L-22-216, Submittal of Pressure and Temperature Limits Report. Revision 4

Submittal of Pressure and Temperature Limits Report. Revision 4

ML22271A258
Person / Time
Site:	Davis Besse
Issue date:	09/27/2022
From:	Tony Brown Energy Harbor Nuclear Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-22-216
Download: ML22271A258 (14)

Text

.a energy

~

harbor Terry J. Brown Site Vice President, Davis-Besse Nuclear September 27, 2022 L-22-216 ATTN: Document Control Desk U.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 Energy Harbor Nuclear Corp.

Davis-Besse Nuclear Power Station 5501 N. State Route 2 Oak Harbor, Ohio 43449 419-321-7676 Submittal of Pressure and Temperature Limits Report. Revision 4 In accordance with Technical Specification 5.6.4, "Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR)," Energy Harbor Nuclear Corp. hereby submits Revision 4 of the PTLR for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS), which was approved on December 16, 2021, and effective on September 14, 2022.

Revision 4 revises the effective full power years (EFPY) from 32 EFPY to 43.5 EFPY.

There are no regulatory commitments contained in this submittal. If there are any questions, or if additional information is required, please contact Mr. Phil H. Lashley, Manager - Fleet Licensing, at (330) 696-7208.

Enclosure:

Pressure and Temperature Limits Report for Up to 43.5 Effective Full Power Years, Revision 4 cc:

NRG Region Ill Administrator NRG Resident Inspector NRR Project Manager Utility Radiological Safety Board

Enclosure Pressure and Temperature Limits Report for Up to 43.5 Effective Full Power Years, Revision 4 L-22-216 (12 pages follow)

ENERGY HARBOR NUCLEAR OPERATING COMP ANY DA VIS-BESSE UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT FOR UP TO 43.5 EFFECTIVE FULL POWER YEARS Revision 4 Prepared by: ~ ~

John R. Marko Reviewed by: 4,4-J~.2 'Y'h~

Michael L. Nelson Date:

12 / I /1. I Date: 11/1 /1\\

jlria_n A. Kanney Approved by: ~

~. ~

~-- Date: / 6.Q SC d-:: ~ l-1 BrianA~

1.0 Introduction EnergyHarbor Nuclear Operating Company Davis-Besse Unit 1 Pressure and Temperature Limits Report for up to 43.5 Effective Full Power Years 43.5 EFPY PTLR Rev.4 Page 2 of 12 This Pressure and Temperature Limits Report (PTLR) provides the information required by Davis-Besse Nuclear Power Station (DBNPS) Technical Specification 5.6.4 to ensure that the Reactor Coolant System (RCS) pressure boundary is operated in accordance with its design. The limits provided are valid to 43.5 Effective Full Power Years (EFPY).

The PTLR provides the RCS Operating Limits in Section 2.0, which satisfies Technical Specification 5.6.4.a. The Analytical Methods used to develop the limits, including dete1mination of the vessel neutron fluence, are provided in Section 3.0, fulfilhng Technical Specification 5.6.4.b. The information and formatting of Section 3 follows the guid,mce of Attachment 1 to Generic Letter 96-03. The PTLR requirements are provided in Section 4.0 of the report, fulfilling Technical Specification 5.6.4.c.

Revision O was the initial issue of the 32 EFPY PTLR after issuance of License Amendment 282, which authorized use of new methodologies.

Revision 1 is re-issuing the 32 EFPY Pressure-Temperature 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 from the Midland nuclear power plant have been removed. No methodology changes occurred in this revision.

Revision 2 is re-issuing the 32 EFPY Pressure-Temperature limits to incorporate Revision 4 of ANP-2718, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station" (Reference 5.7).

Revision 4 of ANP-2718 combined the Heatup/Cooldown Curves into a single Figure.

This results in the re-numbering of the In-Service Leak and Hydrostatic Tests Figure to Figure 2. No methodology changes occurred in this revision.

Revision 3 removes the restriction of exceeding the operating limit date of April 22, 2017 which was included in earlier revisions. This change is the result of the Nuclear Regulatory Commission issuing a renewed operating license to Davis-Besse which extends the period of operation to midnight April 22, 2037, which made 32 EFPY limiting. This change also con-ects 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 body of the document.

Revision 4 is the initial issue of the 43.5 EFPY PTLR. CR-2019-03982, Reactor Vessel 52 EFPY Projected Neutron Fluence (Exposure) Higher Than Previous Projections (Reference 5.17), identified that due to the reactor vessel (RV) fast neutron tluencc accumulating faster than previously predicted, the RV would reach the analyzed tluence

43.5 EFPY PTLR Rev.4 Page 3 of 12 sometime prior to 52 Effective Full Power Years (EFPY) of operation. Document 32-9300671-000, Davis-Besse Reactor Vessel Embrittlement Fluence Reconciliation Through 60 Years (Reference 5.16) and CR 2020-07840 "Revised Vessel P-T Curves Limited to 43 EFPY" (Reference 5.18), has determined that the most limiting part of the RV will reach the analyzed fluence at approximately 43.5 EFPY. Data is based on fluence projections of Fuel Cycles 18 and 19. No methodology changes occurred in this rev1s10n.

Revisions to the PTLR are to be submitted to the NRC after issuance.

2.0 RCS Pressure and Temperature Limits

a.

The Reactor Coolant System ( except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines and ramp rates shown on Figures 1, 2, 3 and 4 (Reference 5.7) during heatup, cooldown, criticality, and in-service leak and hydrostatic (ISLH) testing with:

1.

A maximum heatup of 50°F in any one hour period, and

2.

A maximum cooldown of 100°F in any one hour period with a cold leg temperature of~ 270°F and a maximum cooldown of 50°F in any one hour period with a cold leg temperature of < 270°F.

b.

During periods oflow temperature operation (Tavg <280 °F), 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 determined using methods generically approved by the NRC.

43.5 EFPY PTLR Rev.4 Page 4 of 12 Figure 1: Reactor Coolant System Pressure-Temperature Heatup and Criticality Limits 3000

~iolimit 2800 i*n'(ol)

~ T,em,;,

70 S19 190 75 Sl9 lllS 2600

$,)

S19 200 85 519 205 2400

,o 519 210

~$

Sl9

~lS 2200

.too Sl9 220 105 519 225 110 519 226 2000 lH 519 226 120 S:19 230 llO 1800 U :',

519 235 L'IO 519 240

~ 1600 13~

519 245

~

140 519 250

<ll 1400

~

ct 1200 145 Sl 9 2SS 150 319 2&>

lSO 704 265 lSS i.M~

270 160 178 275 1000 163 820 280

!70 86t 2!!5 800 1?5 917 290

ISO 973 29S 185 1ag5 300 600 400 200 0

0 50

~

UC*:!

11i7 l2SS 134!1

!448 15$7 u;n 1810 2810 1810 l>l3/4 2116 2293 2487 27!'.i2 2938 1950 2950 2960 Z%0 2960 2960 2960 2950 1%0 100 Crih;ally i,-r*

1erm~

21,S 226 230 2:lS 24-0 245 2>0 255 260 265 270 27~

s.o 285 290 295 300 150 r&

0 104$

1102 1* 7 1258 349 1448 1557 1677 18 0 1956 2116 2293 2487 2702 200 Temperature, "F l Nlawoble~~up1.Me I~ 51} 'f,'}u ~Ramo~ ~d b-t 15 *r s,epdia"'i~

~IQl,t,"ed by ~ 11,miiur. flald

/JlfM;abll-cookbm011iil4Drabo:mi27(1 'rls. ta<)*f,.,,, tR.:impJ_WllildJry

"'1~ *F f1a~(h~IIQll loi'IOiN~dl)~,- '(J.~ut,rlbi,JO

> Alow;ati.cockk,i,mat!!bftCrw:}.10'Fis-50"F*1'rlAMr1pl limiledh!':I 1!i

~iu,-p chang.f~w,itbyan 1~1""""~

"' A l'll,l,ljmufP s-lt-P rt'fl1Pf1,11ure cNng,p or 15 *f b ancw*~e *""" tt~

~Rt P4Jl'l'P1 l\\'omop,M.-011 w111 !'Ml OHR ~Y~11!!ffl-6pt,r~ lh" ~*P.

.. mpa,atvr.. d'l*ngwt is-d..,,.,.d M RC I.mi> mi,ius lht OOA r<<lurn ~~

ta ifht,e~r-t,OOIRM \\',flffl flf!t,\\I' to !,1GH)tlQ,JI pUmps

~ 'A'h~ lhe d11<:;ry.,nl r~nJO"f'".-1 ~*,stwm (OH)i,. opttr:,ling *ithat1t.any RC 1£JUl'l""lfl'- 00tl1'fif,Q in$c:#ltd'U1 ttluin tft'ilpt-1,AJvO ~ tN-fo@o&C:W ¥1!1'~

ffltJlbeus-ed

~ f"" a,caPilb't (>f~'>wr,e arwt lffll~.:iltlff,(l~.in@M.ire~ ~nlttft 1l!r,qhtaflf'l*liml1cuo-!.

lnS'INnw:.il ffiOl ilo Ml ikCOUIINd ior WI ttlH* trntt.

250 300 350 400 Heatup Limit Criticality limit

3000 2800 2600 2400 2200 2000

-~ 1800 0..

~ 1600

I

"' 1400 (11 0: 1200 1000 800 600 400 200 0

0 43.5 EFPY PTLR Rev. 4 Page 5 of 12 Figure 2: Reactor Coolant System Pressure-Temperature Cooldown Limits

~:e,o

,,.u. l"".. t.*-m.,t

!=-

-~

ll::Oll2-Si.9 41-8 1102 S19 4&i 1177

!,!'P 478 125S.

5-1"'

Sl9 1349 S.19 5 19 14'18 539 519 1557

019 519 1577 S-19 519 1310

S-l':'-

$1~

1810

.19

=-1-s-1 UO

$1'!'

sa 1956

.$.l9 519 2116 51!1 SH 2293 519 519 2487 519 519 2702 S19 S~9

.1938 Sl9 519 29~

70~

1054 29W 7J!l

~05':0 2%0 778 1184 2:960 s:io

'1277 2950 866 1574 2960*

'!U7 1494 2950 973 1620 2.';160

1035

!752 2950 50

?toi~,S 24311 24:33 2438 2438 2;133 2438 2438 2438 2438 2438 243S 2li38 2438 2<>3-8 2451 24SS

!455 2457 2.SS~

2~~1 2463 100 Notes:

I. Alla..:able healup role is SD 'F/br /Ramp). lim~ed by a 15 'F step chanye lollowad by an 1s.,,,inv1e hold

2. Allowable cooldown rate al or abose 271! 'Fi* 100 'flhr(Ramp}.llmlted!>y

, 15 'F ~ttp*ehange fo*ow..t by a 9,minut-1.' hok!.

,. Allowablecoold~n rale below270 'Fi& SO 'Flhr (Rampi lim;ted by a 15 "F s1ep chonge followed by an 18-minute hold J A ma>rimum ~tep tempe<amredtange or 15 'Fis atlQwablewhen temoving

• II RC pump,, from DP4'rallon v.ilh !he DHR *.vstern operating. The step tampera1ure change i* deimed u RC ll!mp minus the OHR re1um lemps to the reactor coolan1 sy;tem prior to slopping,o pumps

5. WIIM !ht daca~ hMt il>l>>O<al ~~!QM (OH) i!; opora!iftg wilhoul' any RC pump,, oper~1'11g_ ind",c~1ed DH return 1emper,1ure to tile ru~1or ~=-4 shall be used

6. The acceptable preuure-and tem~rafU're combioations are beJow and *o 1h41' 1igh1 ol th* ~mit CU1'¥8 lnstrum.en, -errot is not a.ccount~d for ;n 1hase limils.

150 200 250 300 350 Temperature, °F

-e-Cooldown Limit

43.5 EFPY PTLR Rev. 4 Page 6 of 12 Figure 3: Reactor Coolant System Pressure-Temperature Heatup Limits for In-Service Leak and Hydrostatic Tests 4000 3800

~ 1'-f':?

Pr=

~

70 519 180 3600 n

S,

~

18$

80 s1, 190 3400 s,

51~

195 3200 5,0 519

z 95 519

~Q3 3000 100 519 1:10 105 519 215 2800 110 sa 220 2600 115 519 z~

l20 519 21'0 o,, 2400

125, 519 235

n. 2200 a,*

2000 J..5(1 519 2~

135 519 245 ldO 519 2~

145 S19 23\\S

"' 1800

~

a. 1600 l50 519 250 150 976 2-S-S.

lSS.

1023 i1 1400 u,o lll1S 275 1200 15~

113-1 280

!70 11!13 28S 1000 17,;

1261

~!lO IS 133S 29:,

800 175 U61 WO 600 400 200 0

0.00 50.00

~

l335 1417 1507 l&lo 1716 183~

19(>&

211 2274 Z::l5'.!

2645 28S~

w;;s 3334 3640 39:,4 3991 3991 o!>!il 3991 3991 3391 3.991

~!',91

)991 100.00 150.00 200.00 Temperature, °F Notes*

t. i\\llawablo heatup rate Is SO 'Mu (Ramp). limtted by a 15 *F step <han9e followed by an lB*minute-holid

!. Allowable <Mld<>wn ral* al or abbv* 270 'F It 100 'Flhr (Ramp). r.mhed b~

a 15 "'F slep cbange-follow9dby a 9*minute t,old

.l Allowable <Mld<>Wn rale below 270 'F ;s 50 'Flhr (Ramp). Tim~ed by a 15

'F $lop change followed by an 18-mlnute hold.

.J~ A maxirnum s1ep len,pereture cllenge of 15 ~ is aMowab(e when removing an F\\C pumps Imm operation with the OHR5y>tem operating. The step t@mpara1ure change o, defined** RC lemp minus lhe OHR rolurn temps lo tho re*ctor coolBN system poor to stepping. ell pumps.

5. Whon Iha dee.ay ~**t romoval *y&1*m (OH) Is oparating willlout any RC pump!. op@rating, indicaled OH r@tum temp@rature to lhe reactor ve$~el 11hall bo u~ad.

6 Tht ace@plable ptessrJre and temperature-comblnations a,e below and to

,h~ tight or the 1:ma w*"*

-. lr,strvm&tit error is nQt accqunted for in lhese Nmrts 250.00 300.00 350.00 100.00

43.5 EFPY PTLR Rev. 4 Page 7 of 12 Figure 4: Reactor Coolant System Pressure-Temperature Cooldown Limits for In-Service Leak and Hydrostatic Tests 3400 I!!£

~

70 51 3200 7S 51 3000 SU

!5 519 2800 90 S19 95 519 2600 100 51 105 Sl9 2400 110 51 115 519 2200 120

~19

-~ 2000 m

519 BO 519 Q.

ai' 1800 135 51 140 519

i 1600 Ill V>

143 519 148 S19

~ 1400

0.

ISO 51~

150 1.£43 1200 153 14 158 161 1000 163 1741 163 1869 800 173 203 600 l7S 2198 400 200 0

0 50

!a. fa.

183 2S7<

188 ma 193 2$14 19!

3056 203 3288 208 3283 2B 32S&

218 3288 m

32S&

228 328S m

3288 238

zas 243

"$288 248 528!

1.S3 3288 258 3288 263 S288 270 3~:1 27S sm U!O S316 m

3318 290 Sl21 295 3324 300 3327 100 150 Notes:

I. Allowable heatup rate is 50 'F/hr (Ramp). limited by a 15 'F step change followed by an 18-minute hold.

2. Allowable cooldown rate at or above 270 'Fis 100 'F/hr (Ramp). limited by a 15 "F step ehange followed by a 9*minute hold.

. Allowable cooldown rate below 270 'Fis 50 'Flhr (Ramp). limited by a 15

'F step change followed by an 18-minute hold.

~. A maximum step lemperature change of 15 'f is allowable when removing all RC pumps from operation with lhe OHR system operating. The step temperature change is defined as RC temp minus the OHR relum temps to the reactor coolant system prior to stopping all pumps.

~- When the decay heat removal system (DH) is operaling wilhoul any RC pumps operattng. indicated DH return temperature to the reaclor vessel shall be used.

6. The acceptable pressure and temperature combinations are below and lo the right of the limit curve.

". Instrument error is not accounted for in these limits 200 Temperature, "F 250 300 350

-+- ISLH Limit 400

3.0 Analytical Methods 43.5 EFPY PTLR Rev. 4 Page 8 of 12 3.1 The limits provided in Section 2 and Figures l, 2, 3, and 4 are valid until the Reactor Vessel has accumulated 43.5 Effective Full Power Years (EFPY) of fast (E> 1 MeV) neutron fluence (Reference 5.16).

3.2 The neutron fluence is calculated (Reference 5.7) consistent with Regulatory Guide 1.190 using the NRC-approved methodology described in BA W-2241 P-A (Reference 5.5). Table l provides the neutron fluence values used in the adjusted reference temperature calculations. The listed fluence values are based on 52 EFPY of operation.

3.3 The Davis-Besse Reactor Vessel Material Surveillance Program complies with the requirements of Appendix H to 10 CFR 50 and is desc1ibed in BA W-l 543A (Reference 5.6). This information was approved by the NRC in the SER of Amendment 199 (Reference 5.1 ). The specimen capsule withdrawal schedule is contained within the supplements of the topical report. All plant specific specimen capsules have been withdrawn from the reactor vessel. The ART values were not calculated using surveillance data (Reference 5.14) since it was determined to be non-credible.

3.4 Low Temperature Overpressure Protection (L TOP) limits are addressed in Section 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 determined using ASME Section XI, Appendix G, as modified by the alternative mles provided in ASME Code Case N-588 and ASME Code Case N-640 (Reference 5.9).

3.5 Table 1 provides the Adjusted Reference Temperature (ART) for each reactor vessel beltline material. The ART values were calculated in accordance with Regulatory Guide 1.99, Revision 2. For welds in the reactor beltline region, the initial RTNDT values used (in part) to detennine ART were calculated using an alternate methodology described in the NRC-approved BA W-2308, Revisions 1-A and 2-A (Reference 5.10). The NRC required licensees to obtain an exemption from 10 CFR 50.61 and IO CFR 50, Appendix G to use the alternate initial RTNDT values provided in BAW-2308 Revisions 1-A and 2-A. The required exemption was granted by the NRC in Reference 5.15. The NRC confirmed the limits and conditions for using the methodology were satisfied in the SER of Amendment 282 (Reference 5.8).

3.6 The Pressure-Temperature (PIT) limits of Section 2 and Figures 1, 2, 3, an<l 4 (with applicability as stated in 3.1) were generated consistent with the requirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99, Revision 2, using the methods described in BAW-10046A (Reference 5.4) and ASME Section XI, Appendix G (Reference 5.9), as modified by the alternative rules provided in ASME Code Case N-588 and ASME Code Case N-640.

43.5 EFPY PTLR Rev.4 Page 9 of 12 3.6.1 The NRC has reviewed the methods described in BAW-l0046A (Reference 5.4) and approved the topical repo1i by issuance of a Safety Evaluation Report (SER) dated April 30, 1986. Section 1.2 of BA W-10046A states that it is applicable to all current B& W nuclear steam systems.

3.6.2 ASME Code Cases N-640 and N-588 have been incorporated into ASME Section XI, Appendix G, 2003 Addenda, which are the edition and addenda codified in 10 CFR 50.55a (effective May 27, 2008) and thus may be used per NRC Regulatory Issue Summary (RlS) 2004-04. Specific approval for application at DBNPS is included in Ref. 5.8.

3.7 The minimum temperature requirements ofl0 CFR 50, Appendix Gare included on Figures 1, 2, 3 and 4. Figures 3 and 4 provide the In-Service Leak and Hydrostatic (ISLH) Test Limits. These limits were calculated in accordance with the requirements of 10 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.

3.8 Davis-Besse has removed 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 surveillance data. The Measured ~RTNoT-Predicted ~RTNoT data scatter was less than 2cr, so the Regulatory Guide 1.99, Rev. 2 Chemistry Table values used in the ART calculations are conservative.

4.0 PTLR Requirements 4.1 The PTLR has been prepared in accordance with the requirements of Technical Specification 5.6.4 (see Reference 5.11 for plant implementation). The PTLR shall be provided to the NRC upon issuance for each reactor vessel fl uencc period and for any revision or supplement thereto. Davis-Besse will continue to meet the requirements of 10 CFR 50, Appendix G, and any changes to the Davis-Besse Pff limits will be generated in accordance with the NRC approved methodologies described in TS 5.6.4.

Table1: Davis-Besse Nuclear Power Station Reactor Vessel Beltline Region Data

{Applicable as noted in Section 3.1)

Wetted hmer Surface Flucncc, n/cm' (E > 1.0 ART MeV) Current Projected EFPY

@1/4 T Licensing B!U;il:i to Reach CLB

(°F)

Reactor Vessel Material (CLB)@52 Fluence

@52 EFPY Location Identification EFPY (Ref. 5. 16)

(Ref. 5.16)

(Note I)

Nozzle Belt forging ADB 203 2.20E+18 47.4 67.1 Nozzle Belt to Upper Shell Weld WF-232 2.17E+l8 45.9 Note 3 (ID9%)

Nozzle Belt to Upper Shell Weld Wf-233 2.17F.+18 45.9 98.3 (OD 91%)

Upper Shell AKJ233 l.64E+l9 51.4 71.4 Forging Upper Shell to Lower Shell WF-182-1 l.64E+l9 51.4 150.9 Weld Lower Shell BCC 241 1.64E+19 51.4 89.5 forging I

Note 1: Reported ART values (Ref. 5.7) are based on Regulatory Guide 1.99, Revision 2.

Note 2: Values from Ref. 5.16, which arc based on the location specific clad to vessel interface fluence at 52 EFPY.

Note 3: This weld material does not extend out to the 1/4Tor 3/4T location.

Note 4: This weld material is not present at the clad to vessel interface. so RT PTS doe5 DOI apply to iL ART

@3/4 T (Of)

@52 EFPY (Note I) 58.7 Note 3 66.2 56.9 101.2 78.4 43.5 EFPY PTLR Rev.4 Page 10 of 12 Limiting RTPTS Mat'l?

(Of)

(Yes/No)

(Note 2)

No 82.0 No 120.5 No Note4 No 79.2 Yes 181.J No 95.5 Note 5: BWW 279/BWW 249 (Inlet Nozzle Forgings (at Lower Nozzle Helt to Inlet Nozzle Forging weld)) and WF-232/\\Vf-233 (Lower Nozzle Bell to Inlet Nozzle Forging Welds) are projected to reach the CLB fluence at 43.5 EFPY (Ref. 5.16).

5. 0 References 43.5 EFPY PTLR Rev.4 Page 11 of 12 5.1 Safety Evaluation by the NRC Office of Nuclear Reactor Regulation Related to Amendment No. 199 to Facility Operating License No. NPF-3 Davis-Besse Nuclear Power Station, Unit No. 1, attached to correspondence dated July 20, 1995.

5.2 Technical Specification 5.6.4, Revision 339, "Reactor Coolant System (RCS)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."

5.3 Technical Specification 3.4.12, Revision 339, "Low Temperature Overpressure Protection."

5.4 BAW-10046A, Revision 2 "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G."

5.5 BAW-2241P-A, "Fluence and Uncertainty Methodologies," dated April 1999.

5.6 BAW-1543A, "Master Integrated Reactor Vessel Material Surveillance Program.

5.7 ANP-2718, Revision 7, "Appendix G Pressure-Temperature Limits for 52 EFPY for Davis-Besse Nuclear Power Station FirstEnergy Nuclear Operating Company," dated April 2019.

5.8 Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 282 to Facility Operating License No. NPF-3, FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station, Unit No. 1, (FENOC Ltr. Rl 1-030), dated 01/28/2011.

5.9 ASME Code Section XI, Appendix G, as modified by the alternate rules provided in ASME Code Case N-640 and ASME Code Case N-588. ASME Code Cases N-640 and N-588 have subsequently been incorporated into ASME Section XI, Appendix G, 2003 Addenda, which are the edition and addenda codified in 10 CFR 50.55a (effective May 27, 2008).

5.10 BAW-2308, Revision 1-A and Revision 2-A, "Initial RTNoT of Linde 80 Weld Materials," dated August 2005 (1-A) and March 2008 (2-A).

5.11 Calculation C-NSA-064.02-037, Revision 3, "Davis-Besse 43.5 EFPY Pressure-Temperature Limits," dated 7/7/2021.

5.12 Not used 5.13 Not used 5.14 AREVA Document 32-9031157-000, "Davis-Besse Revised ART Values at 52 EFPY," dated 9/20/2006.

5.15 NRC Letter to FirstEnergy Nuclear Operating Company, "Davis-Besse Nuclear Power Station, Unit I-Exemption from the Requirements of 10 CFR Part 50.61 and 10 CFR Part 50, Appendix G," (FENOC Ltr. Rl0-298) dated December 14, 2010.

5.16 AREVA Document 32-9300671-000, "Davis-Besse Reactor Vessel Embrittlement Fluence Reconciliation Through 60 Years," dated 12/10/2019.

5.17 CR 2019-03982 "Reactor Vessel 52 EFPY Projected Neutron Fluence (Exposure)

Higher than Previous Projection," dated 4/30/2019.

43.5 EFPY PTLR Rev. 4 Page 12 of 12

5. 18 CR 2020-07840 "Revised Vessel P-T Curves Limited to 43 EFPY," dated I 0/9/2020