RA-14-0015, Duke Energy Carolinas, LLC (Duke Energy): 10 CFR 50.46 - Annual Report for 2013 for Catawba and McGuire Nuclear Stations

Duke Energy Carolinas, LLC (Duke Energy): 10 CFR 50.46 - Annual Report for 2013 for Catawba and McGuire Nuclear Stations

ML14205A287
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	07/07/2014
From:	Waldrep B Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-14-0015
Download: ML14205A287 (14)

Text

S8D fl c. wd*g w

ElidERGYvice peý;ýPt Co*fp"!Ot Gr-* ce

& Ckwnatlons sumtqp 52f, S. GIur-t~Strewir Wi~ng Aodiusa:

ECO7H 'P OBox 1006 Clv.,,tMeN- 28201-1006 704-4824182*

Serial: RA-14-0015 10 CFR 50.46 July 7, 2014 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-.001 Catawba Nuclear Station, Units I and 2 Docket Numbers 50-413 and 50-414/Renewed License Numbers NPF-35 and NPF-52 McGuire Nuclear Station, Units I and 2 Docket Numbers 50-369 and 50.37(oRenewed License Numbers NPF-9 and NPF-17

Subject:

Duke Energy Carolinas, LLC (Duke Energy): 10 CFR 50.46 - Annual Report for 2013 for Catw*ba and McGuire Nuclear Stations

Reference:

1) Letter, 8. C. Wadrep (Duke Energy) to USNAC, Report ruam to 10 CFR50.46, Changes to or Errorsin an Evautmion Mode/, February 25, 2014 (ADAMS Accession No. ML14063A278) 10 CFR 50.46 (aX3)(ii) requires the reporting of changes to or errors in Emergency Core Cooling System (ECCS) evaluation models (EMs). Thig report cors the time period from January 1, 2013, to December 31, 2013, for the Catawba Nuclear Station (CNS) and the McGuire Nuclear Station (MNS). The most recent 10 CFR 50.46 report for CNS and MNS (Reference 1)provided the impact to Peak Cladding Temperature (PCT) due to an error inthe cladding burst strain model, which affects the Large Break Lou of Coolant Accident (LOCA) analysis of record for CNS and MNS. The PCT impactsrepoed in Ref*rnce 1 are inClud in this report for completeness.

Several other changes/errors were associated with the Large Break LOCA and Small Break LOCA evaluation models during the reportn period. The specific details of these changes/errors are provided in Attachment 1, Table 1. These changes/errors were evaluated by Westig s as having no impact on the calculated PCTs. Since there was no PCT impact due to these changes/errors, they are not included inthe LOCA PCT margin summary sheets, Tables 2 through 4, for MNS Units 1 and 2 and CNS Units I and 2.

U.S. Nuchr Reguar Commisson RA-14-00115 Page 2 There are no new regulatory cmmtmmt catned In this latter.

Please address any cmmsent or qtstio regarding t materto Michael K. Leisure at (980) 373-3619 (MlkeLelsur*Odiw-onogy.oom).

Sincerely, Be*mki C.W CommposeGovernane S OPeuulOn SwrPoit A#ttacmnt 2013 Annual Repoet of Peak Cladding Temporatr

U.S. Nuclear Regulatory Commission RA-14-0015 Page 3 xc (with astachment):

V. M.McCree Regional Admchsranor U.S. Nuclear Regu4a y Co~mtsson - Region 11 Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 G. Edward Miller NRC Protect Manager (Catavba & McGuWr)

U.S. Nuclear ReuAo Commiss One White Flint North, Mal Stop W"M 11555 PRodmile Pke Rockvlle, MD 20852-2738 NRC Senior Resident Inopeco Mc(ure Nuclear Ston NRC Senior es t Inspector Oamwba Nuclear Ston j, "-I

U.S. Nuclea Regulatory C~ommlm~on RA-14-0015 Attachmenit pape lof 11 Otis'; Qa a, L WPj l,nrg Cooing Syaeme for UgtW*tsr Nuchr Omm Ifhsew 2013 Annal ftp of PealtCladding Tempermtwo TableIt EwmmI Evsblon Model Cbange Tabl

• LOCA Peak Cbdiikitme e anolswumwy - WoQ** 10"Ift I a Tabl 3 -1 OCA Aedi Cadding 0"ysatr &Mrs% Suseay - Catawbe Una I Tabl Clddig 4-LOCTe~emur Pee sgi ~ v ~1~- Uit J

U.S. Nuckl Regula Commission RA-14-0015 Attachment Page 2 of 11 Errors I vEklol Mode Changes Oww" Codi Ma nWe Affected Evaluation Model(s): 1996 Westinghouse OWs Estimat Large Break. LQCA, Various changes have bean made to ertimce the usability of -odw and to strWMI*ne future anye. Examples of these chnes nclude modifi kWAt vW41a00 definitions, units and defaults; ipoving the input diagnost enanin the oode output optimizing active checks;n, coding; and Ol*%Ona irnicive coding. These changes represent Oiscket ry Chnes that will be i m on a forward-fit basis.

The nature of thee code changes leads to an estimated peak cladding temperature Impact of OOF.

-ENona for "meo To"m tiO uea Affected Evaluation Model(s): 1995 Westinghouse OB Estimate Large Break LOCA An envr was discoveredý inWCORAA(TRA whereby an incorrect value would be used inthe inta fuel rod temrperture cam*Mt for a ,*rohat tanse node if # r4damlevation was specified outside of the bounds of the tempokr~iaeinitiatiat~o tabl. Tbi* problve hasbeen evaluated for impact on existi analyses, andfsrsolutionreorss a VIscretonary Chenge.

Based on Inpection of plant arnalyis iaut. itwas cornc!d9edW the inpx eck for existing analym are not impacted by this error, leading to,an 1s*.ped!.k cadbding tenperature Impact of 0F.

fetTrWAOsr LOgi Cer"Wsol fer a"tBu" C III*eatOf Affected Evaluation Model(s): 1008 Westing1ho Beet Eme Large.Break LOCA Achange was made to the WCOBRPAJTRAC coding to cowrrc an ermo, which had disabled rod burst Inseparate effect test simulations. This change, represents a Discretionary Change.

Based on the nature of the change and the evaluation model requireenofts for plant modeling in We*irouse bes estimaet large break LOCA analyses with ILCODBRA/TR , it is judged that existing analyses are not impacted by this change, ladg to an estimated peak cladding temperature impac of O0F.

U.S. Nuclear Regulatory Commission RA-14-0015 Attachment Page 3 of 11 WCOBRA/TRAC U19 Fle DR non*re' aCrection Affected Evaluation Model(s): 1996 Westinghouse Best Estimate Large Break LOCA A problem was identified in the dimension of an array used to generate the u09 0le in WCOBRA/TRAC. The u19 file Is read during HSDRIVER execution and provides information needed to generate the HOTSPOT thermaydraulic history and user input fles, The array used to write the desired information to the ul9 file i dimensioned to 2000 in WCOBRA/TRAC.

It is possible, however, for more than 2000 curves to be written to the u19 fle. If that is the case, it is possible that the curv would not be stored correctly on the u19 file. A survey of current Best Estimate Large Break LOCA analyses Indicated that the-majority of plants had em than 2000 curves in their ul9 files; therefolr, these plants are not affected by the change. For those plants with more than 2000 curves, plant-specific sensit calculations indicated that resolution of this issue does not impact the peak cladding temperature calculation for prior analyses. This represents a Discretionary Change.

Resolukion of this issue does not impact the peak cladding temperature calculation for prior LBLOCA analyses, leading to an estimatedpeak cladding temperature impact of O°F.

Weat Tnefer Model EnOr Comllene Affected Evahaton Model(s): 190 Westinghouse Best Estimate Large Break LOCA Several related changes were made to WCOBRAMTRAC to correct errors discovered, which affected the heat trnsfer models. These errors Incluled calculation of the entrainedlu fraction used in calulation of the drop wall heat flux, qap on of the grid enhancement factor for grid temperature calculation, calsulation of the Reynolds number used in the Wong-Hochrleter carelation for the heat transfer coefficient from fuel rods to vapor, fuel rod initialzation and calculation of cd inner radius with creep, application of grid and two phase enhancemen factors and radiatio o in single phas vapor heat transfer, a rese of the critical heat flux temperature when J4. These errors have been evlaed to estimate fte kpect on etbsing LBLOCA analysis resulls. Corection of tse erromrs represents a closely-reated group of Non-DIscretionary Changes.

Base on the resuft of 1representaive plat cali" . spM te effects and integral effects test sWimu , it is cornckded that the error corrections have a negligible local effect on heat transf, leadi to an estimate peak cld temperature Inpc of O0F.

U.S. Nucler RegWula Comfmson RA-14-0015 Attachenet Page 4 of 11 Crevlon to Heat T-mostr Nod&taltlatlzaton Affected Evaluaton Models): 1996 WstInghd u Best Estirmae ge Break LOCA An error was discovered InIe hea transfer nods nlxo bog in MCOBWRTR" ,

whtreby Ote heat tnser node enter locations ould be iconstt vh the Uometi node center elevations. The prmary feft of fths iuea eon the htepolelu propewte d grid tJwbolte mixing enhlancemnt X0heat atI trIslfer"ode, Thi problem has been evaltd for imact on exakig anays, and ts reouMn rpmetisa n-nCieretoy Chag.

Basd on en*gei udgme and the rest frm a m-at-x of repese*ai plant calculations, t is co*ckde Vt te efec of this error Was' n#0 oodeslto , leading to an estimted peakKcladding twemueratmr inpect of 0..

ume Connesr-mdon EOw Ft Affecltd Evaltn Mods): 199 We ghouse Best Estimat Large Breek LOCA It was Idni tha mass was not onserved in WCOSR/iTRAC ons-dinsonslr copnn osb when void baction vaesm wer cauae to be sghl dt ephys raem (gmter than 1.0 or smaller twh 0.0). This ws observed to relt i tl mageneation on the seondalI y side of stem gweerato mponents. *tis Coyrftn of probem rprets a Non-Oiscrnnay Change.

Tthi error obsorvd to prikarly a the mass n ft seowifay e of ft seam genera. Ti suem an ued to ha negigals b eon slmi MOM analysi, results laing to an esoliMaed peak clddingtemperatue iMpac of 0FF Cmoweolion to SONt Canne wm~me a Eutnow Affected Evaluftin Models) 1996 Westinghouse I"a Estbimale LareBra LOCA An eror was disovmd inMe mment o*qation cu igs fo spli channe In WCOSRArRA. Thi serrinipc the*(I orbutya of the Ifharomtasuncwy boftio cell; (2) bofto and top connuy m iniý for the cfn nl at the boto of a sectik and for the chael outlet at t top of e and(3) op eralmem a rte per unit volm and & drop unit m = Won to fth mew n mbution caWati for sp* chane. This prdom ha been va ted for imact on esn anlye", and Its resoutin%reprstsf ak Non-Olsretonay hag.

a9sed on ret* fm a vk ro s 000stIv plant cablons, Itis monludedtat the e d ts err on the quate drecl*y impacted by the momeu equation calcut for spli channel (veocities, Now, eftc.) Is aNegIbls, leafin Wan aestiated peak caddin temerte hn~ac of 0F.

U.S. Nucle Regulatory Commission RA-14-0015 Attachment Page 5 of 11 ChanMge to Vilel S u -tissdStum 1-opeutle Affected Evalusion Model(s): 1998 Westinghouse Best Estimate Large Break LOCA Several related changes were made to the MCOBRA/TRAC coding for the vessel super-heated water p etihe, *lg updating the HOAS subroutine coding to be consisent with WCAP-12945.P-A Equation 1046, updatkothe approlm lodat of the enihalpy in the TIAS subroutine to be consistent wtthe HGAS subrotne coding, and upda" th temperau teration wmehd and convergence criteria in the TGAS subroutine. These changes represent a closely-related group of Non-Discretionary Changes.

The updates to te calculation of thesumerteMted seam poperties had generally lesshan 1OF mpat on the resulting arn temperature values, leading to an estmated peak cladding tperature m t of OF.

U t U t D sMy toMe s"09m, Tma Affected Evaluation Model(s): 1996 Westinghouse Best Estimate Large Break LOCA it was ideriled that for one-dimensional components In which heo buer to n steel 304 or 316 is modeled, he reference tempersure forthe metal denslt cstdtlon was allowed to vary; as a repsl* th total metal mass was not lserved. Correction this problem represnt a Non.Discrelonay Change. I This change primarily impacts the resior coolant system, loop piping modeled in the lMWg break los-of-oolant accident WOOBARAC models. Itwas Judged ta theedff of this dhange on the peak cladding temperature reesut was negligibe, Wlea 10 an estimated peak cladding tmperature impact of 0PF.

Deesy Nu odl Enw CeweUone Affected Evsauaion Model(s): 1996 Westinghouse Best Estimft Large Break LOCA The decay heat moe inthe WCO NWTRAC code was upda to cornrt the erroneously coded value of the yield fraction deolly from fission for Group 190of Pu- and to Include the term for unceriainty Inthe prompt enerw per fission inthe calculatio of the decay hea powr uncertain. Correction of these errors repreents a closely-rlated group of Non-Dbsceny Changes.

These changes hvem a negligile impact-onfte calculated decay heat poe, leadinig to en estimted peak claddingi temperature Impac of 0OF.

U.S. Nuclea Regulatory Commission RA-14-0015 Mlachnur*

Pae 6 of 11 Corselon to osVa mia Itxl Psesnm orfo Uuror Affected Evauation Models): 1998 Wst*nhoe Bests ima*Lag Break LOCA An error was diommwe inWCOBNRAMC whrebyt*e* oriopreem Od at the sot break TEE owncio to the BREAK cmVoe* was inoe cuMl*t*un the TER, hydraulic dWi ,t instead of the BREAK conent O ikput. This er*or h been valed for impact on xis anmlyse,-ad f resWut un repre t* o-icretioay Chang.

Based on the results from a matrix of repreew I lat doit isconcludkd thtO tW ef of dthis ero on the prsure at the bmk mid the brea flow is fogNgW, eadfng to an estiated peak claddin temlperatureqx pct o OO'F, ONd Nt TrnW*r Enhmoeemmat eulation Affected Evaluman Mod*s): 1936 Wesingmuse Best Esthmole Large Break LOCA An ise was idente which ould aec the ada of the hat trans at grkdded ertonm kr Best-Estiae Larle-Break Loss-4Coolml Accident Evauation Models. For a specific kvt* coanrxit*n the grid heat lia*feonw'itrcement fctor iscalmculed based on an ermro"n or gem*". which can ca M over-diction of "hhet transwe Coeffet at g e elWat*n. This iu ha ben evaloatmd to oesiat t *exis*ti kim on LMBOCA analysis ,mru. The ro*Mko of this "a reproeft a Non-4Ocrt.onay Cham.

The efectdeerdebva wajuged o hoaveaI~globl.efcin existig LSLCA anaysi muls, Wanto lan oe*mdppa codds teonmperae kmpac of 0F.

Clump to a"w SiwuP now amu Pei" Affected Evalation Mode(s): lM Weftn oue St O simam Large Bra LOCA AchaWin the mhodoll k og me toocalcu io ad e f Wesinhlouse ful esLfted Ina hange to Mhe grid two*for knrg brea low-skooclwt accfdt (LBLOCA) wan ri inpib affect hut trsm e inthe cr do n a LLOCA.

This ca reprMOs a Non-irmenry Chag.

The updates to lh ehodoolog to calclat grid blockg rati and porosity used as inpu in

• mode LBL.OCA Westinghous resulted ina

• hn* to het onsfer inte cwe for the fuelWt"e tsed in McGuMr Unis* I & 2 an CMwe Units I &2. The e stimted penalty associated with the dcange Is 0IF for both Ruflood I sad Refood 2 for 10 CFR 50.46 reporting purpose.

U.S. NM er Roauao Comm o RA-14,0015 Page 7of 11 OS.O*A Ckdng tran for Fut Rod wtM kw Affected Evaluation Modelks) 19865 West~nouse Sina Breask LCOCA, NOTRUIMP,_,

An error was discovered Intwo f*vwrn lc stran requredor burst for ZIRLOO claddkV in the SBLOCTA cods. The coding does not enforce reoching the minimm percwet local strain tshold prior to mlcu fu rod burs. Howevw, a review do ice n bslsan seýs revaled no intne of this error c An ed resub. ftsestlon of 9s Isu represenl its a Non-Discretloay Change.

Based on a review of curnt licensing basis analyses, and the phenoea and physlcs of a smaMl brak LOCA trnn, Itis co xdta thise r has a r g SeNa break LOCA ana s reVts lea*di to an uet d peak cadng tmetum of 0P.

U.S. Nuclar Regulatory Commission RA-14-0015 Attachment Page 8Of 11 Tabe 2 LOCAf tak Ote Te-w-w..u. 0.Smm . N o Unb I &2 Ev,@£ q model: QO1 9 CODION.... _ _ _ ___,

Mc~uire ULkt# I &

Analysis of record PCT (Ref~o 2) 2028 2/Catewba Units I

&2Composle Prior errors (APCT)

1. Decay heat in Monte Caro cculalons 8 Rolero"l M

2. MONTECF power uncerty correcon Reference B

3. Safety Ietion temperaum rane so Reereoe C

4. Iprut mo remulting in an inowoplete sdk*tn matr46s 25 RDfernc )

5. Rlevised Slowdown Heetup Uncertainty DWlsb*Ion 5 RAefernc E

6. Ves Unheat Conductor Nodng 0 RF

7. Thermal Deg i w a IS Refereh H,I Factor Bumdown

8. Revised HNet Transfer Multiplier 85 Reference K

9. HOTSPOT Clawd Burs Strain PM. 70 Re.rence L Prior evaluastion ode" changs (APCI

1. Revised AlgorOm for Average Fuel Twrpwatwe 0 Reference, F

2. PAD 3.4 to PAD 4.0 -75 RefereameH,1

3. Peak FQ = 2.7in boftomlhrd d core 0 Resere iesH,

4. MI Uprtep to Ol.7% d 3411 .Wt 1 Referenoes H, I Now errors (APCT)

1. None Now evaluatin model change (APCT)

1. None Final PCT L200 Evaluation model: ______

Anys o re PCT Ford 1323 2.........

_________________________Reference (3 Prior errors (APCr)

I1. Evaluation of Fuel Pelle Thermal Condu*tAy 0 Refrence J Prior *vakt on-model canges, (APT)

1. None 0 New errom (APCT)

i. None New evaluation model change (APCT)

1. None Abeolute V u Of ... na......for 0.t.s . .

Net chng in PCT for fhsreot 0_____

IFinal POT 1323_ _ _

U.S. Nuclear Regulatory Commission RA-14-0015 Attachment Page 9 of 11 Tallb3 LOCA Peak Cladding Tunpertur t Mrgn S&mmary - Cetawbe Unit I LIBLOCA Cladding To" Comnentm Evaluat nofx deW  : CO:' ,,! ....... :. ..... ..

FMcGuire Units 1 Analysis of record PCT (Reflood 2) 2028 & 2/Catawba Units 1 &2 Prior errors (APOT)

1. Decay heat in Monte CaIo calcukltions 8 Reference A

2. MONTECF power ucertaty correction 20 Reference N

3. Safety Injection temperature range 59 Reference C

4. Input error resulting in an incomplete solution maftix 25 Reference D

5. Revised Blowdown Heatup ntaknt DIs~tiuon 5 Reference E

6. Vessel Unheated Conductor Noding 0 Reference F

7. Thermal C*onduc Degradation with Peaking 15 References H, I Factor Bumdown

8. Revised Heat Transfer Muftie DMitr on -85 Reference K

9. HOTSPOT Clad Burst Strain Error 70 Reference L Prior evaluation model dnges (APCT)

1. Revised Agollhm for Average Fuel Temperature 0 Reference F

2. PAD 3.4 to PAD 4.0 -75 References H, i

3. Peak FQO 2.7 in bottm thrd of core 0 References H, I New errors (APCT)

1. None New evaluation model cun. (APCT)

1. None _ _ _

Final PCT 91o"2070___ __ _

Evauation model: NOTRUFMP Analysis of record POT 1323 2dWebrWeak, Prior erors (APCT)

1. Evaluation of Fuel Pelet Thermwal Conductvity 0 Reference J Prior evauation model change (APCT)

I. None 0 New eo (AMT)

1. None New evaluation model cwh e (APOT)

1. None Absolute VOW~ of eroscagsfor thi reort(PT 0 Final PCT)'*

N*t char in PCT for this re.ort.0 Ig0 .... i::

. .. Iii

U.S. Nuclea R"ultOry Comnmion RA-14-0015 Aftacment Page 1oof 11 WOCA Palk CMldin Temra Ma S uy- Ctawbe UnM 2 ILKOCA claddfn T"wi 000, Evalution modW: WLCOBRfTRAC, COD 1996 Analysis of record PCT (Reflood 2) 208&2cawtawb Unis 1&2 Prior error (APCT)

1. Dec heat in Monte Carlo caloulsin 8 Reteence A

2. MONTECF power unrtainty correction 9Rofrence N

3. Safety n t mper e rage 50 RieferNc

4. Input ewr resuM ingan incomplele sokution ntrx 25 Rqfern* D

5. Revised Sokwdoen Heau** Uncetainty Dltbbstlon 5 frec E

a. Vessel Unheaed Conductr Nodlng Rfernce i F

7. Ttermm Cwonduciy Dgdainwlth PeaMn 5 drncsH Factor Bumdown

8. Revsed Heat TransferM,*Dip D"tstbut-on rller -5 OeWrnoe K

.. HOTSPOT Clad sumt Strn ErMr *.. 70 Refereoe L Prior avaluation model change (6PCT)

1. Revised Algorithm for Average Fuel Temperature 0 Reference F

2. PAD 3.4 to PAD 4.0 -15 s ifI

3. PealtFO =2.7 in bcom thldocore 0RokwernO H,li New awe (APCT)

1. None ___

New evalualtion model chage (APCT)

I. None_ _

Final PCT 2070,. ______

gik, n model U.. .. ..

Analysis d record PCT

________________________ 123

________Reference 4 inch brek,0 Priorewmrs (APCT)

1. Evaluation of Fuel P.1s Themal COni~uoMly 0 RetdrnmS Prio evakiation model change (A#"I

i. None 0 New errs (APT)

I. Noe _ __

NOW INVaILuOn model dwgm (APC1 Absolut valu of Orroufthie IoA ~ report (16=T) 0_____

NeftMcag inPCT for IVe gn eot0 ____

Final PCT I23BM,_____

References for Tables 2,3, and 4:

U.S. Nuclear Regulatory Commission RA-14-0015 Attachment Page 11 of 11 A) Letter, G. R. Peterson (Duke Energy) to USNRC, Report Pursuantto 10 CFR 50.46, hwaes to or Errorin an ECCS Eva/uaton Mode/, April 11,2001 (ADAMS ML011070266)

B) Letter, M. S. Tuckman (Duke Energy) to USNRC, Report PAnuant to 10 CFR 50.46, Changes to or ErrorsIn an ECCS Evaluation Mode/, April 3, 2002 (ADAMS ML021070672)

C) Letter, W.R. MoCohn, Jr. (Duke Energy) to USNRC, RPepod fPruatto 10 CFR 50.46, Changes to or Erors n an ECCS Evauaton ModW, J* 29, 2003 (ADAMS ML03217063)

D) Letter, W. R. McColum, Jr. (Duke Enegy) to USNRC, Report Prmantto 10 CFR 50.46, Changes to or Eomr inan ECCS Evalan Model, May 26, 2004 (ADAMS ML041560349)

E) Letter, J. R. Morris (Duke Enemy) to USNKC, Report Pursuatto 10 CFR 50.46, Change to or Errors inan EC=S EWuaon Model, June 21, 2005 (ADAMS ML051790210)

F) Letter, T. C. Geer (Duke Eneg to USNAC, .eport Pursuantto 10 CFR 50.46, Chwa* to or Ermor an ECCS Evalaion Aki March 13,2007 (ADAMS ML07060058)

Gi) Lette, T. C.Geer (Duke Energ) to USNAC, Repor Pursuant to 10 CFR 50.46, Changes to or Errors Inan ECCS Evak I ModAL, May 22, 2007 (ADAMS ML071500297)

H) Letter, D.C. Cu. (Duke Enery) to USNRC, Reqone to formaon Reque Purnuamn to 10 CFR 5.54(0 Relaed to the Esaad ENdt on PeakClalding TeaperaunResultf from Trmal t Degr h the Wetnghous-Fu e Raf*in Emereny Con COiN SW"m Ewaai and 0-Day RWo Prua*t to 10 CFR 50.46, OChn to orErrors inan Eva&tion I.% March 16, 2012 (ADAMS ML1 207*I90)

I) Leter, J. Thompson (USNRC) to K. Henderson and S. D. Ceppe (Duke Enerwgy), Catawba Nuobaa Stton ULb 1and 2, and AoGut M Stati Units 1and2, Closur ar Evaluato for Repoftsuwt to TO 10 of Ow Cod of F, dw ROtoNs, Ft 50, Seclion 50.46, Parspwh (,a4X3) Conce g Si ft Enw cWy Con, Cooing Symm Evaluati AMl ErnorRlated to Mwx*e Fuel Themial Conr *Av4L Degradaton(TAC Nos. WS44", AER446, ME6449, and PEBS), November 16, 2012 (ADAMS ML12314A031)

J) Letter, M. J. kinaoone, (Duke Energy) to USNAC, Repot Pummafto 10 CFR 50.46, Changes to or Errors inan Evaltw Asxl, July 11,2013 (ADAMS MU 3199A29)

K) Letter, R. J. Duncan, (Duke Energy) to USNRC, Report PuwW to 10 CFR 50.4, Changes to or ErrorsInan E= EVation Afode, August 29,2013 (ADAMS ML13246AI03)

L) Letter, B. C. Wakirep (Duke Energy) to USNRC, Repot Pusmut to 10 CFR 50.46, Changisto or ErrorsIn an EvAalton Mode, Fbruary 25, 2014 (ADAMS ML 4063A278)

M) Lifter, M. S. Tudmn (Duke Energy) to USNAC, Report PuruWn to 10 CFR 50.46, Chva s to or Emar Inan ECCS Evaiafn ModW, May 3,2001 (ADAMS MIOI 1310891)

N) LOeter, M. S. Tuckmnn (Duke Energy) to USNRC, Report P:aauWa to 10 CFR .40, Chanp- s to or Erm inan ECCS Evakuaon ModW. Apr#3, 2002 (ADAMS ML02107068)