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{{Adams
#REDIRECT [[L-16-229, Submittal of Pressure and Temperature Limits Report, Revision 3]]
| number = ML16210A393
| issue date = 07/28/2016
| title = Submittal of Pressure and Temperature Limits Report, Revision 3
| author name = Boles B D
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000346
| license number = NPF-003
| contact person =
| case reference number = L-16-229
| document report number = 32 EFPY PTLR, Rev. 3
| document type = Letter, Report, Technical
| page count = 12
}}
 
=Text=
{{#Wiki_filter:FirstEnergy Nuclear @emting Company55Al North State Route 2 Oak Harbor; Ohio 43449Brkn D. tu'rre Vice President, Nuclear 419-32'l-7676 Fax: 41*32t-7582 Jufy 28, 2016 L-16-229 ATTN: Document Control DeskU.S. Nuclear Regulatory Commission Washington, DC 20555-0001
 
==SUBJECT:==
Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 Submittal of Pressure and Temperature Limits Report. Revision 3In accordance with Technical Specification 5.6.4, "Reactor Coolant System (RCS)Pressure and Temperalure Limits Report (PTLR)," FirstEnergy Nuclear OperatingCompany hereby submits Revision 3 of the PTLR for the Davis-Besse Nuclear PowerStation, Unit No. 1 (DBNPS), which was approved on June 24,2A16.Revision 3 removed the restriction of exceeding the operating limit date ofAprif 22, 2017, which was included in earlier revisions. This change is the result of the Nuclear Regulatory Commission (NRC) issuing a renewed operating license to DBNPS extending the period of operation to midnight April 22,2A37 , which made 32 effective full power years limiting. This change also corrects an administrative error that existed in the previous revision where Figure 2 was combined with Figure 1 and the original Figure 3 was designated as Figure 2; however Figure 3 was still referred to in the bodyof the document.There are no regulatory commitments contained in this submittal.
lf there arc any questions or if additional information is required, please contact Mr. Thomas A.
Lentz, Manager - Fleet Licensing, at (330) 315-6810.Sincerely, tultu Brian D. Boles Davis-Besse Nuclear Power Station, Unit No. 1 L-1 6-229 Page 2
 
==Enclosure:==
 
Pressure and Temperature Limits Report for Up to 32 Effective Fulf Power Years, Revision 3 cc: NRC Region ll I Administrator NRC Resident lnspector NRC Project Manager Utility Radiological Safety Board EnclosureL-1 6-229Pressure and Ternperature Limits Report for Up to32 Effective Fufl Power Years,Revision 3 (Nine Pages Follow)
FTRSTENERGY NUCLEAR OPERATINC COMPAhIYDAVIS-BESSH U}{IT IPRESSURE AND TEMPERATURE LIMITS REPOR'['FORUP TO 3?EFFECTIVE FULL POWER YEARS Revision 3 Prepared by: Reviewed by: 6al, At/x fwit David W. Gerren 32 EFPY PTLR Rev. 3Page 2 of I FirstEnergy Nuclear Operating CompanyDavis-Besse Unit I Pressure and Temperature Limits Report for up to 32 Effective Full Power Yearsl,0 lntroduction This Pressure and Temperature Limits Report (PTLR) provides the information requiredby Davis-Besse Nuclear Power Station (DBNPS) Technical Specification 5.5.4 to ensure that the Reastor Coolant System (RCS) prcssure bormdary is operated in accordance withits design. The limits provided are valid to 32 Effective Full Power Years (EFPY).The PTLR provides the RCS Oporating Limits in Section 2.0, which satisfies Technical Specification 5,6.4.a. The furalytical Methods used to develop the limits, including determination of the vessel neutron fluence, are provided in Section 3.0, ftlfilling Technical Specification 5,6.4.b. The information and formafting of Section 3 follows the guidance ofAfiachment I to Generic Letter 96-03.
The PTLRrequirements areprovided in Section 4.0 of &c report, fulfi lling Technical Specification 5.6.4.c.
Revision 0 was the initial issue of the 32 EFPY PTLR after issuance of License Amendment 282, which authorized use of new methodologies.
Rcvision I is re-issuingthe32 EFPY Pressure-Temperatr.ne limits to include the limits for the Reactor Vessel Closure Head (RVCH) installed in October 2011 Cycle 17 Mid-cycle Outage. The limits associated with the RVCH obtained frorn the Midland nuclearpower plant have been removed. No methodology changes occurred in this rpvision Revision 2 is re-issuing the 32 EFPY Pressure-Temperatre limits to incorporate Revision 4 ofANP-27l8, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station" (Reference 5.7).Revision 4 of AlrlP-2718 combined the Heatup/Cooldown Curves into a single Figne.This results in iire re-numbering of the ln-Service Lealc md Hydrostatic Tests Figure toFigure 2. No methodology changes occurred in this revision.Revision 3 removes the restriction of exceeding the operating limit date of Aptil 22,2Anwhich was included in earlier revisions. This change is the rezult of the Nuclear Regulatory Commission issuing a renewed o;rerating license to Davis-Besse which extends the period of operation to midnight AWi122,2037,which made 32 EFPY limiting. This change also conects an administradve error that existed in the previous revision where Figure 2 was combined with Figure 1 and the original Figure 3 was designated as Figure 2, however Figure 3 was still referred to in the body of the document.Revisions to the PTLR are to be submitted to the NRC aft*r issuance.
2,032 EFPY PTLR Rev.3Page 3 of 9 RCS hessure and Temperature Limits a. The Rsactor Coolant Syslem (exccpt thc pressurizer) temperfiire and pressure shall be limited in accordance with the limit lines and rarnp rates shonm on Figures I and2 (Reference 5.7) during heatup, cooldown, criticality, and in-service leak and hydrostatic (ISLH) testing with:l. A manimum heatup of 50oF in any one hourperiod, and
: 2. A maximum cooldown of 100oF in any one horr period with a cold leg temperature of > 270oF and a maximum cooldown of 50"F inany one hour period with a cold leg temperature of
< 270oF.b. During periods of low temperalure operation (Tag <280 oF), Technical Specification 3.4.12 (Reference 5.3) provides additional requirements for RCS pressure and temperature limits. Those limits are maintained in the Technical Specifications because they are not deternrined using methods generically approved by the NRC, 2600 2400 22A0 2000 1800 1600 14001 2001 000 800 600 400 200 03a EFPY#1.:
PeF4 dg Figure 1: Composite Normal HeatuilCooldown Limit - Both Hot Leg Pressure Taps'ilts l'leatuplCooldoum Lirnil-*r Criticality Limit ilotrr:l. Atilnbh lrrtup rrtc b 50 T/fr Grnpl, lffid by r 15 T rtcp drrrga fdlil,ld by m l&rnlnub ffi.2. A0ilabb cookbtn rrlr d srbop 270'F b f il T/nr tRemp). lmlcd by r 15'F dtp chrqn HIM W r $"mhut hold,3, Alornblc cooHorrn reb bcbt n0'f ir 50 'Flhr (Rrmp), Imffi by r 15'F @ chrtqe fullomd by rn f&mlnuD hofd, 1. A mlrlnum rhp tunprfrrturt chuee sl l3'F b rtonblo x,hdr tltrrottlt*
dl RC trttnpo hom oparrtlon rfh he DFlR rftrm opct$fq. Ttr thp hnryrntu't drrngc b .bfrrcd rs RC gnp r$xn lhr Dln ctumhn* b thc ltr&t coolant ry&tn Rrbr lo stoppltf et;PunFr.5, tffhon trt dccry horl nmovd 3yrt m (0H) 3 operdng$ffhotil rny FC puttPt opcr$U, fittllcabd il{ nrurrt lrmprdrrt b$. na6r vcllel rhellbr urill, 6. Ttla eetrrHt prtlrur! end lacperrtuic mmbhdlonr an bclcr and to lhc ftht dthr lmlt snt.7. lnrfturpnt alor b nd rmul$d br In tn* llffi.100 150 200 Tcmprnfurr, cF 250 i I a!I I.9 gt CL E 3 o tn E o.
32 EFPY PTLRRev. 3Page 5 of 9Figure 2 Reactor Coofant System Pressure-Temperature Heatup and Cooldown Lirnits for ln-Service Leak and Hydrostatic Tests P.qu! bgl grcrglesut la*p Hrps$h 70 s71 I C 1J0 t296 ._ ..t_.l-----Y-----
ts ale I D ili Mga 150 l5B3 I195 168I *- +2600 2400 22AA 20001 8001 600 1400 120a1 000 800 600 400 200 0 00 0Bg ss 909 e0 933 95 961 100 sg3 E 160 1793165 1859170 1924 tl.tt Altowabb heatup rate b 50 oFlhr (Ramp). fi'nited by e 15 'F slep change folh,vad by an 1&rnlnute hold.2. Albntrable cooldown rate at or aboue 27goF b 100'F&r tRamp), lln$ted by a l5'F step change follo$rsd by a 9+ninutc hold.3 Allowabh cooldom rate befff 270 oF is 50'F/hr (Rarp), limited by a 15'F slep change lolhwed by an 1&minuie hold.4. A rnexmum step temporaturt change of 15'F ig allol,able when rarnovttg all RC purnp from operation with ttre DffR s)*tern operatirry.
The rlep tcmperature changc is delined as RC t*rnp minrcthe DHR refum lcmprlo thc rpacior coolantsystern prior to stpppir* allpump*.5, When the decay heat prnoralsystem (DH) is oparating without any BC pumps opcrating, hdicEted Dfl return lernpergtura to the reedor tnssel shafl bo used.6, The amptabfe presourc and ternperntuts combindhnr lrc below snd to the right of the limit curw"?. Instrurnent enor b not accounted for ln th$e llmt$, 250 300 350 400 ,9!t o CL E 3 n o E o.105 102e l 175 1997 110 1S6* I F 180 2078 f15 1108 I 185 2170 50 100 0 150 200 Temperature, oF I 32 EFPY PTLR Rev. 3 Page 6 of I 3.0 Analytical Methods 3,2 3,3 3.4 3.1 Ihe limia provided in Section 2 and l'igures I and 2 are valid until the Reactor Vessel has accumulated 32 llffective Full Power Years (EFPY) of fa* (E> t MeV) neutron fluence.The neutron fluence is calculated (Reference 5.12 with Reference 5.13) consislent with Regulatory Guide 1.190 using the NRC-approved methodology described in BAW-2241P-A (Reference 5.5). Table 1 provides the neutronfluence values used in the adjusted reference temperaftre calculations.
The listed fluence values are based on 52 EFPY of opentiou.The Davis-Besse Reactor Vessel Material Surveillance Program complies withthe requirements of Appendix H to l0 CFR 50 and is described in BAW-I543A (Reference 5.6). This infomration was approved by the NRC in the SER of Amendment 199 (Reference 5.1). The specimencapsule withdrawal schedulc is contained within the supplements of the topical rcport.
All plant specificspcimen capsules have been withdrawn from the reactor vessel. The ART values were not calculated using surveillance data (Reference 5.14) since it was detennined to be non-credible.
[,ow Temperature Overpressure Protection (LTOP) limirs are addressed inSection 2.b, above, and Technical Specification 3.4.12 (Reference 5.3).Reference
 
===5.7 discusses===
 
the methods used to determine the temperature at which LTOP must be active. The pressure limit was detemnined using ASME SectionXI, Appendix G, as modified by the a:ternative rules provided in ASME CodeCase N-588 and ASME Code Case N-640 (Reference 5.9).Table I provides the Adjusted Reference Tempetahre (ART) for each reactor vessel beltline material. The ART values were calculated in accordance with Regulalory Guide 1.99, Revision 2. For welds in the reactor beltline region, the initial RTNor values used (in part) to deternine ART were calculated using an altemafe methodology describcd in ttre NRC-approved BAW-2308, Revisions l-A and 2-A (Reference 5.10). The NRC required licensees to obtain an exemption from I 0 CFR 50.61 and 10 CFR 50, Appendix G to use the altemate initial RTNor values provided in BAW-2308 Revisions l-A and 2-A. The required exemption was granted by the NRC in Reference 5.17. The NRC oonfirmed the limits and conditions for using the methodology were satisfied inthe SER of Amendment 282 (Reference 5.8).The Pressure-Temperature (P/T) limits of Section 2 and Figures 1, a:td 2 (wittl applicability as stated in 3.1) were generated consistcnt with the requirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99, Revision 2, using themethods described in BAW-10046A (Reference 5.4) and ASME SectionXl, Appendix G (Reference 5.9), as modified by the altemative nrles provided in ASME Code Case N-588 and ASME Code Case N-640.
3.5 3,6 3,7 3.832 EFPY PTLR Rev.3 Page 7 of 9 3.6.1 The NRC has reviewed the methods dessibed in BAW-10046A (Reference 5.4) and approved the topical report by issuance of a SafetyEvaluation Report (SER) dated April30, 1986.
Section 1.2 of BAW-10046A states that it is applicable to all cunent B&W nuclear steam systems.3.6.2 ASI{E Code Cases N-640 and N-588 have been incorpomted into ASME Section XI, Appendix G, 2003 Addenda" whish arethe edition and addenda codified in l0 CFR 50.55a (cffcctive May 27,2008) and thus maybe used perNRC Regulatory Issue Summary GIS) 2004'04. Specific approval for application at DBNPS is included inRef. 5.8.The minimum temperature requirements of l0 CFR 50, Appendix G are ineluded on Figures I and 2. Figure 2 provides the In-Service kak and Hydrostatic (ISLH) Test Limits. These limits were calculated in accordance with therequiremeirts of l0 CFR 50, Appendix G and ASME Code Section XI, Appendix G, 1995 Edition, with Addenda through 1996 and ASME Code Cases N'588 and N-640.Davis-Besse has rernoved more than two surveillance capsules. The capsule test results have been evaluated and found to be non-credible (Reference 5'14).Consequently, ART calculations are not based on the strnreillance data. TheMeasured ARTnor - Predicted ARTNoT dala scatter was less than 2o, so theRegulatory Guide 1.99, Rev. 2 Chemistry Table values used in the ARTcalculations are conservative.
4.0 PTLR Requirements 4.1 The PTLR has been prepared in accordance with the requirernents of TechnisalSpecification 5.6.4 (see Reference 5.1 l). The PTLR shall be provided to the NRCupon issuancs for each reactor vessel flue,nce period snd for any revision orsupplement thereto. I)avis-Besse will continue to meet the requirements of l0CFR 50, Appendix G, and any changes to the Davis-Besse P/T limits will be generated in accordance with the NRC approved methodologies described in TS 5.6.4.
32 EFPY PTLR Rev. 3 Page I of I Tabbl: Davis-Besse Nuclear Power Station Reactor Vessel Beltfine Region Data (Appficable as noted in Section 3.1)
I Reactor V6gel I tnsation ldaterial Idcntification Flucntx@ 52 EFPY (Wetted Surface)(n/cm2)(E> I McV)ART@%T{"F)@52 ErPY{Notc l}ART@'/tT fn@s2 EFPY (Notc I )limiting Mat'l?(Ye#No)RTrn ("n{Note 2)Nozde Bclt ForeineADB 203 2.:9F+18 74,8 64.E No I r.2Nozzle Belt to Upper Shell Weld (lD 99o)w-232 2.29E+18 Note 3 Note 3 No l5?.9Nozzle Belt to Upper Shcll Weld (OD 9l7o)wF-2332.29E* l8 100.4'67 ,E*No Note 4 Upper Shell Forgiur AKI233 1.69B+t9 7l.E 57 -3 No 79.4 Upper Sbgll to Lower Shell WeldwF-r E2- l I .69E+ l9 156.2r 1ffi.4't Yes 182.2ilnwer Shelt Fors,ine BCC 24I1 ,79f,+ l9E9.978.8 Yes 95.7 Notcl:ReponcdARTvrluesurobasedonRegularoryGuidel.gg,Revision2(Ref 5.15). P/TLimitcalcuhionwasbsscdonateinpcralurcvrluo which is more comervative tbefl tho llstcd ART ralue. (Rcf, 5. l 3)Note 2: vrluss fi,om Ref. 5.16, whic,h m brscd on tbe location spccific clad ro vessel interfacc fluonco at 52 EPPY.NoJe 3: This wcH rnatsrirl docs not oced ou ro rh* %T or %T location.Nco 4: this wcld macrirl is nol present at ate clad to vcsscl intorfrcc, so RTrrs does not aprply to it.
I Basod on thc iuitial RTxor Fovkled in the NRC Safety F,valuation Rcpons ro BAW-2308, Rev. l -A rnd 2-A (R.f. 5. l0).
32 EFPY PTLR Rev.3Page 9 of9
 
==5.0 References==
 
===5.1 Safety===
Evaluation by the NRC Officc of Nuclear Reactor Regulation Related to Amendment No. 199 to Facility Operating License No. NPF-3 Davis-BesseNuclear Power Station, Unit No. l, attached to correspondence dated July 20, t995.5.2 Technical Specification 5.6.4, "Reactor Coolant System ECS) PRESSURE Al.lD TEMPERATURE LIMITS REPORT (PTLR),"5.3 Technioai Specification3.4.12, "Lolv Temperature Overpressure Protection."5.4 BAW-10046A, Revision 2 "Methods of Compliance with Fracture Toughness andOperational Requirements of l0 CFR 50 Appendix G." 5.5 BAW-224IP-A, "Fluence and Uncertainty Methodologies," dated Aptil 1999.
5.6 BAW-1543A, "Master Integrated Reactor Vessel Marerial Sunreillsnce Program.'n 5.7 ANP-2718, Revision 4, "Appendix G Pressure-Temperatrne Limits for 52 EFPY,Using ASME Code Cases for Davis-Besse Nuglear Power Station"" datedDecember 2013.
 
===5.8 Safety===
Evaluation by the Office of Nuclear Reactor Regulation Related toAmendment No. 282 to Facility Operating License No.
NPF-3, FirstEnergyNuclear Operating Company Davis-Besse Nuclear Power Station, Unit No. I, (FENOC Lh. Rl l-030), dated}ll2&l20rl.
5.9 ASME Code Section XI, Appendix G, as modified by the alternate nrles providedin ASME Code Case N-640 and ASME Code Case N-588.
ASME Code CasesN-640 and N-588 have subsequently been incorporated into AStttE Sedion XI,Appendix G,2003 Addenda, which are the edition and addenda codified in 10CFR 50.55a (effective May 27,2008).5.10 BAW-2308, Revision l-A and Revision 2-A, *Initial RTNpr oflinde 80 Weld Materials," dated August 2005 (l-A) and March 2008 (2-A).5.11 Calculation C-NSA-064
.02-037, Revision l, "Davis-Besse 52 EFPY PT Limits -Chalon RV Closure Head," dated9l23/2011.
5.12 AREVA Repon 86-9015129-000, *DBl - Cycles 13-15 Fluence AnalysisReport,o' datrd 4 I 21 12006.5.13 AREVA Report 5l-9123331-000, "Davis-Besse - EOL Fluence Reconciliation o'dated 10/8/2009.
5.14 AREVA Document 32-90311 57-000, "Davis-Besse Revised ART Values at 52 EFPY," dated 9 120/2006.
5.15 AREVA Document 32-9017744-003, "Davis-Besse ART Values at 52 EFPY," darrdl0f29l2B9.5'16 AREVA Document 32'9123247-000, *RTrrs Values of Davis-Besse Unil I for 52EFPY, Including Extended Beltline," dated Ll I L2l 09, 5.17 NRC Lrtter to FirstEnergy Nuclear Operating Companyn "Davis-Besse Nuclear Power Station, Unit 1-Exemption from the Requirements of l0 CFR Part 50.61and 10 cFR Paft 50, Appendix G," (FENOC Ltr. R10-298) dated Deccmbcr 14, 2010.}}

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