us in g a clamp assembl y to preclude future degradation . ............. .l. ............ .L ............ L!?..r:Y. .. .. <?.n. .. Y.r.! .. .. 0.! .. ................... ............ ................................................... .. !?.Y. .. .. ru..s.!.f.O.!.rT1 .. ............................. ......... . ! ! ! ! Nothing emanating from the annulus region was confirmed. 22 l X l j jLea k so u rce -crack i n the Canopy Sea l 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. .. . .'?.11 .. Y.r.! .. ...................... ............ ................... .................................................. .. .. !?.Y. .. .. .. ...................................... . ! ! l !Nothing emanating from the annulus region was confirmed. 23 ! X ! ! jLea k so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) ! ! ! !Lea k was repaired using a clamp assembly to preclude Mure de gradati on. ! ! !Dry loose particulate on UH side l Structural int eg rity NOT compromised by the surface rust form of degradation. jNothing emana tin g from the a nnulus re gi on was confirmed. 24 x so urce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 }, and 3142.3 (a) l Leak was repaired using a clamp assembly to preclude futu re degradation . ............. .l. ............ .l. ............ .l.1?..ry .. .. C?.ri .. lJ.lj. .. 0.! .. ............................................. ................................. .... ... .. ?.Y. .. .. ru..s.t .. .. ............................... ....... . ! ! ! !Nothing emanat in g from the annulus region was confirmed. 25 1 X 1 1 1 Leak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : ! : l Leak wa s repaired using a clamp assembl y to preclude future de grada tion . .............. i. .............

i. ............. . .o.r:i .. ..

.. .. .o.f .. ................... .................. .. .. .. .. .. ...................................... . i i i :Nothing emanating from the annulus re gio n was confirmed. 26 1 X 1 1 \eak source -crack in the Canopy Sea l weld of Nozzle 77. (314 2.1 (1 ), and 3142.3 (a) : ! l !Lea k was repaired using a clamp assemb ly to prec lu de future degradation.

• • • • • • • • • • • • • • !**********
      • !**************!.1?..':Y.

.. . .C?.rl .. lJ.lj .. s.!9.!7-............. .................................. ...................... ............................................... .. N..9.!. .. .. ?.Y. .. .. .. ...................................... . ! ! ! !Un a ble t o vis u ally confirm no nozz l e leakage (3142.1 (b) and (c), 3142.2 , 3 1 30 , and 3132.1 (a)) 2 7 i X i l l Supplemen tal Volumetr ic Examin a tion performed i dentifying n o c hange in structura l ! ! ! !cha ra cteristics from previou s Vo lu metr i c exams o f all n ozzles pe rf ormed in 2006 and 2013 . .............. L ...... -.... .t ............. L!?..':Y. .. .. .. ............................................................................ -............................ .. .. .. ............................................. -... . : : : l Nothing emanating from the annulus re gion was confirmed. 28 i X i j jLea k source -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! ! ! ! Leak wa s repaired using a clamp assembly to preclude future degradation . .............. L. .......... .J. ............ .l.1?..':Y. .. . .C?.rl .. lJ.lj __ .. 0.! .. .............................................................. .................... .. .. .. ....................... ............... . : ! : !Not hing emanat in g from the annulus region was confirmed. 29 i X i i i Leak source -crack i n the Canopy Sea l 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?.ri .. lJ.lj __ .. o.! .. ............ ................................................................ ...... .. .. .. .. ...................................... . : ! l !Nothing emanat ing from the annulus region was confirmed. 30 1 X 1 1 1 Leak source -crack i n the Canopy Sea l weld of Nozzle 77. (314 2.1 (1 ), and 314 2.3 (a) : : l : Le ak w as repa ired using a clamp assembly to preclude future degradation . .............. L. .......... J ............ .. ?.r:1 .. .. CJ.f .. ......................... ............................... ... .. .. .. ...................................... . ! l ! !Nothing emanating from the annulus re gion was confirmed. 31 l X l j \eak source -crack i n the Canopy Sea l weld of Nozzle 77. (314 2.1 (1 ), and 3142.3 (a) : : : !Leak was repaired using a clamp assembly to preclud e future degradation . .............. 1 ............. .J. ............ .l.1?.r:x .. .. lJ.lj .. .. C?.! .. .................................... ................................. ............. .. .. .. .. ...................................... . : : l l Nothing emanating from the annulus region was confirmed. 32 i X i i !Lea k source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) ! ! ! !Lea k was repaired using a clamp assemb ly to preclude future degradation . ............. .l. ............ .L ........... .l.1?..':Y. .. .. C?.ri .. lJ.lj .. .. C?.! . ................................................... ...................... ......... .. ?.Y. .. !!2El .. .. ...................... ................ . : ! : !Nothing emanating from the annulus re gion was confirmed. 33 i X ! i !Lea k source -c rac k in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : ! : !Lea k was repaired us ing a clamp assembly to preclude future degradation . .............. 1 .......... ... .i. ............. i!?..':Y. .. .. lJ.lj __ .. o.! .. .. .. ........................... ... .. ?.Y. .. .. .. ru..s.!.f.o.r.ri:i .. ...................................... . ! ! ! ! Nothing emanating from the annulus region was confirmed. 34 l X l j jLe ak so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) i i i iLe ak was repaired using a clamp assembly to preclude futu re degradation. i i iSl igh Umino r dry particulate on UH side of nozzle !structural integrity NOT compromised by the surface rust form of degradation.

to visually confirm no nozzle leakage (3142.1 (b) a n d (c), 3142.2 , 3130 , and 3132.1 (a)) 35 : X : : : Supp l emental Volumetric Examination performed i dentify i ng no cha n ge in str u ctura l ! ! !characteristics from previous Volumetric exams of a ll n ozzles performed in 2006 and 2013 . ............. .l ............. .L ........... .lPr.Y. .. .. .. .................... ................................. ..................... .............................. .. .. .. .. .. .. .. .................................. ................ . : : j :Nothing e manating from the annulus re g ion was confirmed. 36 1 X 1 1 jLeak so u rce -crack t he Canopy Sea l weld of Nozzle 77. (3142.1(1),_and3142 .3 (a) ! ! ! ! Lea k was repaired using a clamp assembly to preclude futu r e degradation . .............. L .......... .L ........... .lPr.Y. .. .. ?..n. . .. . .......................................... ............................... ........... .. .. .. .. .. .. ................. ........... .......... . : l : !Nothing emanating from the annulus region was confirmed. 37 1 X 1 j j Leak so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ). and 3142.3 (a) : : : :Leak was repaired using a clamp assembly to preclude future degradation . .............. i .............. i .............. .. LJ.f:! .. .. 9.f .. .......... .............................................. .. .. .. .. .. ...................................... . ! ! ! !Nothing emanating from the annulus region was confirmed. 38 l X l l jLeak so ur ce -crack i n th e Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : : : : Lea k was repa i red using a clamp assembly to preclude future degradation . .............. L ............ .t .......... ... L!?..r.Y. .. .. . .. .. .. o.! .. ........................ ............ .............................................. ... .. .. .. .. ...................................... . : : : j Nothing ema n ating from the annulus region was confirmed. 39 ! X ! ! !Lea k so ur ce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : : : : Leak was repaired using a clamp assembly to preclude future degradation . ........... .. + ............ j .............. .. o.r.i .. Y f:! .. . ..................................................................... ... ... .. .. .. .. ................. .................... .. : l : : unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2 , 3130 , and 3132.1 (a)) 40 ! X l l !supp l ementa l Volume t ric Examination performed i dentifying no change in str u ctura l ! ! ! kharacteristics from previous Volumetric exams of a ll nozz l es performed i n 2006 and 2013 . .............. i .............. i ............. .!P.r.Y. .. . .?..n. .. .. .. ............. ..................... ............................ ................. ................ .. .. .. .. .. .. .......................... .................... .. : ! : : Nothing e ma n ating from the a nnulus region was confirmed. 41 l X l l ! Leak so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) i i i l Leak was repa i red using a clamp assembly to preclude future de g radation . .............. L .......... .!. ............ .1P.r.Y. .. .. . .. .. .. o.t .. ........................ ............ .............................................. .. .. ?.Y. .. t.h_e. __ .. .. ...................... ................ . ! l l !Nothing emanating from the annulus region was confirmed. 42 ! X ! ! !Leak so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1),_ and 3142.3 (a) : : : :Le a k was repaired using a clamp assembly to preclude future de g radation.

! iDry particula t e at nozz le interface (small amount) !Structural integrity NOT compromised by the surface rust form of degradation.

• • • • • • • • • • • • • r***********r*********
  • r******************** .. *******************

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  • • • • • • • • .. **************** .. *************************************** .... 43 l X l l !Lea k so u rce -crack i n t he Canopy Sea l 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. .. .. .. .. .. 0.f .. ................................ .................................................. ... .. .. .. .. ...................................... . 1 ! 1 ! Nothing emanating from the annulus region was confirmed. 44 1 X 1 1 jLea k so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : j : !Leak was repaired using a clamp assembly to preclude future degradation . ............. .i. ............ .i.. ........... .i.!?..r.Y. .. .. .. .. .. .. 0.f .. ................................................................................ .. .. .. .. .. ...................................... . ! ! ! ! Nothing ema n ating from the annulus region was confirmed. 45 ! X ! ! ! Leak so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : : : iLeak was repaired using a clamp assembly to preclude future degradation.

l iSl i ghUminor dry particulate on 90 deg r ee a n d UH s id e of nozzle )Structural integrity NOT compromised by the surface rust form of degradation.

to vis u a ll y confirm no nozz l e leakage (3142.1 (b) and (c), 3142.2 , 3130 , and 3132.1 (a)) 46 : X : : : Supplemen t a l Volumetr i c Examination performed i de n tifying no change in structural ! ! !characteristics from previous Volumetr i c exams of all n ozzles performed in 2006 and 2013 . .............. 1 .............. .. .. .. Y.tt .. .. 0.f .. .................................................. ................................ .. .. .. .. .. .. .................................................. . ! ! ! !U n able to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2 , 3130 , and 3 1 32.1 (a)) 47 : X : : : supplementa l Volumetr i c Exam i nation performed i dent i fying no change in structu r al ! ! ! !characteristics from previous Vo lu metric exams of all n ozzles performed in 2006 and 2013 . ............. .l. ............ .l. ............ .. ... .. ............................................................... ......................................... .. .. .. .. .. .. ................................................ . : : l l Nothing emanating from the annulus region was confirmed. 48 [ X [ [ [Le ak so ur ce -crack the Canopy Sea l wel d of Nozzle 77. (3142.1(1).,and3142.3 (a) ! ! ! ;Le ak was repaired using a clamp assembl y to preclude future degradation . ............. .L ........... .L ........... .. ... o.r.i .. .. ................................... ........... ... .. .. .. ...................................... . : : l lNot hing emanating from the annulus region was confirmed. 49 f X f f l Leak source -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : : : :Leak was repaired using a clamp assembl y to prec l ude future degradation . .............. L ........... L ........... .. . .. .. ................................. ... .. !?.Y. .. .. .. .. ................. .................... .. i i i i Nothing emanating from the annulus re gion was confirmed. 50 [ X ! i i Leak so ur ce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : : : : Leak was repaired using a clamp assembl y to preclude future degradation . ............. .L ........... J ............ . .. .............................. ... __ !?.Y. .. .. .. .. ...................................... . ! ! ! !N othing emanating from the annulus region was co nfirmed. 51 ! X i [ iLe ak so ur ce -crack i n the Canopy Sea l wel d of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : : : l Leak was repaired us in g a clamp assembly to preclud e futu re degradation . .............. i .............. i .............. .. .. .. ?..n. .. Y.tl .. .. <?.f .. ................................................... .............................. ... .. .. .. .. ................. .................... .. : : : : Nothing emanat in g from the a nnulus region was confirmed. 52 ! X ! ! !Le ak so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) i i i l Leak was repaired us in g a clamp assembl y to preclude futu re degradation . ............. .L ........... L ........... .. .. o.r.i .. ............... ......................... .............................................................. .. .. !?.Y. .. .. .. .. .................... .................. . : : : :Nothing emanating from the annulus re gion was confirmed. 53 [ X [ [ !Leak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1)., and 3142.3 (a) : : : : Lea k wa s repaired using a clamp assemb ly to preclude future degradation. i i i Dry loos e particulate on UH side of penetra t ion 1 Structural integrity NOT compromised by th e surface rust form of degradation . ............. T ............. 1" ........... T .............................................................................................................................. ................. .............................. fNoi'tiiri9"eiiii nai iri 9"trom .. itie"aii*nuiu*;;

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..................................................... ................ .......... . 54 ! X ! ! jLea k so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : 1 : : Leak was repaired using a clamp assembly to preclude future degradation . ............. .L ........... .L ............ .. .. .. .. .. !C? .. ................ ................................. .................. ... t:-!.9.!. .. ... .. .. .. .. ...................................... . ! ! ! ! Nothing emanating from the annulus region was confirmed. 55 [ X ! [ iLea k source -crack i n the Canopy Sea l weld o f No zzle 77. (3142.1 (1), and 3142.3 (a) : 1 : 1Leak w as repa i red using a clamp assembly to preclude future degradation . .............. L ............ L ............ .. .. .. ?..n. .. Y.tt .. .. o.f .. .................................................... ............................ ... t:-!.9.!. .. .. !?.Y. .. .. . .. ...................................... . ! ! ! ! Nothing emanating from the annulus region was confirmed. 56 ! X ! ! ! Leak so ur ce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ). and 3142.3 (a) : : : iLea k was repa i red using a clamp assembl y to preclude futu re degradation. i 1 i Dry particula t e on UH s ide of nozzle ) Structural integrity NOT compromised by the surface rust fo rm of degradation. jNothing emanating from the annulus region was confirmed. 57 x so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 }, and 3142.3 (a) l Leak was repaired using a clamp assembly t o preclude future degradation.

• • • • • • • • • • • • • • l***
                    • l**************l.1?..ry __ P..?.!.!!0.!.1.?.t.e.

.. <?.n..Yf:i .. .. .............. ..................................................................... ... .. .. .. .. ...................................... . ! ! ! !Unable to visua ll y confirm no nozz l e leakage (3142.1 (b} and (c), 3142.2 , 3 1 30 , and 3132.1 (a)) 58 ! X ! ! !Supplementa l Volumetr i c Examination performed i dentifying no change in structural

: l [characteristics from previous Vo l umetr i c exams of all n ozzles performed in 2006 and 2013.

.. .. .O.ll .. .. .. ............................................................. .. .. .. .. ............................. ................ ..... . ! ! ! !Unable to visually confirm no nozz l e leakage (3142.1 (b} and (c), 3142.2 , 3130, and 3132.1 (a)) 59 ! X ! ! ! Supplementa l Volumet ri c Examination performed i dentifying no change in structural l l lBoron accumulation (lo ose part ic ulate and hard caked boron) on UH side of l characteristics from previous Vo l umetr i c exams of all nozzles performed in 2006 and 2013 . ............. .1. ............ .1. ............ .. ....................................... ..................................... ........... .. .. .. .. ........................................ .......... . : : l lNothing emanating from the annulus region was confirmed. 60 f X f f l Lea k source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : l l [Leak was repaired us in g a clamp assembly to preclude future degradation . .............. L. ........... L. ........... .. .. .O.ll .. .. . ................. ............. .................. .. .. !?.Y. .. .. .. ...................................... . l ! : : Nothing emanating from the annulus region was confirmed. 61 ! X ! i !Leak so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ). and 3142.3 (a) : l : l Leak was repa i red using a clamp assembly to preclude future degradation . .............. l ............. .l. ............. . .. .O.ll .. Y.i::t .. . .<?f..ll.<?!.<:!.e. ......................................... .. .. ?.Y. .. .. .. ...................................... . ! ! ! !Nothing ema n ating from the annulus region was confirmed. 62 ! X ! ! !Leak so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : ! : l Leak was repa i red using a clamp assembly to preclude future degradation.

• • • • • • • • • • • • • • !***

.. *********!**************j.l?..r:Y. .. .. . .<?.n. .. Y.f:i .. .. .. .................... .............................................. ................ .. .. .. .. f.<?.!.ll:' .. ...................................... . ! ! ! !Unable to visua ll y confirm no nozz l e leakage (3 1 42.1 (b) and (c), 3142.2 , 3130 , and 3132.1 (a)) 63 i X i i l Supplemental Vol u metr i c Examination perfo r med i dentifying no c hange in structural ! ! [Dry rust colored boron cake on 90 side of nozzle -Dry loose particulate on UH !c haracteristics from previous Vo l umetric exams of all nozzles performed in 2006 and 2013 . ............. .l. ............ .L ........... . .0.f .. ..................................................................................... ............ ........................................... .. .. .. .. .. .. .. ....................... ........................ ... . : l l [Nothing emanat in g from the annulus region was confirmed. 64 [ X ! ! [Le a k so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) : ! : l Leak was repa ire d using a clamp assembly to preclude future degradation . ............. .l. ............ .L ........... .l.1?..r:x .. .. <?.n. . .Yf:i .. .. .o.! .. ...................................................................................... .. .. .. .. ...................................... . : ! : l Nothing emanating from the annulus region was confirmed. 65 ! X ! j i Leak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ),_and 3142.3 (a) ! ! ! ! Leak was repaired using a clamp assembly to preclude future degradation . ............. .l. ............ .l. ............ .. .. .o.11 .. .. .............................................. .. .. .. ...................................... . : : : l Nothing emanating from the annulus region was confirmed. 66 ! X ! ! iLeak so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1), and 3142.3 (a) : : : !Leak was repaired us in g a clamp assembl y to preclude future degradation . .............. L. .......... .!. ............. .. !9.0.?.e. .. .. . .o.f . .................. ................................ ................................ ... ... .. . ...................................... . ! ! ! i Nothing emanating from the annulus region was confirmed. 67 ! X ! ! ! Leak so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ). and 3142.3 (a) ! : l jLea k was repaired us ing 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 emana ting from the annulus region was confirmed. 68 x so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 }, and 3142.3 (a) l Leak was repaired using a clamp assembly to preclude future degradation . ............. .!. ............ .!. ............ .l.1?..ry .. .. C?.11 .. lJ.lj. .. 0.! .. ............................................. ..................................... ... .. .. .. f<?.r.".1 .. ...................................... . ! ! ! !Nothing emanat in g from the annulus region was confirmed. 69 1 X 1 1 1 Leak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! ! ! l Leak was repaired us i ng a clamp assembly to preclude future de grada tion. .. .. .. .. ................................. ............................................... ........... .. .. :0.Y. .. .. .. ...................................... . i i i !unable to visually confirm no nozzle leakage (3142.1 (b) and (c), 3142.2 , 3130 , and 3 13 2.1 (a)) 70 i X i i i Suppleme nt a l Volumetric Examination pe rformed i dentifying no c hange in structural i i !Significant boron accumulation (loose particulate and hard caked boron) on !characteristics from previous Vo lu metric exams of all n ozzles performed in 2006 and 2013. .. Clr:':d. .. .. .. 9.r1 .. ................................................. ............ ............................................. .. .. .. .. .. .................................................. . i i i !u n able t o visually confirm no nozz l e lea kage (3142.1 (b} and (c), 3142.2 , 3130 , and 3132.1 (a)) 71 ! X ! ! !Supplementa l Volumetr i c Exam ina tion performed i dentifying no c hange in structural i ! !Significant boron accumulation (loose particulate and hard caked boron) and !cha racter istics from prev ious Volumetric exams of all n ozzles performed in 2006 and 2013 . .............. L. ........... L. ........... .. .. 9.ri .. .. ......................................................... ................................ ............. ............ ...... .. .. .. .. .. .. ........................... ..................... . ! ! ! ! Nothing emanating from the annulus region was confirmed. 72 l X ! l j Leak so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ). and 3142.3 (a) ! l ! l Leak was repa ir ed using a clamp assemb ly to preclude futu re degradation . ............. .L ........... .J. ............ _l.1?..ry .. .. O..r1 .. lJ.t! .. .. °.! .. .................................... ..................... ............ ............. .. .. :O.Y. .. .. ...................................... . ! ! ! !Nothing emana ting from the annulus region was confirmed. 73 ! X l l jLea k so u rce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) ! ! ! l Leak was repa i red using a clamp assembl y to preclude future degradation . .............. l ............. .l. ............. .. d..rx .. .. .. .. . ............. .. .. .. .. .. ...................................... . ! ! ! ! Nothing emanat ing from the a nnulus re gion was confirmed. 74 i X ! ! !Lea k so u rce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ), and 3142.3 (a) i i i l Leak was repaired using a clamp assembly to p re clude futur e degradation . ............. .l. ............ .l. ............ .l.1?..ry .. .. . .o..ri .. lJ.lj .. .. a.f .. .................................................................................. .. .. !?.Y. .. .. .. .. ...................... ................ . ! ! l !Nothing emanating from the annulus region was co nfir med. 75 ! X ! ! ! Leak so ur ce -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 )._and 3142.3 (a) ! i i : Lea k was repaired using a clamp assembly to preclude future de g radation. ! ! ! Dry loose particulate on UH side of nozzle l Structural integrity NOT compromised by the surface rust form of degradation.

• • • • • • • • • • • • • r***********r***********r*****
                  • TNoi"tii n9**e-n;-a-r;a-i in 9 .. trom .. itie**a-;;*r;uiu*;;

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  • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 76 1 X 1 1 jLea k so u rce -crack i n the Canopy Sea l 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**************l.1?..ry

.. .. !?.°.!.C?.rl.9.r:' .. .. . .......................................... ................................. .................. .. ... .. .. .. ...................................... . i i i !unable to visually confirm no nozzle lea k age (3142.1 (b) and (c), 3142.2 , 3130 , and 3132.1 (a)) 77 ! X ! ! !Suppleme nt a l Volumetr i c Examination performed i dentifying no change in structural i [ i !characteristics from prev io us Volumetric exams of all n ozzles performed in 2006 and 2013 . .............. L .......... .!. ............. .. .. .. .. .. .. .. 9.r19 .. .. .. .............. .. .. .. .. .. .................................................. . ! ! ! i Nothing emanating from the annulus region was confirmed. 78 j X j ! j Leak source -crack in the Canopy Sea l weld of Nozzle 77. (3142.1 (1 ). and 3142.3 (a) ! ! l !Lea k was repaired using a clamp assembly to preclude future de gradati on. l ! !SliqhUminor amount dry loose particu l ate on UH side of nozzle lStr uctura l inteqritv NOT compromised by the surface rust form of deqradation. Vent x Vent X Boron on vent line appears to have come from a bove line Nothing emanating from the annulus region was confirmed. Leak so ur ce -crack i n the Canopy Sea l weld of Nozzle 77. (3142.1 (1 }, and 3142.3 (a) Leak was repaired using a clamp assembly to preclude future de gradati on. Structural inteqrity NOT compromised by the surface rust fo r m of deq rada tion. DE-McDermott Periodic Briefing , September 8 , 2016 PLANTS I A PR14 00 D esig n Ce rti fica ti on R eview (E E EB) >>Phase 2 Chapter 8 SER and Chapter 14.3.6 SER comp l eted and provided to NRO. S ta ff i s aw a it ing OGG interpretation of SECY 91-0078 to address the app li cant's conformance w i th the g ui dance i n the S E CY. Additional RAI was p r ovid e d 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. NuSca l e Sma ll Mo d ula r Reactor T opica l Report (EEEB) >> NuScale s u b mi t t ed TR-0815-16497, "Sa f e t y Class i fication of P assive NPP E l ectrical Systems" fo r NRG review a p proval of cond i tions o f a p p li cabi l ity. T he cond i tions o f applicabil i ty compr i se a se t of passive r eactor plant desig n a n d opera t ional attr i butes that , if met i n f ull , justify the appl i cant's determinat i on that n one o f the pla n t electr i ca l systems f u lfi ll func ti ons t hat wou l d wa rr ant a Class 1 E class ifi catio n. 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 (E EE B) >> Cab l e TI A. 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 c ollabo r at e on a draft response to Duke. DPR has scheduled a meeting on September 8 to begin this collaboration effort. >> DVR T I A. Reso l v i ng D E management comments prior to send i ng the d r aft T I A response to DPR for r eview. T he c urr e n t sched u le wi ll exceed the 2 year metric fo r O L Ts. Ginna Offs i te P ower TI A (EEE B) >> DPR i s revising and processing the Tl for the licensee's fact check. P er r y Deg r aded U ndervo l tage TI A (E EEB) >>N ew TI A from R egion Ill. Nor mal r eview sc h ed u le. I ndian Point 2 (EV I B) >> Li censee has replaced all 227 po t ent i a ll y defective bolts i n I P2 plus an addit i ona l 51 bo l ts to add safe t y m arg i n. Th e li censee performed an operability determ i na ti on fo r U nit 3 , wh i ch was found to be sufficie n t. Region 1 has issu ed a violaUon for the lic e nse e's handling of the issu e. Browns Ferry (EMCB) >> The st a ff has been notified of a pot e ntial hearing on the Browns Ferry EPU. Aft e r clarification discussions with the licensee, the staff is in agreement with their approach and is awaiting their final written respon se to RAls. Seab r ook (E M C B) >> The staff is reviewing the license amendment request related to ASR. Following the acceptance review , the staff ha s d e cided to non-accept with the opportunity to supplement.

The staff ha s provided input detailing the need for missing information and the PM is assembling the letter to the licensee. Wolf Creek (EPNB) Wolf C r eek exper i enced a leak in a canopy seal we ld above a reactor head penetration. This leak caused the plant to enter its re fueling outage ea rl y. This is not a pressure bo undary l eak per t ech specs or ASME code as the p r essure boundary is a threaded connection. Repa ir wil l be via a mechanical clamp. Re gion I V has the lead. The leak resulted i n significant boric acid residue at various locations inc l ud ing the head. The re is no known damage to the head at this ti me. Cleanup and inspec tion to ensu r e that none of the bo ri c acid came from a leaking nozz l e is an i nvo lve d process. T here will li kely be one or more relief requests. Exact nature of the requests has not yet been dete rmine d. 1 DE-McD e rmott Period ic Bri e fing , Sept e mber 8 , 2016 PROCESS I OPC (EEEB) >> Th e sta ff re ce ived comment s from the r e gion s and i s working to finaliz e the fir s t Tl (on interim corrective actions and compensatory measures). The Tl will be issued for regional implementation afte r th e Commi ss ion approves the IEP. , (C om m e n t IRMI: Do we h a ve a d at e? I ! I : D l&C Action Plan (E I CB) >> Issued to the Comm i ssion. D l&C CCF Po li cy Review (EICB) >> N EI also announced that they are developing a technical methodo l ogy for NRC endorseme n t. A second tabl e top was held on Augu s t 22. NEI is now fo cused on potential changes to our current guida 11c e for CCF (BTP 7-19). I ! f I D l&C 50.59 Guidance Review (E I CB) : >> The staff is holding bi-weekly meetings with NEI to resolve the key first-of-a-kind conceptual 1/ a pproaches being propo se d in App e ndix D and the deviation s from NEI 96-07. MRP-227-A (EVIB) f >>Staff is currently reviewi n g PWROG-15032 regard i ng CASS mater i als (Action Item 7) and expects to ! issue a safety assessment this *** J addresses the co l d work aspect of Act i on Item 1 that was recently submitted fo r informat i on. Rev. 1 to MRP-227 -A, which addresses mos t of the l i censee/applicant act i on items, has been received for staff review. Baffle-Former Bolts (EVIB) >> EPRI MRP started a baffle-former bolt focus group to develop an i ntegrated i ndustry response to the gene r ic issue for PWRs . The MRP focus group i ssued interim guidance on Ju l y 25 and the gu i dance generally lines up with the informa t ion i n the N SAL. LIC-504 h as be e n c ompleted a nd is in concurrence. Plan to issue an IN once test results come in this fall. GALL Report-BMI and Capsule w i thdraw, liner bulge new i ss u e PEOPLE On Board Budget EOY On Boa r d 78 82 74 Khadijah West on rotation from STSB to EEEB through December. Serg i u Basturescu on rotation from EEEB to STSB through February. Yong Li, Yuan Che n g, and Amit Ghosh will move to NRO COE on October 2 from EMCB. FY17 Budget 76 Sub i noy Mazumdar , Deirdre Spaulding, Euge n e Eagle (EICB), Jack McHale (EVI B BC), and Robert Hardies (S L S) opted fo r ear l y ret ir ement. Depart October 1. Developing staffing plan i nput to address 3 vacanc i es in EICB. Vacancies >> SLS Electrica l: Preparing ERB package t o support backfill for Hardies >> GG-14 E l ectronics Dl&C: EICB beginning process >> GG-14 E l ectronics Dl&C: EICB beginning process >> GG-13 E l ectronics D l&C: EICB beginning process >> GG-14 Materials: Agency-wide SOI closes this week. >> GG-14 Materials: Reassignment from JLD identified. Expec t ed on boa r d in J a n uary 2017. Overages >> EEEB: one unfunded position in FY17. Reassignment from JLD w i thout 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 Ob l iaations Remainina 1 061 448 484 384 0 1065* 2,544 2 , 544 2,544 0 Total 2,992 3 , 028 2,928 0 *Topica l Reports are budgeted in DPR and funding is moved to DE at the time i t is committed. The peen in g topica l report will be moved to RES because in essence it is confirmatory research. Final contract action for NRR/DE is in negotiation and shou l d be awa r ded and obligated by the end of the month. 3 Attachment 2 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 volumet ric or surface examination coverage distance below the toe of the J-groove weld (i.e. dimens i on "a") be 1.5 inches for incidence angle , e. less than or equal to 30 degrees; 1 inch for incide nce angle , e , greater than 30 degrees; or to the end of the tube , whichever is less. These cove rag e requirements are applicable to Wolf Creek Generating Station (WCGS) reactor vessel head penetrations as shown in Table 1. Table 1: WCGS Reactor Vessel Head P enetration Coverage Requirements Penetration Numbers 1 to 29 30 to 78 4. Reason for Request *red Coverage , " a" (i nches) 1.5 1.0 styles of ends , referred to gh 73 are Type "Y" that are 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 dis tan ce from the lowest point Id to the top of the threaded re gi on could be less than the "a" shown in Figu re 2 of ASME Code Case N-729-1. i red inspection coverage is sought for reactor vesse l , as th e require d cove rag e for these two penetrations

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From: Sent: To:

Subject:

Ba l want, Muilenburg William T 11 Oct 2016 19:44: 39 +0000 Singal , Balwant [External_ Sender] Participants from Wolf Creek on today's call On the call today for Wolf Creek were: Steve Smith, Plant Manager Richard Flannigan, Manager Nuclear Engineering Cindy Hafenstine, Manager Regulatory Affairs Dennis Tougaw, I Si Engineer Mark Barraclough, Boric Acid Program Owner Bill Muilenburg, Licensing Supervisor On the othe r request Debbie Hendel l , our Senior Counsel, said she will call this afternoon. From: Good N i cole R Sent: 28 Oct 2016 17:24:3 0 +0000 To: Drake, Jame s;All e y, D a vid;Collin s , Jay; Cumblidge, Stephen; Dod so n , Doug l as;Kopriva, Ron;Taylor, N ick;Thomas, Fab i an;S i nga l , Balwant;lingam, Siva Subj e ct: [Exte rn al_Sender) RE: Penetration pictures for Relief Requests 14R-03 and 14R-04 The upload is complete. Thank you, Nicole Good From: Good N ico l e R Sent: Friday, October 28, 2016 10:24 AM To: James.D r ake@nrc.gov; A ll ey, David (Dav i d.Alley@nrc.gov); Collins, Jay (Jay.Co ll ins@nrc.gov); Cumblidge, Stephen (Stephen.Cumb l idg e@n rc.gov); Dodson Douglas E; Ron.Kopriva @nrc.gov; Nick.T ay l o r@nrc.gov; T h o mas Fab i an D; Balwa n t.Singal@nrc.gov

Subject:

Penetration pictures for Relief Req u ests 14R-03 and 14R-04 Pictures have been uploaded for you in the Certrec IMS Sept 2016 Forced Outage folder I tem #27. The upload of the pictures to the folder in IMS is still in process. I will email when the uploads arc complete. T h ank y o u, Nicole Good Licensing n i lyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek Nucleor Operat i ng Corporotion From: Sent: To: C c: Subj e ct: Good N i co l e R 14 Oct 2016 16:09:15 +0000 Lingam, Siva Singal , Balwant;Collins, Jay; Kopriva, Ron;Dodson , Douglas;Thomas, Fabian [External_Sender) RE: WCNOC RV pictures I forgot to i nclude this. -Pi ct u re # -p e n etrat ion numb ers DSC00006-53, 64, 59, 36 , 47 , 60 DSC00039-63, 52, 34, 26, 20, 27, 58, 70, 75, 57, 33, 16, 12 DSC00029 -46 D SC00019-47, 71, 77 DSC00018 -40, 46, 70, 71 From: Good N ico l e R Sent: Fri d ay, October 14, 2016 11:05 AM To: 'Lingam, Siva' Cc: Singal, Balwant; Colli n s, Jay; R o n.K op r iva@nr c.gov; Dodson Doug l as E; T h omas F abia n D

Subject:

RE: WCNOC R V p ictures Mr. Lingam, I have b een 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 severa l pictures of the penetrations with labels. Incl u ded is M-706-00009 Reactor Pen, i t will help with orientation of t h e picture labels. Thank you, Nicole Good From: Lingam, Siva [mai l to:S i va.Linqa m@nrc.gov] Sent: T hursday, Octobe r 13, 2016 3:15 P M To: Good N icole R Cc: Singal, Balwant; Collins, Jay

Subject:

RE: WCNOC RV p ictures Please prov i de me the Certrec link and password with pr i nting r i ghts. Thank you. From: Good N icole R [mailto: ni l y o n@WCNOC.com ] Se nt: Thursday, October 13, 2016 4:05 PM To: Lingam, Siva Cc: S i ngal, Balwant <Balwant.S i ngal@nrc.gov > Subj e ct: [Ext erna l_S end e r] WCNOC RV picture s I was told you would like pictures of the p enetrations w i th l abe l s of the penetrat i on number. I have on l y been ab l e to locate a few pictures , at this point. I have granted you access to the Certrec I MS Sept 2016 F orced Outage. I tem #14 has five picture s t h at may be helpful (DCS00006, DCS00039, DCS00029, DCS00019, and DCS00018). I w i ll 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 nil y on@wcnoc.com (620) 364-8831 x 4557 Wolf Nucteor Operat i ng Corporot 1 on From: Good Nicole R Sent: 21 Oct 2016 13:52:47 +0000 To: Pascarelli, Robert;Lingam, Siva;Singal, Ba l want;Taylor, N i ck;Kopriva, Ron;Kennedy, Kriss;Dodson, Douglas;Thomas, Fabian

Subject:

[External_Sender)

Reactor Head Images Uploaded to Ce rt rec IM S To all: Images of the reactor head have been up loaded to the respective items in the CERTREC Inspectio n Management System folder for the September 2016 Forced Outage In spect ion: F l ange an d Head with Clamps installed -IM S item 23 F l ange Area As Found -IMS item 24 Eyebolt and Upper Canopy Seal Weld Design -IMS item 25 Videos of Head In spection -IMS item 26 Th ank you, Nicole Good Lic ensi ng nilyon@wcnoc.com (620) 364-8831 x 4557 Wolf Creek Nucteor Operoting Corporotion From: Sent: To: C c: Subj e ct: Number WO 16-0052 Att a chm e nt s: Hafenstine Cynth i a R 12 Oct 2016 22:10:42 +0000 Singal , Balwant;'siva.lingman@nrc.gov' Mu i lenburg William T;Tougaw Dennis E;Barraclough R i chard M [External_Sender) Wolf Creek -Draft revision of Relief Request Document W016-0052R5dt.pdf Attached is our current draft revision of the re l ief request. We have not yet incorporated the questions listed in the draft RAI that you provided. We wou l d like to have a call on Thursday at 1:00 pm Eastern Time I Noon Central Time. Please let me know if that wi ll 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: Sent: To: Subje ct: Ba l want, Muilenburg Wil l iam T 7 Oct 2016 21:17:37 +0000 Singal, Balwant [External_ Sender] Wolf Creek Re l ief Request Anticipated 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 (ML 12353A241 provided NRC Safety Assessment of the request), and second, we will be asking to perform a n alternate exam v. that specified in code case N-729. Can you he l p us assemb l e the right peop l e to have a phone call regarding this request on Monday morning? I wi l l be in Saturday and Sunday if there are any questions I cain help answer. Thanks, Bill Muilenburg 620-364-4186 From: Sent: To: Subje ct: Att a chment s: Ba l want/Nick, Muilenburg Wil l iam T 1 Nov 2016 14:41:44 +0000 Singal , Balwant;Taylor, Nick [External_ Sender] Wolf Creek Re l ief Requests ET16-0030.PDF See l i sting of Records A l ready Av a i l a ble to . . Here is the Relief Requests related to the WCNOC reactor vessel closure head. Key Points Bill l. This supersedes all previous correspondence on this topic 2. The reques t now encompasses all penetrations on the head From: Sent: To: Subje ct: Att a chment s: Hafenstine Cynth i a R 13 Oct 2016 17:00:49 +0000 Singal, Balwant [External_Sender] FW: Re l ief Request for Code Case N-729-1 M-7 06-00009_REACTOR PEN.JPG One-page attachm e nt withhe l d in full under ex4. New drawing for the draft relief request ... From: Barraclough R icha rd M Sent: Thursday, October 13, 2016 11:53 AM To: Hafensti n e Cynt h ia R Cc: Tougaw Dennis E

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R el i e f Reques t for Code Case N-729-1 This i s th e imag e I had Salvador Ferrara put together for the r elie f request R. Mark Barraclough Wolf Creek Nuclear Boric Acid Engineer I Program Owner Fluid Leak Management I Program Owne r AOV Engineer 620-364-8831 x8148 I r i barra@wcnoc .com Fax: 620-364-4154 (b)(4) From: Sent: To: S ubj e ct: System (IMS) Siva Lingam, I MS 14 Oct 2016 10:14:41 -0500 Lingam, Siva [Externa l_Sender] Log i n Informat i on -Certrec I nspect i on Management Welcome! You have been granted access to the CerlTec Inspection Management System in pr eparat i on for an upcoming NRC In spect i on. This reque s t 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 you r account. http s://i m s.cert r ec.com/ve r i fy/ Usemame: D b ( Verification Code: lf72 hour s 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 ex pired. ln the futur e, you will not n ee d to rep ea t this process to "Sign I n" t o th e Certrec I n spec tion Manag e m e nt System. The Ccrtrec Inspection Management System can be accessed at: http s://ims.c crt rcc.com As a lw ays , please feel free to contac t the Certrec Support Team i f you have any questions or concerns. Th ank you, Certrec Corporation Ce rtr ec Support Team s upport@certrec.com 817.738.7661 From: Sent: To:

Subject:

Attachments

and Corrosion.docx Wilson, George 31 Oct 2016 14:34:22 -0400 Orf, T racy FW: Wolf Creek Vessel Head Corros ion -One Pager and Q&As WOLF CREEK-RCS-Q&A.docx, Wolf Creek Reactor Vesse l Head Nozzle Leakage 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 <Geo rge.Wilson@nrc.gov>

Subject:

FW: Wolf Creek Vessel Head Corrosion -One Pa ge r 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, Septembe r 22, 2016 7:33 PM To: Wilson, George <George.Wi l son@nrc.gov>; Boland , Anne <Anne.Boland@nrc.go v> Cc: A ll ey, David <David.Alley@nrc.gov>; Evan s, Michele <Michele.Evans@nrc.gov>; Dean, Bil l <Bill.Dean@nrc.gov >; McDermott, Brian <Brian.McDermott@nrc.gov >; Lubinski, John <J ohn.Lub i nski@nrc.gov>; Ross-Lee, MaryJane <MaryJane.Ross-Lee@nrc.gov >; Pascare ll i, Robert <Robert.Pascarelli@nrc .gov>

Subject:

FYI: Wolf Creek Vessel Head Corrosion -One Pager and Q&As From: Taylor, Nick Sent: Thursday, September 22, 20 1 6 7:19 PM To: Kennedy, Kriss <Kriss.Kennedy@nrc .gov>; Morri s , Scott <Scott.Morris@nrc .gov>; Pruett, Troy <Troy.Pruett@nrc .gov>; Lantz , Ryan <Ryan.Lantz@nrc .gov>; Vege l , Anton <Anton.Vegel@nrc.gov >; Clark, Jeff <Jeff.C l ark@nrc.gov >; R4DRP-BC <G-R4-DRP-BC@nrc.gov >; Werner, Greg <Greg.Werner@nrc.gov >; Dricks, Victor <Victor.Dricks@nrc.gov >; Mai er , Bill <Bill.Maier@nrc.gov >; Mor e no, An g el <Angel.Moreno@nrc.gov >; Bowen , J e remy <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.Prou l x@nrc.gov>; Dodson, Doug l as <Douglas.Dodson@nrc.gov >; T hom as, Fab i an <Fabian.Thomas@nrc.gov >; G a l e mo r e, Su s an <Susan.Galemore@nrc.gov >; Kopriva, Ron <Ron.Kopriva@nrc.gov > S ub ject: 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 inspect i on of the vessel head! (anticipated within the next week). Please feel free to contact me with any quest i ons that you have. Thanks, Nick Tay l or Chief , Projects Branch B Division of Reactor Projects USNRC Region I V 0: (8 1 7) 20 0-1141 c: l<bl<5 i I E: n i ck.taylor@nrc .gov R Wolf Creek Reactor Vessel Head Nozzle Lea k age a n d Corrosio n Q's and A's 1. Do e s the r ed rust indicate damage to the re a c t or vess e l head? How soon will the extent of possibl e damage be known? Th e red co l ora t io n cou l d indicate corrosion of the reacto r vesse l h ead f r om t h e reac t or coo l ant syste m lea k age t h at became apparent on August 3 1 , 20 1 6 and conti n ued unt il the licensee f ull y depressur i zed the plant. The extent of this co r ros i on is curre ntl y un kn own, but believed t o be l imited to a sma ll secto r o f the head area. I t will take severa l weeks to c l ean th e vessel head and determine the extent of the cor r os i on. 2. What is the cause of the rust on the head? T h e r u st was li ke l y caused by a sma ll l eak spraying on t o the vessel head from a sea l weld on a m ec h a n ical j o i nt on a vesse l head nozzle above t he a r ea. T his leak man i fes t ed i tself on A u g u st 1 1, 2016 and cont i nued u ntil the plan t was dep r essurized.

3. Ple a se describe the extent of the discoloration

? How and when was this discovered ? Th e d i scolo r ation is l i m i t ed to a s m all sec t or u n d erneath the l eaking head nozzle. T h i s was discovered following a requi r ed shu t down on September 2, 2016. The l icensee was ab l e t o observe the disco l o r ation on the head after remov i ng i nsu l at i on from the vessel head o n Septembe r 1 9, 2016. 4. If there is damage , will the reactor head have to be repl a ced or can it be repaired? Depending on t he extent of co r rosio n , the li censee w ill eva l uate the r epa ir/rep l ace options, pr i o r to t he end o f the cur r e n t 55-day r ef u eli n g outage. 5. Will the licensee need NRC permission to restart? A t this time t h e r e is no indication t hat the licensee wi ll need N RC permission t o r es t a rt. NRC i n spec t ors w i ll verify that t he li censee has adequately add r essed the issue in o r de r to safe l y resta rt and opera t e t he pl an t. H owever, NR C perso nn e l are i n c l ose communica ti on with t he licensee, and w ill closely mon it or t h e l icensee's actions to repair t h e leak and the vessel head, i f needed. 6. Following th e incident a t Davis Bes s ie aren't all reactor lic e ns e es required to condu c t periodic vess e l h ea d inspe c tions? Has Wolf Cr ee k done the i rs? Yes. Wolf Creek performs i nspections of the t h eir vesse l head each r efue lin g o u tage in accordance wit h the i r approved in-serv i ce i n spection p r ogram. T he last s u ch i n spect i on w as in t he S p ring 20 1 5 outage. Contact: Nick T aylor, Chief, Reacto r 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 w it hi n the design basis of the plant. T he l i censee shutdown the plant afte r indi cation of a sudden inc r ease in l eakage beyond that allowed by the Technica l Spec ificati ons. 8. What role is the NRC playing i n this? The NRG resident in s p ecto rs and Region I V pe r so nnel monitored the li censee's actions promptly when indication of a leak first surfaced, and verified th at the l icensee too k act i on pr omp tly and i n accordance with t h eir operat i ng li cense. NRG i nspectors from the Region I V off i ce will be onsite beginning September 26 to assis t the residen t i nspe c t ors with their follow-up of this issue. NR G management is in communicat ion with Wolf Creek managem ent o n their path forward. 9. Is the licensee required to file written reports with the NRC regarding this? Yes. NRG r egu lation s require a written Li censee Even t Report (LER) for co mp le ti on of the T ec hn ica l Spec ifi cat i on-r equ ir ed shutdown , on Se p tember 2, 2016. Contact: Nick T aylor, Chief , Reacto r 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 C r eek noted an u pward trend in un id entified RCS leakage on Augus t 31, 2016. On Septembe r 2, 2016, Wol f C r ee k observed RCS unidentified lea k age in excess of 1.35 ga llon s per m i nut e (gpm), exceeding the T S l imit of 1.0 gpm. As a result , the li censee initiated a TS required s hut down on Sep t embe r 2 , 20 1 6.
• The resident inspectors mon i tored r eacto r coo l ant system l eakage throughout th e opera ti ng cyc l e. Data i ndicated a steady very sma ll leak rate (app r oxi m ately 0.05 ga ll ons pe r m in ute), that s u dde nl y began to in crease on A ug ust 31, 2016.
• Following shu t down and con t ainme n t entry , the so ur ce o f the leak was id entified as th e ca n opy seal weld on pene trat ion 77 on t he reactor vessel head, w h ic h serves one of the core exit thermocouples.

Th e threaded mechanical joint serv in g the core exit thermocoup l e nozz l e assembly is no t considered p r essu r e boundary l eakage.

• Following the s hutd own, th e li censee r emained sh utd own to commence the ir refueling outage, wh i ch is p l anned for 55 days. During this outage, th e licensee i s evaluating plans t o repair the leak using an applicable ASME code a ll owab l e methodology.

P revious minor leaks on mecha ni cal join ts o n the reactor vesse l head have been r epaired wit h code-approved me c h anical clamps. Th ere are 10 of th ese clamps c urr ently ins talled on vessel head nozz l e assemblies.

• The reactor vesse l head is the o ri g i na l head and i s app r ox im ate l y 30 years o l d. The l icensee ha s pe riodi cally in spec ted the head for leakage in accordance with the ir approved in-serv i ce inspection p r ogram. The last such inspe c ti on was in the Spr ing 2015 refueling outage.
• The license plans t o re m ove the vesse l head and place i t o n an i nspection stand. The r eac tor vessel head will b e clea n ed and exa min ed t o determine the exten t of th e corros ion, and if repairs are necessary.
• Reg i on I V i n spec t ors from the Divisi o n o f Rea cto r Safety are schedu l ed t o ar r ive on Sep t ember 26 to ass i s t the resident i nspectors in spec tion of this issue. Co nta ct: Nick Tay lo r, Chief, Reactor Proj ects Branch B {817) 200-1 141 September 20, 20 1 6 Wolf Cre e k Nucle a r Opera t ing Corpora t ion -0009 3 697 Co nd it i o n R eport AR#: 0 0 093697 Seve ri ty T y p e: CAQ L eve l: SSC Du e Da t e: 1211112015 S t at u s:COMPLETE Sta tu s Date: 12110 120 15 A R Subjec t: STS PE-040E Penetration 20 Canopy Seal Weld Le akage lndicati A ge In D ays: 264 Owed T o N a m e: DORATHY, BRIAN D Origi n at i on Date: 0311812015 Owed T o Depart m e nt: 4050050 -Dorathy Br ian Ini t i a t or: HA LL, JOHN F Owed To A l er t Gr o u p: Orig D e p artment: 0060030 -Heffron Jason Cond it ion R epor t Summary: T y p e AR#-A ssig n#-Sub-A ssign# O we d/Assign T o D ue D ate Status CAQ 00093697 BRDORAT 1 21 11/20 15 CO M P L ETE 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 RA CT 00093697-01-04 OPS REVIEW COMPLETE RACT 00093697-01-05 OPS REVIEW COMP L ETE RACT 00093697-01-06 OPS REVIEW COMPLETE RACT 00093697-01-07 OPS REVIEW COMPLETE RACT 00093697-01-08 OPS REVIEW COMP L ETE RACT 00093697-01-09 OPS REVIEW CO MPL ETE RACT 00093697-01-10 OPS REVIEW COMPLETE RACT 00093697-01-11 OPS REVIEW COMP L ETE RACT 00093697-01-12 OPS REVIEW COMPLETE RACT 00093697-01

-1 3 OPS REVIEW CO MPL ETE RACT 00093697-01-14 OPS REV IEW CO M P L ETE RACT 00093697-01-1 5 OPS REVIEW COMP L ETE RACT 00093697-01-16 OPS REVIEW COMPLETE RACT 00093697-0 1-17 OPS REVIEW CO MPL ETE RACT 00093697 1 8 OPS REVIEW COMP L ETE RACT 00093697-01-19 OPS REVIEW CA N CELED RACT 00093697-01-2 0 OPS REVIEW CA NCELED BLL 00093697-02 BR DORAT 0412012015 CO M PLETE PLAN 00093697-03 BR DORAT 1 2/11/20 1 5 CO MPLET E ACT 00093697-03-01 DAGIEFE1 11/20120 15 CA NCELED ACT 00093697-03-02 BR DORAT 1211112015 COMPLETE ACT 00093697-03 -03 BR DORAT 10130 120 1 5 COMP L ETE Attachme n ts: CR D eta il A sse t/Equip: RBB01 Work R , equest: 1 5-111261 Desc r iption: Dur in g the performance of STS PE-040E evidence of l ea k age was identified at the canopy sea l we l d on penetration

20. The indication appears on t h e east side of the sea l weld face (as the RPV c l osure h ead sits). Boron sta i ning is evident above, adjacen t and b , elow the seal weld indication.

R u st sta i ning is a l so prevalent at and below the i ndica t ion running down the CROM noz z l e and onto the top head surface. There is no evidence of degradatio n to the CRO M or closure head surfaces. This sea l weld ind i ca tio n appears to have been an active leak in the recent past. Photos/video ava i lable from QC (Jason Heffron). Recommend th i s sea l w e ld be eva l uated f or insta ll a tion of a sea l weld clamp assembly. CR Detail Repo rt Page 1 of 28 12 110120 1 5 1 1 :42: 24PM Wolf Creek Nuclear Operating Corporation Immediate Concern: N Imm edia te Actions: Initiate CR Ext ent of con diti on: RPV Canopy seal welds Recommended Re so lution: Instal l canopy seal weld c l amp assembly Screening R ev i ew Op era bility: CR Detail Report 00093697 Condition Report SM Notified: N/A 3 OPER/DNC The initiator ident i fied during the perfo r mance of STS PE-040E evidence of leakage at the cano py sea l we l d on penetration

21. The indication is on the eas t side of the sea l weld face (as the RPV closure head sits). Boron staining is evident above, adjacent and be l ow the seal weld indication. Rust staining is also prevalent at and be l ow the i ndication 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 l ocated at K:\Data\NDE\Photos\RF-20\CRDM Head In spect i on. In these pictures there are vis i ble traces of dried boron and some small amounts of discoloration on the sea l weld. As the i nitiator identified , there i s no significa nt accumulation of boron or wastage of any ca rbon steel on the head pe n etration directly below the subject sea l weld. Technical Specifications defines Pressure Boundary LEAKAGE as LEAKAGE through a noniso l able fault in an RCS component body, pipe wall or vessel wal l. TS 3.4.1 3 contains the operating li mits for RCS Operat i onal LEAKAGE. In MODES 1 through 4 , no pressure boundary is allowed, unidentified LEAKAGE is l i mited to 1 gallon per minute, ident i fied LEAKAGE is l i mited to 10 gallons per minute, and primary to secondary LEAKAGE i s l i mited to 150 gallons per day in any one Steam Generator. Th e Contro l Rod Drive Mechan i sm is what is used to raise, l ower , and trip contro l rods. The internals of this mechanism is exposed to RCS pressure. The Drive Mechanism Latch Housing is in ternal l y threaded and torqued down onto a seating surface at the interface between the housing and the top of the Reactor H ead Adapter. This connection is a mechanical jo i nt and l eakage via this pathway is not Pressure Boundary LEAKAGE as defined by Technical Specifications. The WCGS reactor ve ssel head and CRDM assemblies are classified as ASME Boiler and Pressure Vessel Code Section Ill Class 1 item s. Th e R eac tor Vess e l was designed and fabricated to the 1971 Edition through Winter 1972 Addenda and the CROM housing assemblies were designed and fabricated to th e 197 4 through Winter 197 4 Addenda of Section Ill of the ASME B&PV Code. Sec tion Ill paragraph NB-3671.3 states that threaded joint s un whi c h threads provide the on ly seal shall not be used. The sea l we l d is not a structural part of the pressure boundary and is not required to meet the structural requirements of ASME B&PV Page 2 of 28 lnit DNC: N 1 2/10/20 1 5 1 1 :42:24PM Wolf Cre ek N u cle a r Op e ra t ing C o rpora t ion Reportab l e: En v i ronme n ta l I ssue: Tec h Spec Sec 5: Personne l Safety I ssue: Reac t iv i t y I ssu e: Impa c t R i s k Asse ss ment: OPS Review: CR/WR Screening: Signifi ca n ce C a t: Screen/SRT N o t es: Gene r a l N otes: 0009 3 697 Co n d i tion R epo rt Code, Section Ill, NB-3000. The threads are the load ca rr ying part of the joint design. T h e in d ustry indications and past operating ex pe rience at WCGS of l eaks in the subject seal welds are pinholes or s m all local i zed cracks. T hese flaws have resulted i n leak rates that a r e bound by the l imits establish e d in Technical Specification 3.4.13. Com p l e te d p erf ormances of STS B B-006 w ere r evie w ed fr om the l ast operating cycle and RCS leakage l im i ts were not challenged. Th e Reactor Vessel and th e subject CROM is OPERABLE bu t degraded due to the flaw in the l ower seal we l d. References

Technica l Spec i fica t ions 1.1, 3.4.1 3 and Bases; TR 3.4.17 and Bases; N RC Inspection Manual Part 9900, WCGS Correspondence CT 02-0029 , Westinghouse I nstruction and Opera t ing B ook for M agne t ic Contro l Ro d Drive Mechan i sm for Full-Length Control Rods, and STS PE-040E. TSS 3/21/1 5 N ote-The r efe r ence to NRC inspection manual 9900 has been superceded by N RC i nspection manual 0326-Additionally, investigation by engineer in g has determined th e actual condition is on the canopy seal we l d on penetrat i on 20. Th i s does not change the ope r ablity basis. N N N N N N BELL , SETH A BEL L , SETH A 99 -N OT APP LI CABLE Updat e d By This is a l ong standing issue in the industry.

Westinghouse has prev i ous l y identified this t o be transgranular stress corrosio n crack i ng (TGSCC) suscept i bility in austenitic sta i nless steel influenced by the environment , stress, and microstr u cture. The residual stress associated with the BR DORAT BRDORAT BRDORAT BRDORAT BR DORAT weld i s sufficien t to promo t e TGSCC on the annealed type 304 s t a i nless s t eel ca n opy sea l in a co rr osive environm e n t. One type o f environment in which annealed stainless steels a r e known to be susceptib l e to TGSCC is in hig h tempera t ure, B R DORAT BRDORAT BRDORAT BRDORAT L a s t Upd a ted 03/20/2015 03/20/20 1 5 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/20 1 5 03/20/2015 CR Detail Repor t Page 3 of 28 1 2/10/20 1 5 1 1 : 42: 24PM Wolf Creek Nuclear Opera t ing Corporation 00093697 Condition Report stagnant chloride/oxygen env i ronments. This is associated with Head Adaptor Plug Canopies, which is l ocation of this leak. Westinghouse has recommended to perform i nspections of these locations to identify any telltale rusty colored spots, and white boric acid deposits and tracks. Thi s is what was occurring when this l eak was i de n tified. With the cause already identified and actions i n place , which identified this l eak, an apparent cause would not be beneficia l to this issue. This i ssue was discussed with the initiator. Oth e r R e l ated Inf ormation Assignment Status Summary: Total Assigns/Subs

Open Assigns/Subs:

Overdue Assigns/Subs: C r oss Refer e nces: 3 -23 0 -3 0 -1 Typ e MPAC WORK REQUEST Numb e r 15-111261 Status & Du e D ate H istory: Respon s ibl e P erson HALL , JOHN F HALL, JOHN F NEILSON , RHONDA G BELL , SETH A DORATHY, BRIAN D DORATHY , BRIAND D a t e U pdated 03/18/2015 03/18/2015 03/21/2015 03/19/2015 04/20/2015 12/10/2015 Evaluation/Check! ist BR DORAT BRDORAT BRDORAT BRDORAT BRDORAT BRDORAT BRDORAT BRDORAT BRDORAT BRDORAT Sub Number Status INPROG H/APPR APPROVED PRE-APRV COMPLETE BLL A ss ignment #: 00093697-02 Due D ate: 04/20/2015 St a tu s: COMPLETE

Subject:

STS PE-040E Penetra t ion 20 Canopy Seal Weld Leakage lndicati Age In Days: 30 A ss igned To N a m e: DORATHY , BRIAND A ssigned T o Organizat ion: 4050050-E NGINEERING DAILY WORK CONTROL SU PV -DORATHY 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/2015 03/20/2015 1 0/26/2015 Due D a te 04/20/2015 1 2/1 112015 St a tu s D a te: 04/20/2015 Total Ag e: 30.00 D escr iption: Perform a basic evaluation in accordance with AP 28A-100 and Al 28A-100. Use form AIF 28A-100-1 2, Basic Cause Evaluation. This evaluation does not require a qua li fied evaluator. Contact the CAP group for further assistance if determined this assignment is not needed. DO N O T status the assig nm ent as Complete. CR Detail Report Page 4 of 28 1 2/10/20 1 5 1 1 : 42:2 4PM Wolf Creek Nuclear Operating Corporation Condition Statement: E x tent of Condition: Op e rating E x peri e n ce: CR Detail Report 00093697 Condition Report PROBLEM STATEMENT: The canopy seal weld on penetration# 20 was found to be leaking during the performance of a remote bare metal v i sual examination (VE) of the Reactor Vessel (RV) head penetrations (S T S PE-040E performed during RF20). The indication appears on the east s i de of the seal we l d face. Boron staining is evident above, adjacen t and below the seal weld indication. Rust staining is also prevalent at and be l ow the ind i cation 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 sea l weld i ndication appears to have been an active leak in the recent past. CR INITIATOR CONTACT: HALL, JOHN F. The ISi Engineer involved in th i s evaluation was also involved i n the examinations that identified the canopy seal weld leak. Th e I Si Engineer thoroughly discussed this issue with the CR Initiator during and after the examinatio

n. This condition is l i mited 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 descr i ption of the design and configu r ation). This CR addresses all of these lower canopy sea l welds in the 78 RV head nozzle penetrat i ons. Therefore , no further ex t ent of condition considerations are applicable.

An Operating Experience searc h of the INPO website ident i fied a number of sim il ar i ssues at plants resu l ting in a leak i n the canopy seal weld area. In dustry experience using a canopy seal clamp assemb l y has been successful. The tabl e in the Evaluation section below lists a number of canopy seal leaks that have been successfully repaired at WCGS. As not e d i n WCAP-12088 , the transgranu l ar st r ess corrosion cracking of the canopy seal weld is a known industry issue. Some applicab le OE that can be used for referenc e, are: OE4046 (Indian Po i nt 3), OE15763 (I ndian Point 2), OE23609 (VC Summer), OE27722 (M i llstone 3), OE31028 (North Anna 2), Plant Event #40228 (Seabrook), 120388 (Wolf Creek). In addition, WCAP 12088 describes the results of failure ana l ysis of canopy seal weld leaks from S different plants that led to the c reation of the WCAP. Wolf Creek has also had history of leaks at the canopy seal weld , as ev i denced by the number of canopy seal clamps i nstalled (see the table in the Evaluation section below). Most of the canopy seal c l amps are i nstalled at spare penetration locations , whi c h see m to b e the most s us cep tibl e to the transgranular stress corrosion crac king. As id entified in WCAP-12088, the failure mechanism of the canopy seal weld i s known. Th e method of repair is also well-proven and has been successful throughout the industry. Pages of 28 1 2/10/20 1 S 1 1 :42:24 PM Wolf Creek Nuclear Operating Corporation E val u atio n and Con c lu s ion: CR Detail Report 00093697 Condition Report EVALUATION: The reactor coolant system (RCS) transfers the heat generated in the reactor core to the steam generators v i a the reactor coolant pumps. The RCS operates at of 2235 psig and at a co l d leg temperature of approx i mately 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 e leva ted temperatures and wet conditions, the bor ic acid found in the RCS coolant is highly corrosive to carbon steel (the material of constructio n 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 sta i nless stee l head adaptor flange. The stainless stee l 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 piece construction -an ln conel tube sec t ion w elded to a s t ainless steel (type 304) flange , referred to as the head adapter flange. The In con e l 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 des i gned 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. Th is 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 stee l flange on the end of the CROM penetrat i on and a canopy lip. The head adaptor is designed such that when the attachment is threaded onto the stainless steel flange (during the orig i nal constructio n), the two canopy lips come together. They are seal welded since the ASME Sect i on Il l Code states that threaded joints in whi ch threads provide the only seal shall not be used. H ence the canopy seal weld was provided to seal the ACME threads. It is i mportant to note that the ACME threads of the threaded connection provide the structural design st r ength and pressure boundary of the joint. The canopy seal weld provides l eakage co ntrol of the threaded connection, but does not provide any of the ASME Code strength of the connection. Th e canopy seal welds are thin we l ds of about 0.070" thickness that serve to seal the threaded press ure boundary connec tion. The canopy sea l weld configuration forms a " dead end" in which impurities in water that works i ts 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 installed , and the vessel is repressuri ze d , the trapped water in the sea l area is oxygenated. Thi s environment establishes conditions th at appear to increase the probability of stress corrosion cracking. Annea l ed stainless stee l s are known t o be susceptible to transgranular stress corrosion cracking (TGSCC) in this type of high temperature , stagnant chloride/oxygen envi r onment. Page 6 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation CR Detail Report 00093697 Condition Report Westinghouse performed destructive examination and hardware failure analysis on a number of lower canopy seal welds removed from nuclear power plants i n the late nineteen e i ght ie s (WCAP-12088). In addition, Westinghouse assisted WCNOC personne l in performing a hardware failure analysis report for leaking canopy seal welds (References 4 and 5). The se i nvestigations concluded that the failure mechanism was TGSCC. The cracking was id entified both in the sea l welds and in the base metal of the seal weld joints. All cracking was i nitiated from the i nterior of the j oint. No sensitization of the material was identified in the components examined. Very low levels of chloride contamination w e r e 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 residua l s tresses associa ted with the seal welding process, combined with the very low levels of ch l oride in the oxygenated stagnant region, were suffic i ent to promote TGSCC on the joint. The Westinghouse hardware failure analysis also i nc l uded examination of some threaded joints that were removed along with the l ower canopy seal welds. There was no ev i dence of corrosion or crack i ng on any of the threaded joints that were exam i ned. Austenitic s t ainless stee l in the presence of chlor i de/oxygen in the high pressure and temperature environment is susceptible to Transgranular Stress Corros i on 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 corros i ve 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 l ikely chloride, and oxygen contamination present i n the "dead e n d cav ity" 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. Th ere have been no industry reports of degradation of canopy seal welds resulting in significant l eakage flow rates (Ref. 3). Considering the h ead adapter f l ange des i gn, l eakage through a c r ack in the non-pressure boundary sea l weld wou l d be expected to be limited by the load carrying component , the flange connection thr eads. Based upon the WCGS and industry experience, along with the fact that the canopy seal weld is not the load carry i ng part of the joint design, a gross failure on a lower canopy seal weld is unlikely to occur. If a gross failure of a sea l weld does occur, it is expected that leakage would be recogn i zed us i ng i nd i cat i ons typical of a small leak inside containment and would be subject to the un i dentified leakage T echnical Specification limitations. The canopy seal associated with penetration

1. 20 was determined to be l eaking during the RV head inspection in RF20. Thi s penetrat i on i s one of the 8 for capped latch housings.

Historically, repairs us i ng a Canopy Sea l Clamp Assembly (CSCA) have been made to l owe r canopy seal welds at WCGS at the following locations: Penetration

1. Type 10 RV Leve ll Indication System 13 Spare 17 Ac tiv e CROM 22 Spare 24 Spare 25 Spare 27 Spare 28 Spare Page 7 of 28 1 2/10/20 1 5 1 1 :42: 24PM Wolf Creek Nuclear Opera t ing Corpora t ion CR Detail Report 00093697 Condition Report 29 Spare Th e CSCA is des ign ed to be installed remotely from above the n ozz l e housings. The installation can be accomplished with the reactor vessel head lo ca t ed e ith er in the r eactor head s t and or on the reactor v essel. The above lis t of canopy seal repa i rs has been performed at different times during the operating life of the reactor at WCGS. This evaluation and probable cause id entify 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, -lea kage is based on the ind i vidual cond itions associated w i th each canopy seal weld. A leaking canopy seal w i ll allow bor ic acid and radioact i ve contaminants to leak from the primary system. The boric acid i s non-corrosive to the piping and v essel cladding associated with the primary system (whic h contains the Borated Wate r). Bor i c acid leak ing onto the hot external surface of the reactor vessel could result in cor r osive damage. The extent of the damage depends on the extent , duration of the leak and moisture co nt ent of the residue on the carbon steel closure head. At t he b eg inning of eac h outage , a boric acid wa l kdown at Mode 3 is conducted in accordance w i th STN PE-0400. I ncluded in the STN PE-0400 in spect ion l ocations are the RV nozzle head adapter canopy seal welds. The STN PE-0400 inspection conducted at the beginning of RF20 did not i dentify the lea kage from the canopy sea l w e ld at head nozzle penetration
2. 20. It shou l d be noted that th is leak was discovered during performance of STS PE-040E, " RPV Head Visual In spection".

The method of visual in spect i on was by a remote controlled crawler performing a VE ins pection on the RPV head. In accordance with STS PE-040E, this VE inspection is cond uct ed every third refueling outage. Although i n spection of the canopy seal area i s outside the requir ed inspection scope, boric acid staining and a rust stai nin g w ere noticed on the RPV head surface. Thi s discoloration was trac ed to the overhead canopy sea l area. As noted a bove, a head visual in sp e c tion is scheduled for each outage as part of ST N PE-0 400 and a head visual examination is schedu l ed for each refueling outage (e ither a bare metal v i sual examination, i.e., VE, every third refuel i ng or a VT-2 with insulation removed at refueling outages between the VE). Th e boric ac i d walkdown inspections of STN-PE-0400 , along with the visual examinations of STS PE-040E, and the end of outage performance of pr essu r e test S T S PE-040A by a VT-2 examina tion at NOP/NOT , provide several opportunities for discovery of a leaking canopy seal before the RV head would be damaged by the corros i ve boric acid l iq u i d. Past ident i fication of l eaking canopy seal welds have been i dentified by one of these inspections. However, the small extent of leakage i ndicated by the id entified boric acid and ru s t staining at penetration

1. 20 h as id entified a w ea k ness in the abil i ty of these inspections and examinations to identify small leaks. Therefore , enhancements to one or more of these in s pections are determined appropriate to be able to identify sim ilar small leaks. When a suspected l eaking lower canopy seal we l d is identified fol l owing a visual examination as described above, WCGS has i nstalled a Canopy Seal Clamp Assembly (CSCA). The proposed repair method to encapsulate the canopy seal weld with the CSCA has been eva lu ated by Westinghouse and determ i ned to be an acceptab le repair method. Wolf Creek has previously successful l y performed this repair me thod on ac tiv e penetration canopy seal leaks a nd multip l e spare penetration canopy seal l eaks. However , this leak was the first identified on a capped latch assemb ly and WCNOC did no t have a canopy seal clamp assembly ap pr oved for instal l a t i on on Page 8 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Opera t ing Corpora t ion CR Detail Report 00093697 Condition Report this configuration.

Th e CSCAs are designed and fabr i cated as C l ass 1 components. I n the CSCA design, a large ring of Grafo i lŽ (graphite) sea l ant that is compressed against the canopy seal weld region provides the barrier against fluid leakage. The GrafoilŽ is held i n place by the seal carrier halves , top and bottom housings, and attachment cap screws. The CSCA is designed to be i nstalled remotely from above the CROM housings. Th e installation is typically performed with the reactor vessel head removed from the reactor vessel and on the head stand. Th is approach is cost effect i ve and minimizes the rad i ation dose during the repair process. For the leaking canopy seal we l d at penetration

1. 20, WCNOC prepared and i ssued CCP 012962, approving the use of a CSCA on penetration
2. 20 capped latch assembly.

A rev i sion to CCP 012962 will be required to approved CSCAs on any capped latch assembly. CONCLUSIONS

Leakage at these RV canopy seal welds is an i nherit des i gn and configuratio n problem caused by Tr ansgranu l ar Stress Corrosion Cracking. Such leakage has been i dentified and repaired in the past by insta l lat i on 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 ind i cated by the id entified boric acid and rust sta ining at penetration

1. 20 has identified a weakness in the abil i ty of th ese i nspect i ons and examinations to ident i fy small leaks. Therefore, enhancements to one or more of these inspections are determined appropriate to be able to identify similar small leaks. However, to prec l ude delays in repairing leaking canopy seal welds w i th CSCAs , WCNOC design documents need to have evalua t ed a nd approved CSCAs for all the identified configurations of attachments to the nozzle head adapters.

WCNOC Engineering needs to comp l ete 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 s t ainless stee l in the presence of chlor ide/oxyg en in the high pressure and temperature environment is susceptible to Transgranul ar Stress Corrosion Cracking (TGSCC). The residual stress associated with the canopy seal weld is sufficient to promote TGSCC on the annealed 304 stainless stee l canopy seal in a corrosive environment. The cause of the cracking observed in the annealed 304 s tainle ss steel canopy sea l is concluded to be TGSCC from a combination of corrosive media , most l i kely chloride, and oxygen contamination present in the "dead end cavity" that i s formed by the canopy seal (Reference 4). SUPPORTING FACTS: Westinghouse performed destructive exam i nation and hardware fa il ure analysis on a number of lowe r canopy seal welds removed from nuclear power plants i n the late nineteen e i ghties (WCAP-12088). In addition, Westinghouse ass i sted WCNOC personne l in performing a hardware failure analysis report for leaking canopy seal w elds (References 4 and 5). These i nvestigations concluded that the failure mechan i sm was TGSCC. The cracking was i dentified both in the sea l 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 mater i al was identified i n the components examined. Very low levels of chloride con t amination were noted in water samples obtained from removed weld joint areas, as well as residues on the threaded surfaces that w ere analyzed. Westinghouse determ in ed that the Page 9 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation Cause: Extent of Cause: Saf e ty Significance: CR Detail Repor t 00093697 Condition Report residual stresses associated with the seal we l ding process, combined with the very low levels of ch l oride in the oxygenated stagnant region , were sufficient to promote TGSCC on the joint. T he Westinghouse hardware failure ana l ys is a l so i nc l uded examination of some threaded jo i nts that were removed along w i th the l ower canopy sea l welds. There was no ev i dence o f corrosion or crack i ng on any of the threaded joints that were examined. The indus t ry 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 l eakage flow rates (Ref. 3). Considering the head adapter f l ange des i gn, l eakage through a crack in the non-pressure boundary sea l we l d wou l d be expected t o be limited by the load carrying component, the flange connection t hreads. Page 10 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation Actions Taken: Information Sources: R ev i ew and Appr ova l s QA Review: Rad Protection Review: Indepe n d e nt Review: CR Detail Report 00093697 Condition Report ACTIONS TAKEN: Engineering approved (Reference

2) the installation of a canopy clamp seal assemb l y at the capped latch penetration

1. 20. Add i tionally , Engineering approved (Reference

2) the installation of a new design " short" canopy seal clamp assembly as a d i rect replacement for the old design "long" canopy sea l clamp assembly.

The short CSCA can be installed in any l ocation where the long CSCA i s currently approved for installation . The proposed repair method to encapsulate the canopy seal weld with the CSCA has been eva lu ated by Westinghouse and determined to be an acceptable repair method. Wolf Creek has previously successfully performed th i s repair method on severa l active penetration canopy seal leaks and multiple spare penetration canopy seal l eaks. As of Jan. 15, 2008, West i nghouse information had identified 150 CSCAs i nstalled on both l eaking (56) and non-leaking (94) lower canopy seal welds on head penetrations in 35 plants s i nce 1988. Westinghouse a l so reports there have been no re-occurrences of leakage from previously l eaking welds (Reference 2). Th e " sho rt" canopy seal clamp assembly was success fully installed on penetration

1. 20. After installation of the CSCA, i t was found that the " dummy can" would not fit over the CSCA. The "dummy can" was modified prior to installation.

An inspect i on will be performed dur in g performance of pressure test STS PE-040A to determine if the CSCA installation was successful. ACTIONS PLANNED: 1) The small extent of leakage i ndicated by the identified boric acid a n d ru s t staining at penetration

1. 20 has identified a weakness in the ability of the normal i nspect i ons of the RV nozzle penetrations to ident i fy small l eaks. Th erefore , enhancements to one or more of these inspections a r e determ i ned appropriate to be able to identify similar small leaks. The following inspection procedures should be evaluated for enhancements to assure sma ll leaks a t the canopy seal welds are ide ntifi ed early in the outage to allow time for repairs (installation of CSCAs): STN-PE-040D , STS PE-040E , and STS PE-040A. 2) West i nghouse will be providing follow-up documentation that w i ll approve CSCAs on any capped latch housings. Engineering plans t o issue a revision to CCP 012962 to update the appropriate WCNOC documents after receipt of th e Westinghouse documen t ation. IN FORMAT ION 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 I ntegrity" (CT 02-0029).

2. Change Package# 012962 , " Canopy Sea l Clamp Assembly in Capped L atch Location" , 3/25/2015.

3. WCAP-12088 , "Metallurgical Failure Analysis of Leaking Canopy Seals" , C. M. Pezze (January , 1989). 4. Hardware Failure Analys i s Request (HFAR N o. MA 92-008 , March 1992). 5. Westinghouse letter SAP-92-148 transmitting Westinghouse Report MED-PCE-1 1788. Page 11 of 28 1 2/10/20 1 5 1 1 : 42: 24PM Wolf Creek Nuclear Operating Corporation CARB R e view: CAP Li a i s on: Sup v. Ap p r ova l: Supt. Approval: Man a ger App r oval: V.P. Appr ov al: CEO Approv a l: Extentions

1. of E x tention s: Ext e nti o n Note s: Supv. E x t. Appro va l: Supt. E x t. Approval: Manager Ext. Appro v al: V.P. E x t. Ap p ro va l: CEO E xt. App r oval: Oth e r R e l ate d Info r mati o n 00093697 Condition Report APPROVED BRDORAT -04/20/2015 The evaluation was performed by David Giefer and Richard Gimple. I t was i ntended to attached all the documents to this CR but t he folder in Net FYI is locked and wil l not allow any files to be added. Everything was copied into t he applicable categories.

The evaluation has been reviewed and approved with the required actions created and assigned. 0 A ss ignm e nt Not es: Updated By L as t Upd a t e d R efe r e n ce s: EVAL S t a tu s & Du e Dat e Hi s to ry: BRIAN D. DORATHY BRIAN D. DORATHY R H O N DA G. NEILSO N CR Detail Repor t 04/20/2015 04/20/2015 03/2 1/2015 Page 12 of 28 COMPLETE ACC/ASG INPROG 04/20/2015 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Plan and Actions Pl a n Assig in ment #: 00093697 -03 St a tu s: CO M PLETE Plan

Subject:

Penetrat i on 20 Canopy Sea l Weld Lea k age A ss i g ned To Name: DORATHY , BRIAN D As s igned To Organization

4050050-E N GI N EERI N G DA I LY WORK CO N TROL S U PV -DORATHY Description

Action Ass.ignment #: 00093697-03-01 Action Due Date: 11/20/20 1 5 Status: CA N C EL ED Action

Subject:

Revise STS PE-040E Assigned To Name: G I EFER, D AVID L Assigned To Organization

4 05 0 050-EN GI N EE R I N G D AIL Y W ORK CO N TRO L S U PV -D OR A T HY St a tus D a te: 1 2/10/20 1 5 Age In Day s: 234 Status Date: 05/20/20 1 5 Age In Day s: 30 Description

Revise procedu r e STS P E-0 4 0E to inc l ude a came r a on a po l e that cou l d b e manipulated within the CRO M tub i ng to i dentify canopy seal i ss u es. Action Category: REMEDIAL LTCA: S c h e dul e R e quir e m e nt: RCMS#: Commitm e nt: Commit To Ag e ncy: Work P e rform e d: R e vi e w and Approv a l s Indepe n dent Re v ie w: CARB R e vi e w: CAP Li a i s on: Supv. Approval: Supt. Approval: M a n a ger Approval: V.P. Approval: CEO Approv a l: Extension s # of E x ten s ion s: Ext e nsion Notes: Supv. Ext. App r o v al: Supt. E x t. Approv a l: CR Detai l R e port 0 Page 13 of 28 1 2/10/2 0 1 5 1 1 :4 2: 24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Man a g e r Ext. Appro v al: V.P. Ext. App r ov a l: C EO E xt. Approv a l: A c tion A ss.ignm e nt #: 0009 36 9 7-0 3-02 A c tion Du e Date: 12/11/2015 St a tu s: COMPLETE St a tus D a t e: 12/10/20 1 5 A c tion Subj ec t: Revise Inspection Procedu r e Ag e In Da ys: 234 As s igned To Name: DORATHY, BRIAN D Assigned To Organization:

4050050-E N GINEERING DA I LY WORK CO N TROL S U PV -DORAT H Y D es cription: T he small extent of leakage indicated b y the i dentified boric ac i d and rust staining at penetration

1. 20 has identified a weakness i n the ability of the normal inspections of the RV nozz l e penetrations to identify sma ll leaks . Therefore, enhancements to one or more of these inspections are de t erm i ned appropriate to be able to identify sim i lar small l eaks. The follow i ng i nspection procedures shou l d be evaluated for enhancements to assure small l eaks at the canopy seal welds are ident i fied early i n the outage to allow time for repa i rs (installation of CSCAs): STN-PE-0400, S T S PE-040E , and STS PE-040A. Action Category: REMEDIAL LTCA: Schedule Requirem e nt: RCMS#: Commitme*nt: C ommit To Ag e n c y: Work Perf o rm e d: y CR Detail Repor t The procedure that is required to provide a detailed inspection of t he RF head is STS PE-040E , " R PV H ead Visual Inspection

". T he other two procedures (STN-PE-0400 and STS-PE-040A) do not provide detai l ed inspections. Therefore , STS PE-040E revis i on should close* this a c t i on. The revision of the proced u re inc l uded prov i ding requ ir ed updating for the Interval 4 IS i Prog r am. I n addition. some word i ng corrections/modifications were made. The main additions to the procedure to support the CR 93697 resolution Includ e: 1) Addition of a note to include the r equ i rements of a genera l inspection of the areas above the head to Include the canopy seals (Note included between Section 8.0 and 8.1) as follows: "A l though not part of the pressure test inspe c tion , the visual i nspect i on (e i ther VE or VT-2) shou l d include a general inspection of the RV head areas above the CROM n ozzle/head interface penetrations that wou l d In c lude t he canopy sea l s. The results of these i nspect i ons shoul d be inc l uded i n Attachment A-Test Performers Comments" 2) A step to identify acceptance of the visual doc u mentation (Sections 8.1.3 and 8.2.3) as follows: "Visua l documentation (photographs and/or video) of examinat i on accepted by Boric A c id Corrosion Control Program Owner." 3) A step to assu re a good Inspection of the inner areas of the head that are d i fficu l t to view (S t ep 8.2.1 , I tem 1) as follows: Page 14 of 28 1 2/10/20 1 5 1 1 : 42: 24PM Wolf Creek Nuclear Operating Corporation Review and Approvals Independent Review: CARB Review: CAP Li a ison: 00093697 Condition Report " To inspect inner areas of the head that a r e difficult to vi ew for performance of a VT-2 inspec t ion, a remo t e viewing device (c a mera , etc.) mounted to an extension tool can be used. Record app r oximate az i muth of th e access points on Attachment D along with the results of the VT -2 exam using a remo t e viewing device." With the above revisions to Procedure STS PE-040E , the CR action 93697-03-02 should be closed. Supv. Appro va l: APPROVED Supt. Appr ova l: M anager Approval: V.P. App rova l: CEO Approval: Exten sio ns # of Extensions: 0 Ext e n sio n Note s: Supv. Ext. Approval: Supt. Ext. Approval: Manager Ext. Approval: V.P. Ext. Approval: CEO E xt. Approval: A c tion A ss.ignm e nt #: 00093697 03 A c tion Due D ate: 10/30/20 15 Sta tu s: COMP L ETE S*tat us Dat e: 10/26/20 15 Action

Subject:

Track R evision to CP 12962 Age In Day s: 189 As si gn e d To N a m e: DORATHY, BRIAN D Assigned To Organization

4010020-CIVIL

-ME C HAN I CAL DESIG N ENG -CURTE N De scr iption: Westinghouse will be providing follow-up documentation that will approve CSCAs on any capped l atch hous i ngs. E ng ineering p l ans to issue a revi sion to CCP 012962 to updat e the appropr i ate WCNOC documents after rec e i pt o f the Westinghouse documen t ation. Action Category: E N HANCE M ENT LTCA: Schedule Requirement

RCMS#: CR Detail R epo rt Page 15 of 28 1 2/10/20 1 5 1 1 : 42: 24PM Wolf Creek Nuclear Operating Corporation Commitment:

Commit To Agency: Work Performed: Review and Approval s Independent Review: CARB Review: CAP Liaison: Supv. Approval: Supt. Approval: Man age r Approval: V.P. Approval: CEO Approv a l: Ext e nsions # of E xtensions: Ext ens ion Note s: Supv. Ext. Approval: Supt. Ext. Approval: Manager Ext. Approval: V.P. E xt. Approv a l: CEO Ext. Approval: y 00093697 Condition Report APPROVED 0 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 actio n is required. 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 prior i ty items preventing t he review of the change package. The RF21 design mi l estone and WANO preparations have taken a priority. This is capturing information provided from Westinghous e i n WCNOC documents that wil l not be needed until the next refueling outage, whi ch the extension due date is wel l before the next outage. There are not any hazards assoc i ated with this extension since it wi ll s till be completed prior to next outage. There are no interim actions requir ed for this extension. This action is not add re ss i ng the restoration of full qualification of an SSC. APPROVED BRDORAT -09/17/2015 Peer rev i ew provided by John Ashley. 9/17/15 Oth er Related Pl a n and Action Information CR Detail Report Page 16 of 28 1 2/10/20 1 5 1 1 :42:2 4PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Plan A s signment Note s: 00093697-03 93697-03-01 was created in error. Changes cou l d not be made after the action was progressed to no t ify so the action was cance l ed. Action Ass.i gnment Notes: 00093697-01-01 Action auto-closed based on work comp l etion for WO 1 5-399448-000. 0009 3 697-01-02 Action auto-closed based on work comp l etion for WO 15-3 99448-001. 00093697 0 3 Action auto-closed based on wo r k comp l etion for WO 1 5-399448-002. 00093697-01-04 Actio n auto-closed based on wor k comp l etion for WO 1 5-399448-003. 00093697-01-05 Ac t io n au t o-closed b ased on wo rk comp l etion for WO 15-39 94 48-004. 00093697-01 -06 Ac t io n a ut o-closed based on wo rk comp l etion for WO 15-399448-005. 00093697-01 -07 Action auto-closed based on work comp l etion for W O 15-399448-006. 00093697-01-08 Action auto-closed based on work comp l etion for WO 15-399448-007. 00093697-01 -09 Action auto-closed based on work comp l etion for WO 1 5-399448-008. 00093697-01-10 Action auto-closed based on wo r k comp l etion for WO 15-399448-009. 0009 3 69 7-01-11 Actio n auto-closed based on wor k comp l e t ion fo r WO 15-399 4 48-010. 00093697 12 Actio n au t o-closed based on wor k comp l etion for WO 1 5-399448-011. 00093697-01-13 Action auto-closed based on wor k comp l etion for WO 1 5-399448-012. 00093697-01-14 Action auto-closed based on work comp l etion for WO 15-399448-013. 0009 3 697-01-15 Action auto-closed based on wo r k comp l etion for WO 15-399448-014. 0009369 7-01-16 Action auto-closed based on wo r k comp l e t ion for WO 15-399448-015. 0009 36 97-01-17 Action auto-closed based on work comp l e t ion for WO 15-399448-016. CR Detai l Report Page 17 of 28 Updated By La s t Updated BR DORAT 05/20/20 1 5 INDUS 04/17/2015 INDUS 03/29/2015 I N DUS 03/24/2015 I N DUS 0 4/15/2015 IND US 03/3 1/2015 I NDUS 03/31/2015 IN DUS 03/22/2015 INDUS 03/30/2015 IN DUS 03/22/2015 I N DUS 04/10/2015 IN D U S 05/0 1/2015 I N D U S 03/30/2015 IN DUS 03/20/2015 I NDUS 04/09/2015 IN DUS 03/21/2015 IN DUS 03/21/2015 INDU S 03/29/2015 1 2/10/20 1 5 1 1:42:24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report 00093697-01-18 Action auto-closed based on wor k comp l etion for WO 15-399448-017. 00093697 01 93697-03-01 was created in error. It was unknown that changes could be made after t he action was progressed to notify so the act i on was cance l ed. Plan Compl e tion Not es: Action Completion Notes: Pl a n Cro ss R e f e r e n ce: T y p e A c tio n Cr oss R e fer e n ce: Plan Status and Due Date History: 00093697 -03 R es pon s ibl e P e r s on DORATHY , BR I AN D DORATHY , BR I AN D DORATHY, BR I AN D A c tion St at u s a nd Du e D a t e Hi s t o ry: 0009 3 69 7 -0 3-01 Responsible Person DORATHY , BR I AN D DORATHY , BR I AN D DORA T HY , BR I AN D DORATHY, BR I AN D DORATHY , BR I AND DORATHY , BR I AN D 000936 9 7 -0 3-02 Responsible Person DOR A THY, BR I A N D DORATHY , BR I AND GIEFER, DAVID L DORATHY , BR I AN D DORATHY, BR I AN D DORATHY, BR I AN D 0009 3 697-0 3-03 R es po ns ibl e P ers on DORATHY , BR I AN D DORATHY , BR I AN D DORATHY , BR I AN D PANKAS KI E , JASO N M DORATHY , BR I A N D DORATHY, BR I A N D CR Detail Repor t Dat e Updat e d 04/20/20 1 5 04/20/2015 12/10/2015 Date Updated 04/20/2015 04/20/2015 04/20/2015 05/20/2015 05/20/2015 05/20/20 1 5 Dat e Updated 04/20/2015 04/20/2015 12/10/2015

12/10/2015 12/10/2015

12/10/2015 D a t e Upd a t e d 04/20/2015 04/20/2015 09/17/2015 10/08/2015 10/26/2015 10/26/2015 Page 18 of 28 Numb e r S t a tu s INPROG ACC/ASG COMP L E T E Statu s I NPROG NTFY/ASG CANCELED I NPROG N TFY/ASG CANCELED Status I NPROG NTFY/ASG ACC/ASG N T F Y/ASG ACC/ASG COMPLETE St a tu s I NPROG N TFY/ASG ACC/ASG NTFY/ASG ACC/ASG INDUS 03/29/2015 BRDORAT 05/20/2015 Sub Num be r Due Date 1 2/1 112015 Due Date 1 1/20/2015 Due Dat e 1 2/1 112015 D ue D a t e 09/1 8/2015 1 0/3 0/2015 1 2/10/20 1 5 1 1 : 42: 24PM Wolf Creek Nuclear Opera t ing Corporation 00093697 Condition Report DORATHY , BRIAN D 10/26/2015 COMPLETE CR Detail Report Page 19 of 28 1 2/1 0/20 1 5 1 1 : 42:24PM Wolf Creek Nuclear Operating Corporation EFU Assignment#: EFU

Subject:

Assign ed To Name: Assigned To Organization:

== Description:==

EFU Effect ive: Review and Approv als Independent Review: CARB Review: CAP Liaison: Supv. Approval: Supt. Approval: Manager Approval: V.P. Approval: CEO Approv a l: Extensions

1. of Ext ensions: Extension Notes: Supv. Ext. Approval: Supt. E xt. Appro va l: Manager !Ext. Approval:

V.P. Ext. Approval: CEO Ext. Appro va l: Other Related Information Assignment Not es: CR Detail Report 00093697 Condition Report Effectiveness Follow-up EFU Due Date: Status: Status Date: Ag e In D ays: Upd ated By La st Upd a t ed Page 20 of 28 1 2/10/2015 1 1 :42: 24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Cross R efere n ces: EFU Status a nd Due D ate Hi s tory: CR Detail Report Page 21of28 1 2/10/20 1 5 1 1 :42: 24PM Wolf Creek Nuclear Operating Corporation RTFQ 00093697 -01 Status: COMPLETE RTFQ

Subject:

RTFQ Description

Equipm ent: On-Line or Refu e l: CR Detail Report 00093697 Condition Report Restore to Full Qualification Status Date: 05/07/2015 WR#: 15-11 1261 STS PE-040E Penetration 21 Canopy Seal Weld Leakage lnd icati During the performance of STS PE-040E evidence of l eakage 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 s t aining i s evident above, adjacent and below the sea l we l d indication.

Rust staining is also prevalent at and below the indication running down the CROM nozz l e and onto the top head surface. There i s no evidence of degradation to the CROM or closure head surfaces. This sea l 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 sea l weld clamp assemb l y. RBB01 REFUEL Page 22 of 28 1 2/10/20 1 5 1 1 :42: 24PM Wolf Creek Nuclear Operating Corporation Operability

3 OPER/DNC CR Detail Report 00093697 Condition Report The initiator id entified during the performance of STS PE-040E ev i dence of leakage at the canopy seal weld on penetrat i on 21. The indication is on the east side of the seal weld face (as the RPV closure head sits). Boron staining i s evident above , adjacent and below the sea l weld indication.

Rust staining is also prevalent at and be l o w the indication running down th e CROM n ozz l e and onto the top head surface. There i s no evidence of degradation to the CROM or closure head surfaces. I reviewed the pictures that are located at K:\Data\NDE\Photos \R F-20\CR D M Head Inspection. I n 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 penetrat i on directly be l ow the subject seal weld. Technical Spec i fications defines Pressure Boundary LEAKAGE as LEAKAGE through a nonisolable fau l t in a n RCS componen t body , pipe wall or vessel wall. TS 3.4.13 conta i ns the ope r ating 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 m i nute, and primary to secondary L EAKAGE is limited to 150 gallons per day i n any one Steam Generator. The Control Rod Drive Mechanism is what is used to raise , lower, and trip control rods. The i nterna l s of this mechanism i s exposed to RCS pressure. The Drive Mechan i sm Latch Housing is internally threaded and torqued down onto a seating surface at the interface between the housing and the top of the Reactor H ead Adapter. This connection is a mechanical j oint and le akage via this pathway is not Pressure Boundary LEAKAGE as defined by Technica l Spec ifi cations. The WCGS reactor vessel head and CROM assemblies are class ifi ed 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 w ere designed and fabricated to the 1974 through W i nter 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 st r uctural part of the p r essure boundary and i s not r eq uired to meet the structural requirements of ASME B&PV Code , Section Ill , NB-3 000. The threads are the load ca rrying 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 i n leak rates that are bound by the limits establ i shed in Technical Specification 3.4.13. Completed performances of STS BB-006 were rev i ewed from the l ast operating cyc l e and RCS leakage limit s were not challenged. The Reactor Vessel and the subject CROM is OPERABLE but degraded due to the flaw i n th e lowe r seal weld. Page 23 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation Operations Fo c us Li s t: Plant System: Ri s k Impact: Risk Revi e w Complete: Risk Significance

BB y H I GH 00093697 Condition Report

References:

Technical Specifications 1.1, 3.4.13 and Bases; TR 3.4.17 and Bases; NRG In spect i on Manua l Part 9900, WCGS Correspondence CT 02-0029, Westinghouse Instruction and Operating Book for Magnetic Control Rod Drive Mechanism for Full-Length Contro l Rods , and STS PE-040E. WABRAND 04/04/2015 Safety Function: RCS Integrity IOA Conclusion

A clamp has been installed. QC will perform PMT at NOP. IOA: Sources CAP: Work Orders: M arg in M a nag e ment: Ops Focu s Li s t: Single Point Vulnerability:

System He a lth Report: T e mporary Modification: Operational Decision Making: Maint ena nc e Rul e: MSPI: PDM W a t c h Li s t: Regulatory Commitment: Other: Vulnerabilities Steam Gen e rator Tub e Rupture: L oss o f Off-Sit e Power Rapid Lo ad R e duction: Inadv e rt en t Safety Inje c t ion: Fire/Flooding: CR Detail Report Page 24 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Inter-system LOCA: Loss ofRHR: Lo ss of $.pent Fuel Cooling: Lo ad Rej ect: Steam Line Break: Lo ss ofESW: Measures Compensatory M eas u res: Monitoring Measures: Mitigations Measures: RTFQ Actions: CR Detail Report Page 25 of 28 1 2/1 0/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report 00093697-01-12 Status: COMPLETE WO#: 15-399448-0 1 1

Subject:

Remove/Replace cables and unistrut on head as requ i red to su Notes: Action auto-closed based on work comp l et i on for WO 15-399448-011. 00093697-01-16 Status: COMPLETE WO#: 15-399448-0 1 5

Subject:

PE has request a SWO for new stock item that is SR stock i te 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 complet i on for WO 15-399448-005. 00093697-01-14 Status: COMPLETE WO#: 15-399 448-013

Subject:

Contingency -Access lower shroud to assist c l amp installat i Notes: Action auto-closed based on work complet i on for WO 15-399448-013. 00093697-01-15 Status: COMPLETE WO#: 15-399 448-0 1 4

Subject:

NS92250263 i s not tied to asset RBB01 BOM Notes: Action auto-closed based on work comp l et i on for WO 15-399448-0

14. 00093697-01-02 Status: COMPLETE WO#: 15-399448-001

Subject:

STS PE-040E Penetration 2 1 Canopy Seal We l d L eakage lndicati Notes: Action auto-closed based on work comp l e ti on for WO 15-399448-001. 00093697-01-01 Status: COMPLETE WO#: 15-399448-000

Subject:

STS PE-040E Penetration 2 1 Canopy Sea l We l d L eakage lnd icati Notes: Action auto-closed based on work comp l et i on for WO 15-399448-000. 00093697-01-20 Status: CANCELED WO#: 15-399448-019

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Res i due on CROM nozzles , adapters and housings During RF20 Notes: 00093697-01-13 Status: COMPLETE WO#: 15-399448-0 1 2

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Engineering to evaluate the use of the following NS Stock It Notes: Action auto-closed based on work complet i on for WO 15-399448-012. 00093697-01-18 Status: COMPLETE WO#: 15-399448-0 17

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Engineering to evaluate !rim i ng the dummy can for capped lac Notes: Action auto-closed based on work comp letion for WO 15-399448-017. 00093697-01-11 Status: COMPLETE WO#: 15-399448-0 1 0

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QC i s requested to perform a Pre-Serv i ce VT-2 examination on Notes: Action auto-closed based on work comp leti on for WO 15-399448-010. 00093697-01-10 CR Detail Report Page 26 of 28 1 2/10/20 1 5 1 1 :42:24PM Wolf Creek Nuclear Opera t ing Corporation 00093697 Condition Report Status: COMPLETE WO#: 15-399448-009

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QC is requested to perform a Pre-Serv i ce VT-3 examination on Notes: Action auto-closed based on work complet i on 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-Serv i ce VT-1 examination on Notes: Action auto-closed based on work comp l etion 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 comp l etion 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 comp l etion for WO 15-399448-003. 00093697-01-03 Status: COMPLETE WO#: 15-399448-002

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QC is requested to p e rform a PT exam i nat ion on the Canopy Se Notes: Action auto-closed based on work comp l et i on for WO 15-399448-002. 00093697-01-05 Status: COMPLETE WO#: 15-399448-004

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Remove/Reinstall Upper Plenum to support insta ll ation of a c Notes: Action auto-closed based on work comp l et i on for WO 15-399448-004. 00093697-01 -07 Status: COMPLETE WO#: 15-399448-006

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QC i s requested to verify/document the Canopy Seal We l d leak Notes: Action auto-closed based on work comp l et i on 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 clean i ng Dur i ng th Notes: 0009369 7-01-CR Detail Repor t Page 27 of 28 12/10/20 1 5 1 1 :42: 24PM Wolf Creek Nuclear Operating Corporation 00093697 Condition Report Reportabllity Evaluation Report RER Status: Status Date: Age: Du e Date:

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D ate/Time of Di scovery: D escription: SCREENING/NOTIFICATIONS (Completed by Shift Manager): Potentially Reportable: RER Number: Per (list applicable r eport ing criteria met): Person Contacted: Corporate Services Notified: ENS Reportability D etermination per 10 CFR 50.72: ENS Wo rkshee t completed and attached: Continuous open c hann e l required: Shift Manager Approval: Last Updated: DISPOSITION (Completed by Licensing): LER#: Ltr. N l umber: Submittal Date: Event Evaluation: Reportability Evaluation Performed by: REVIEW and APPROVAL (Non-Reportable Events Only) Supervisor Licensing Approval: Last Updated: Manager Regulatory Affairs Approval: L as t Upd ated: ENS Retraction needed: Report Crlterl a CR#: 00093697 CR V i s i ble: y EVAL Visib l e: y PLA N Visible: y EFU Visible: y Non QA Visib l e: N RER V i sib l e: y C R D e t a il R epo rt P age 28 o f 28 1 2/1 0/20 1 5 1 1 :42:24 PM From: To: Sub j ect: Da te: Gentleme n , Reimer. Lisa Sjnga l Ba l want: Alley Dayjd; Werne r Greg; Pascare l li Robe rt FYI: Info: Interna l NBC C all to D is cu ss Wo l f Creek Rel i e f Reque s t Fr i day, October 28, 2016 12:43:00 PM George gave Miche l e t h e heads u p that the W o lf C r eek R e li ef R equest scope may need t o be expan d ed. She is leav i ng t he off i ce t oday at 3:3 0 pm, bu t is fi ne w ith a n y b r iefi n g i n fo rmat i on to be sen t to her b y ema il. Pl ease cc: Geo r ge W i l so n as we ll -h e a l so had to l eave early to d ay, b u t is avai l ab l e by p hone. George ce ll: .... r_l<5_l ___ ___. George home: 1 .... (b-)(6-) ___ ..... I am a l so availa bl e in t he o ffice unt i l 3 pm t oday, a n d by cel l a ft er: 410-733-965 9 T ha nk s! Li sa Lisa Regner Sr. PM N RR/DORL/L PL4-1 301-415-1906 08D08 From: Coll i ns, Jay Sent: Fr i day, October 28, 2016 11: 34 AM To: Singal, Ba l want <Ba l want.Singal@nrc.gov >; Al l ey, David <Dav i d.Alley@nrc.gov>; Kal i kian, Roger <Roger.Kalikian@nrc.gov>; Tsao, John <John.Tsao@nrc.gov >; Drake, James <James.Drake@nrc.gov >; Taylor , Nick <Nick.Tay l or@nrc.gov >; Proulx, David <Dav i d.Prou l x@nrc.gov>; Cumbl i dge, Stephen <Stephen.Cumb l idge@nrc.gov>; Regner , Lisa <Lisa.Regner@nrc.gov>; Werner, Greg <Greg.Werner@nrc.gov>; Anchondo , Isaac <lsaac.Anchondo@nrc.gov>; Kopriva , Ron <Ron.Kop r iva@n r c.gov>; Thomas, Fabian <Fabian.Thomas@nrc.gov>

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RE: I nterna l NRC Call to Discuss Wo l f Creek Relief Request ASME Code Case N-729-1 Note 1 T h e VE sha l l consist of the fo l lowing: (a) A direct exam i nation of the bare-metal surface of the entire outer surface of the head, including essentially 100% of the i ntersection of each nozzle w i th the head. If welded or bolted obstructions are present (i.e., mirror insu l ation, insu l ation support feet, shroud support ring/l ug), the exam i nation shal l include =95% of the area i n the reg i on of the nozzles as defined i n Fig. 1 and the head surface uphill and downh i l l of any such obstruct i ons. The exam i nation may be performed w i th insulat i on in place us i ng remote equipment that provides resolut i on of the component metal surface equ i va l ent to a bare-meta l d i rect examinat i on. (b) The examination may be performed with the system depressurized. (c) T he examinat i on shal l be performed with an i l lumination level and a sufficient distance to al l ow resolut i on of lower case cha r acters not greater tha n 0.105 i n. (2.7 mm) in height. -----Or i g i na I From: Singal, Ba l want Sent: Friday, October 28, 2016 10:17 AM To: Col l ins, Jay; Al l ey, David; Kalikian, Roger; Tsao, John; Drake, James; Taylor, Nick; Proulx, David; Cumbl i dge, Stephen; Regner, Lisa; Werne r , Greg; Anchondo, I saac; Kopriva, Ron; Thomas, Fab i an

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Interna l NRC Cal l to Discuss Wolf Creek Re l ief Request When: Friday, October 28, 2016 11:00 AM-12:00 PM (UTC-05:00) Eastern Time (US & Canada). Where: Dave Al l ey's Office Dave Alley and me received a cal l from Wolf Creek (Cyndia and Jaimme McCoy) at 9.30 this morn i ng. An interna l NRC staff meeting is required to discuss path forward based on informat i on prov i ded dur i ng the cal l. Bridge No. I nfo. 866-624-3402 Passcode: 1 ..... (b-)1 6_) _ ___. Li sa: You wi l l need to use Passcode .... 1 15_)16_) _ _.I (as i nitiator of the ca l l). I was not bale to search for conference rooms from ho me. I will be out-of-office .... l (b-)(_6) _______ 1 for about 3 hours and Lisda wi l l be supporting this ca l l. I can be contacted .... '.b_l (_5 l ____ I for any questions. Thank s. ..c I , ro a_ (].) I , C) c ro ID 76 E (].) en _J en .S:2 h en (].) <( E :J Ultrasonic inspections performed to see flaws in welds and penetration tubes need to scan above and below the weld, 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 the interference fit than leaking nozzles. --. . --: A -& fit ----. -.,. . . interference !ff:.- No Leak Leak 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. Very Little refl ec ti on ?% ....-( ( ( ( ( ))))))))))!) ) ) ) ) :: 0% Weld ))})))))))))))))))))))))))

100% * .... ,,,,..--(((((((((_! .A.ir )))))))))M Tot a l reflection

... Cla d d in g (Stain l ess St Butter (Alloy 82/182) 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. . U..**-) .... , ............ ............... --*"*-... -* l..,... .""""' .. ' .. -* ..,.,,, --. "'*'-J .. .............. ......... fl:** ,..,.._,.... ---0 to 170 deg. Circumference 0 ,..... 0 00 0 3 3 )> x J l:...j Q) 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. We s tin g h o u se D ata P NNL 2.25 MHz D a t a Wetted Side Wetted S i de 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. 1 35 Deg r ees 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 Asse ss ment 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)}}