NL-15-0418, Revision 6, Pressure Temperature Limits Reports

Revision 6, Pressure Temperature Limits Reports

ML15064A023
Person / Time
Site:	Farley
Issue date:	03/05/2015
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-15-0418
Download: ML15064A023 (50)

Text

Charles A. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 A

Birmingham, AL 35201 Tel 205.992.7872 SOUTHERN Fax 205.992.7601 COMPANY March 5, 2015 Docket Nos.: 50-348 NL-15-0418 50-364 U. S. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Unit 1 Revision 6 Pressure Temperature Limits Report Unit 2 Revision 6 Pressure Temperature Limits Report Ladies and Gentlemen:

In accordance with Technical Specification 5.6.6.c, Southern Nuclear Operating Company submits the enclosed Reactor Coolant System (RCS) Pressure Temperature Limits Report (PTLR) Revision 6 for Unit 1 and Revision 6 for Unit 2.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

C. R. Pierce Regulatory Affairs Director CRP/RMJ

Enclosures:

1. Unit 1 PTLR, Revision 6

2. Unit 2 PTLR, Revision 6

U. S. Nuclear Regulatory Commission Nl 0418 Page 2 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bast, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President- Farley Mr. M. D. Meier, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. R. R. Martin, Regulatory Affairs Manager- Farley RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. S. A. Williams, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident Inspector- Farley

Joseph M. Farley Nuclear Plant Unit 1 Revision 6 Pressure Temperature Limits Report Unit 2 Revision 6 Pressure Temperature Limits Report Enclosure 1 Unit 1 PTLR, Revision 6

SOUTHERN COMPANY A

Energy to Serve l'Our World" Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 1 Revision 6 November 2014

PTLR for FNP Unit 1 Revision 6 Page 1 of 22 Table of Contents List of Tables oooooooooooooo oooooooooo o o o ooooooooooooooooo o o o o o o o o o o o ooooooooooooooooooooooo o oooooooooooooooooooooooooooooooooooooo 2 List of Figures 0 0 0 0 0 00 0 00 0 0 0 0 0 0 00 0 0 0 0 0 0 0 0 0 0 0 0 00 0 0 0 0 0 00 0 00 0 0 00 0 0 0 000 00 0 0 0 0 0 00 0 0 0 00 0 Oo 0 0 0 0 0 00 0 0 0 0 0 0 0000 0 0 0 00 0 0 0 0 0 00 0 0 00 0 0 Oo0 0 0 0000 0 0 0 3 1 0 0 RCS Pressure Temperature Limits Report (PTLR) oooooooooooooooooooooooooooooooooooooooooooooooo 5 20 0 Operating Limits oooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooo 5 20 1 RCS PressurefTemperature (PfT) Limits (LCO - 3.403) oooooooooooooooooooooooooooooooooooo 5 20 2 RCP Operation Limits ooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooooo 5 20 3 LTOP System Applicability Temperature (LCO- 3 4 12) . 0 oooooooo ooooooo oooooooooooooooooooo 5 30 0 Reactor Vessel Material Surveillance Program ooooooooooooooooooooooooooooooo ooooooooooooooooooooo 12 40 0 Reactor Vessel Surveillance Data Credibility 0 0000 0 0 0 0 0 0000000000000000000000000 oooooooooooooooooooo 13 500 Supplemental Data Tables 0 oooooooooooooooooooooooooooooooo oooooooooooooooooo 000 0000 000 0000000 0 0 000000000 0000 13 600 References oooooooo o o o ooooooooooooooooooooooooooo o o ooooooooo o o o oooooooooooooooo o o o o ooooooooooooooooooooooooooooooooooo 21

PTLR for FNP Unit 1 Revision 6 Page 2 of 22 List of Tables 2-1 Farley Unit 1 54 EFPY Heatup Curve Data Points .... . ............ . ...... ............ . .......... . ... 8 2-2 Farley Unit 1 54 EFPY Cooldown Curve Data Points . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0 3-1 Surveillance Capsule Withdrawal Schedule ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 5-1 Comparison of Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1 .99, Revision 2, Predictions .. . ........................................ . . . . . ................................. . . . . . . . . . . 14 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data .... . ............ ..... 15 5-3 Reactor Vessel Toughness Table (Unirradiated) . . ............. . ........ . ........................... 16 5- 4 Reactor Vessel Fluence Projections at 54 EFPY .... . ......... ... ...... . .

. ........ . ... 17 5-5 Summary of ART Values for the Reactor Vessel Materials at the 1 /4-T and 3/4-T Locations for 54 EFPY ........ . . ............................. . .

. ..... 18 5-6 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Reactor Vessel Material ................... ......... . .............. . .......... . ....... . ........ . .... 19 5-7 Pressurized Thermal Shock (RTPTs) Values for 54 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

PTLR for FNP Unit 1 Revision 6 Page 3 of 22 List of Fi g ures 2-1 Farley U nit 1 Reactor Coolant System Heatup Limitations . . . . . . .. . . .. . . . . . .. . . . . . . . . . . ..... . . . . . 6 2-2 Farley Unit 1 Reactor Coolant System Cooldown Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 7

PTLR for FNP Unit 1 Revision 6 Page 4 of 22 This page intentionally blank.

PTLR for FNP Unit 1 Revision 6 Page 5 of 22 1.0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for Farley Nuclear Plant - Unit 1 has been prepared in accordance with the requirement of Technical Specification (TS) 5.6.6. Revisions to the PTLR shall be provided to the NRC after issuance.

This report affects TS 3.4.3, RCS Pressure!Temperature Limits (PfT) Limits. All TS requirements associated with low temperature overpressure protection (LTOP) are contained in TS 3 .4. 1 2, RCS Overpressure Protection Systems.

2.0 Operating Limits The limits for TS 3.4.3 are presented in the subsection which follows and were developed using the methodologies specified in TS 5.6.6. The methodologies are contained in WCAP-1 4040-A, Revision 4111* The operability requirements associated with LTOP are specified in TS LCO 3.4. 1 2 and were determined to adequately protect the RCS against brittle fracture in the event of an LTOP transient. The limitation on the number of operating reactor coolant pumps (RCPs) is necessary to assure operation consistent with the pressure corrections incorporated in the P!T limits for flow losses associated with the RCPs.

2. 1 RCS Pressure!Temperature (PfT) Limits (LCO - 3.4.3)

2. 1 . 1 The minimum boltup temperature is 60°F.

2 . 1 .2 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 1 00°F in any one hour period.

b. A maximum cooldown of 1 00°F in any one hour period.

c. A maximum temperature change of less than or equal to 1 0°F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1 .3 The RCS P!T limits for heatup a n d cooldown are specified b y Figures 2 - 1 a n d 2-2, respectively.

2.2 RCP Operation Limits 2.2.1 The number of operating RCPs i s limited to one a t RCS temperatures less than 1 1 ooF with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

2.3 LTOP System Applicability Temperature (LCO- 3 .4. 1 2) 2.3.1 The Low Temperature Overpressure Protection (LTOP) System applicability temperature is 275°F.

The Farley Nuclear Plant-Unit 1 pressure-temperature limits for the reactor vessel inlet/outlet nozzle corner region and balance of reactor coolant system ferritic components were evaluated and found to be non-limiting with respect to the reactor vessel beltline pressure-temperature limits (Figures 2-1 and 2-2 below). This technical assessment was confirmed by the NRC via Safety Evaluation Report !21> .

PTLR for FNP Unit 1 Revision 6 Page 6 of 22 Figure 2*1 Farley Unit 1 Reactor Coolant System Heatup Limitations121 (Heatup Rates up to 1 00°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi p at RCS temperatures 1 1 0°F and 27 psi p at RCS temperatures < 1 1 0°F). Includes vessel flange requirements per 10 CFR 50, Appendix Gl3f.

Limiting Material: Lower Shell Plate 869 1 9- 1 with non-credible surveillance data Limiting ART Values at 54 EFPY: 1 /4T = 1 91 oF 3/4T = 1 66°F 2500 I

1 I 2250 1----

I

_ ,_ 1- - --- 1-- -

LUnit 2000 --- -- I I

__ I_ - - - -- ----- -- ---

I Unacceptable Operation 1780 --- J - *--

v I Acceptable Operation I 1:500 J, fJ a-e 12tso I k 1--,l Crklcal I

Limit I

800eg. FJHr D..

1

=

1000 --- --

N Umtt_j -- --

I Critical H oatu p Rate u

c;;

u I GOOeg, F/Hr 100Do .F/Hr 7t50 vr!'

J

500 ""A:: I

,l ,

Hoatup R__

100Dog. FIHr 2t50 - Crtdcallly Limit baaed on *--- ---

i+=

lnsorvice hydroatatlc cost I

Bop temperature (25&*..-, for Cho ITt Temp.* aorvlce .orlod up to 54 EFPY so*F 0 --

--RCS proaaure Ia -14.7 pa1g J

--*--.-- -*--*-* 1

-*--*- --- --.-h -.-*-* --- -

0 50 100 1 t50 200 250 300 3t50 400 4t50 500 550 Moderator Temperature (Deg. F)

PTLR for FNP Unit 1 Revision 6 Page 7 of 22 Fi g ure 2*2 Farley Unit 1 Reactor Coolant System Cooldown Limitations121 (Cooldown Rates up to 1 00°F/hr) Applicable to 54 EFPY (adjusted to incl ude 60 psi p at RCS temperatures 1 1 0°F and 27 psi p at RCS temperatures < 1 1 0°F). Includes vessel flange requirements per 10 CFR 50, Appendix Gl3f.

Limiting Material: Lower Shell Plate 869 1 9- 1 with non-credible surveillance data Limiting ART Values at 54 EFPY: 1 /4T = 1 9 1 °F 3/4T = 1 66°F 2SOO 2250 *--- ---- I 2000 . --- -  !----- -

- -I -- --

Unacceptable Operation 1750 ----* - -*--

--j 1500 . --

I I Acceptable Operation S2 f! 1250 v_

I 111.

! 1000 -- -- --

u lj 760 - -- - -- --

- - ---- ----- --*- -- 1-*

Cooldown Retoa,*F/Hr 260

)Bop "f:: - t'\ 20

• t.eady-wtete

-40 I

..ao Temp. i' -100

-eo*F 0 - l -

I I ,

The lower limit for RCS pro**urola -14.7 palg 1- -*-1-.-* .-1 -* .L A - .. .- J-.r-'""1

• • __ ,_ !--.--.-- - --- 1-..---. -.---.

• 0 50 100 150 200 250 300 350 400 4SO 500 550 Moderator Temperature (Deg. F)

PTLR for FNP Unit 1 Revision 6 Page 8 of 22 Table 2-1 Farley Unit 1 - 54 EFPY Heatup Curve Data Pointsl2l (adjusted to include 60 psi il P at RCS temperatures 1 1 0°F a nd 27 psi il P at RCS temperatures < 1 1 0°F) 60°F/hr. 60°F/hr. 1 00°F/hr. 1 00°F/hr.

Leak Test Limit Heat u p Criticality Heatu p Criticality T P (psig) o T ( F) p T p T p T p

( o F) (psig) (o F) (psig) ( o F) ( ps ig) ( o F) (ps i g )

229 2000 60 -14.7 247 -14.7 60 -14.7 247 -14.7 247 2485 60 594 247 561 60 574 247 541 65 594 247 561 65 574 247 541 70 594 247 561 70 574 247 542 75 594 247 561 75 574 247 542 80 594 247 561 80 574 247 545 85 594 247 561 85 574 247 545 90 594 247 561 90 574 247 549 95 594 247 561 95 574 247 550 100 594 247 561 100 574 247 554 105 594 247 561 105 574 247 556 110 594 247 561 110 574 247 561 110 561 247 561 110 541 247 561 115 561 247 561 115 541 247 561 120 561 247 561 120 542 247 561 125 561 247 561 125 545 247 561 130 561 247 561 130 549 247 561 135 561 247 561 135 554 247 561 140 561 247 561 140 561 247 561 145 561 247 561 145 561 247 561 150 561 247 561 150 561 247 561 155 561 247 561 155 561 247 561 160 561 247 561 160 561 247 561 165 561 247 561 165 561 247 561 170 561 247 561 170 561 247 561 175 561 247 822 175 561 247 670 180 561 247 855 180 561 247 693 180 561 247 892 180 561 247 718 180 822 247 932 180 670 247 746 185 855 247 977 185 693 247 777 190 892 247 1049 190 718 247 826

PTLR for FNP Unit 1 Revision 6 Page 9 of 22 Table 2-1 (continued)

Farley Unit 1 - 54 EFPY Heatup Curve Data Pointsl21 (adjusted to include 60 psi LlP at RCS temperatures 1 1 0°F and 27 psi LlP at RCS temperatures < 1 1 0°F)

Leak Test 60° F/hr. 1 00°F/hr.

60°F/hr. Heatu p 1 00°F/hr. Heat up Lim it Criticality Criticality T p T T T T p P (ps lg) P (psig) P (psig)

(o F) (psig) (o F) ( o F) ( o F) ( o F) ( ps l g) 195 932 250 1082 195 746 250 849 200 977 255 1143 200 777 255 891 205 1027 260 1210 205 811 260 938 210 1082 265 1284 210 849 265 990 215 1143 270 1365 215 891 270 1046 220 1210 275 1455 220 938 275 1109 225 1284 280 1552 225 990 280 1179 230 1365 285 1638 230 1046 285 1255 235 1455 290 1734 235 1109 290 1340 240 1552 295 1839 240 1179 295 1433 245 1638 300 1955 245 1255 300 1536 250 1734 305 2084 250 1340 305 1650 255 1839 310 2225 255 1433 310 1 775 260 1955 315 2382 260 1536 315 1913 265 2084 265 1650 320 2066 270 2225 270 1775 325 2234 275 2382 275 1913 330 2419 280 2066 285 2234 290 2419

PTLR for FNP Unit 1 Revision 6 Page 1 0 of 22 Table 2-2 Farley Unit 1 - 54 EFPY Cooldown Curve Data Pointsl21 (adjusted to i nclude 60 psi p at RCS temper atur es 11 0° F and 27 psi p at RCS temperatures < 11 0°F)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. 1 00°F/hr.

T(°F) P (psig) T(°F) P (psi g) T(°F) P (psig) T(°F) P (psig) T(°F) P ( psig) 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 594 60 594 60 558 60 515 60 427 65 594 65 594 65 561 65 519 65 431 70 594 70 594 70 565 70 522 70 435 75 594 75 594 75 569 75 527 75 440 80 594 80 594 80 574 80 531 80 445 85 594 85 594 85 578 85 537 85 451 90 594 90 594 90 584 90 543 90 458 95 594 95 594 95 590 95 549 95 465 100 594 100 594 100 594 100 556 100 474 105 594 105 594 105 594 105 565 105 483 110 594 110 594 110 594 110 574 110 494 110 561 110 561 110 561 110 541 110 461 115 561 115 561 115 561 115 551 115 473 120 561 120 561 120 561 120 561 120 486 125 561 125 561 125 561 125 561 125 500 130 561 130 561 130 561 130 561 130 517 135 561 135 561 135 561 135 561 135 535 140 561 140 561 140 561 140 561 140 555 145 561 145 561 145 561 145 561 145 561 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 175 561 175 561 175 561 175 561 175 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 561 180 900 180 880 180 862 180 846 180 823 185 934 185 917 185 902 185 890 185 875 190 971 190 957 190 946 190 938 190 933 195 1012 195 1002 195 995 195 991 195 991 200 1058 200 1051 200 1049 200 1049 200 1049 205 1108 205 1106 205 1106 205 1106 205 1106 210 1164 210 1164 210 1164 210 1164 210 1164 215 1225 215 1225 215 1225 215 1225 215 1225

PTLR for FNP Unit 1 Revision 6 Page 1 1 of 22 Table 2-2 (continued)

Farley Unit 1 -54 EFPY Cooldown Curve Data Point s[2J (adjusted to include 60 psi p at RCS temperatures ;::: 1 1 0° F and 27 psi p at RCS temperatures < 1 1 0°F)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. 1 00°F/hr.

T (°F) P (psig) T (°F) P (ps ig) T (°F) P (ps ig) T (°F) P (ps ig) T (°F) P (psig) 220 1293 220 1293 220 1293 220 1293 220 1293 225 1368 225 1368 225 1368 225 1368 225 1368 230 1451 230 1451 230 1451 230 1451 230 1451 235 1542 235 1542 235 1542 235 1542 235 1542 240 1644 240 1644 240 1644 240 1644 240 1644 245 1756 245 1756 245 1756 245 1756 245 1756 250 1879 250 1879 250 1879 250 1879 250 1879 255 2016 255 2016 255 2016 255 2016 255 2016 260 2167 260 2167 260 2167 260 2167 260 2167 265 2334 265 2334 265 2334 265 2334 265 2334

PTLR for FNP Unit 1 Revision 6 Page 1 2 of 22 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is i n compliance with 1 0 CFR 50, Appendix Hl41, and is described in Section 5.4.3.6 of the Farley FSAR. Surveillance capsules are tested and the results 5

reported in accordance with ASTM E 1 85-821 *1 The removal schedule is provided in Table 3-1 . The neutron transport and dosimetry evaluation methodologies used follow the gu idance and meet the requirements of Regulatory Guide 1 . 1 90, "Calculational and Dosimetry Methods for Determining P ressure Vessel Neutron Fluence"161* The results of the Capsule Z examination (WCAP- 1 6964-NP, Revision 0[71) were used to produce Figures 2-1 and 2-2.

Table 3-1 Surveillance Capsule Withdrawal Schedule <a>

Ca ps ule Lead Removal Fluence Ca ps ule Location Factor EFPY !bl (n/cm 2)

(Degreej y (C)

6. 1 2 x 1 01ts

)

343 3.24 1.15 U (c 9

1 07 3.34 3.08 1 .73 x 1 0 1 X (C

) 1 J

287 3.35 6.1 1 3.06 X 1 0 w<c 1

1 10 3.0 1 1 2.43 4.75 X 1 0 9(dJ v<c> 7 . 1 4 X 1 01\eJ

)

3 . 04 z<c 290 20. 1 6 340 3.04 24.26 8.47 x 1 0 19111 Notes:

71 a) Data from Table 7-1,WCAP- 16964-NP, Revision 0 1 b) Effective Full Power Years (EFPY) from plant startup.

c) Plant-specific evaluation.

d) This fluence is not less than once or greater than twice the peak EOL fluence for the initial 40-year license term.

e) This fluence is not less than once or greater than twice the peak EOL fluence for a 20-year license renewal term to 60 years.

f) This fluence is not less than once or greater than twice the peak EOL fluence for an additional 20-year license renewal term to 80 years.

PTLR for FNP Unit 1 Revision 6 Page 1 3 of 22 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1 .99, Revision 2181 , describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1 .99, Revision 2, describes the methodology for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

Per WCAP-88 1 0191 , the Unit 1 surveillance program was based on ASTM E 1 85-7311 01 . All six surveillance capsules (YI1 11 , U 1121 , X 1131 , W l141 , V 1151 , and Z 171) have been removed from the Farley Unit 1 reactor vessel and analyzed. In accordance with the discussion of Regulatory Guide 1 .99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The credibility conclusions for the Farley Unit 1 surveillance plate and weld are described below:

The credibility evaluation of the Farley Unit 1 Lower Shell Plate 8691 9-1 surveillance material is documented in Appendix D of WCAP-1 6964-NP 171 . The credibility evaluation concluded that the surveillance data for Lower Shell Plate 8691 9-1 is non-credible.

The credibility evaluation for weld Heat # 33A277 surveillance data is documented in Appendix D of WCAP-1 7365-NP1161 . The evaluation takes into account surveillance data from Calvert Cliffs U nit 1 and Farley Unit 1 . The credibility evaluation concluded that the surveillance data for weld Heat # 33A277 is credible.

5.0 Su pp lemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1 .99, Revision 2, predictions.

Table 5-2 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5-3 provides the unirradiated Farley Unit 1 reactor vessel toughness data.

Table 5-4 provides a summary of the reactor vessel fluence values at 54 EFPY.

Table 5-5 provides a summary of the ART values of the Farley U nit 1 reactor vessel materials at the 1 /4-T and 3/4-T locations for 54 EFPY.

Table 5-6 shows the calculation of the ART values at 54 EFPY for the limiting Farley Unit 1 reactor vessel material (lower shell plate 869 1 9- 1 ) .

Table 5-7 provides RTPTs values for Farley Unit 1 for 5 4 EFPY.

PTLR for FNP Unit 1 Revision 6 Page 1 4 of 22 Table 5-1 Comparison of Surveillance Material 30 ft-lb Transition Temperature Shift and Upper Shelf E nergy Decrease with Regulatory G uide 1 .99, Revision 2, Predictions <a>

30 ft-lb Tran s ition Upper Shelf Energ y Tem perature Shift Decreas e Fluence Predicted Meas ured Predicted Meas ured Material Ca ps ule (1019 n/cm2, (o F ) (oF) (%) (%)

E > 1 .0 MeV) y 0.612 84.3 64.6 21 7 u 1.73 112.5 110.0 27 21 X

Lower Shell 3.06 126.7 129.2 30 17 Plate 86919-1 w 4.75 136.2 145.7 35 20 v

(Longitudinal) 7.14 143.4 177.7 39 21 z 8.47 145.9 202.2 40 29 y 0.612 84.3 70.1 21 1 u 1.73 112.5 100.4 27 9 X

Lower Shell 3.06 126.7 110.8 30 12 Plate 86919-1 w 4.75 136.2 150.5 35 17 (Transverse) v 7.14 143.4 161.7 39 21 z 8.47 145.9 178.3 40 23 y 0.612 67.4 66.9 24 3 u 1.73 89.9 75.1 31 22 X

Surveillance 3.06 101.2 87.4 36 15 w

Program 4.75 108.7 98.3 40 18 Weld Metal v 7.14 114.5 117.5 44 18 z 8.47 116.5 113.5 46 25 y 0.612 --- 29.2 --- 7 u 1.73 - 155.3 21 X

Heat Affected 3.06 - 132.9 - 15 w

Zone Material 4.75 -- 121.7 - 11 v

7.14 --- 169.7 20 z

8.47 -- - 170.7 20 NOTE:

171 (a) Data from Table 5-10 ofWCAP-16964-NP, Revision 0

PTLR for FNP Unit 1 Revision 6 Page 1 5 of 22 Table 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data Ca ps ule f ARTNoT FF*&RTNoT Material Ca ps ule (1 019 n/cm 2, FF181 FF 2 (o F) ( o F)

E > 1 .0 MeV) y 0.612 0.862 64.6 55.7 0.744 u 1.73 1.151 110.0 126.6 1.324 X

Lower Shell 3.06 1.295 129.2 167.1 1.678 Plate 86919-1 (bl w 4.75 1.392 145.3 202.3 1.938 v

(Longitudinal) 7.14 1.466 177.7 260.5 2.149 z 8.47 1.492 202.2 301.6 2.225 y 0.612 0.862 70.1 60.5 0.744 u 1.73 1.151 100.4 115.5 1.324 X

Lower Shell 3.06 1.295 110.8 143.5 1.678 Plate 86919-1 (bl w 4.75 1.392 150.5 209.5 1.938 v

(Transverse) 7.14 1.466 161.7 237.1 2.149 z 8.47 1.492 178.3 266.0 2.225 SUM: 2146.20 20.118 CFass1s-1=:E (FF*RTNoT ) + :E( FF2)=(2146.20)+ (20.118)=106.7°F y 0.612 0.862 108.38 (66.9)(d) 93.47 0.744 Farley Unit 1 u 1.73 1.151 121.66 (75.1)(d) 140.00 1.324 Surveillance Weld X 3.06 1.295 141.59 (87.4)(d) 183.42 1.678 Material w 4.75 1.392 159.25 (98.3)(d) 221.69 1.938 v

(Heat # 33A277)(cJ 7.14 1.466 190.35 (117.5)(d) 279.07 2.149 z 8.47 1.492 183.87 (113.5)(d) 274.30 2.225 Calvert Cliffs Unit 1 263° 0.505 0.809 81.97 (50.4)(d) 66.34 0.655 Surveillance Weld 9r 1.94 1.181 156.63 (104.5)(d) 185.00 1.395 Material (Heat # 33A277)(cJ 284° 2.33 1.228 120.06 (78.0)(d) 147.49 1.509 SUM: 1590.78 13.618 CF Surv. Wel =L(FF* RTNOT) + L( FF2)=(1590.78)+ (13.618)= 116.8°F d

NOTES:

(028 - 0"11o g(lf l (a) FF = f luence factor = f .

m (b) Information pertaining to Lower Shell Plate 86919-1 is taken from Table D-1 ofWCAP-16964-NP .

117J (c) Information pertaining to weld Heat# 33A277 is taken from Table 6.1-1 of WCAP-17506-NP .

(d) To calculate the best fit chemistry factor (CF) as provided by Reg. Guide 1.99, Rev. 2, Position 2.1, the surveillance weld LlRTNOT values have been adjusted to account for chemistry differences between the reactor vessel weld and the surveillance weld. For the Calvert Cliffs Unit 1 data, the surveillance weld LlRT NOT values have also been adjusted to account for the temperature difference between the Farley Unit 1 and Calvert Cliffs Unit 1 reactor vessels. Pre-adjusted 117J values are in parentheses. See Table 6.1-1 ofWCAP-17506-NP for all details pertaining to the chemistry factor calculation for weld Heat# 33A277.

PTLR for FNP Unit 1 Revision 6 Page 1 6 of 22 Table 5-3 Reactor Vessel Toughness Table (Unirradiated) <aJ Beltline Material Cu Wei g ht Ni Wei g ht IRTNDT

%  % ( oF )

Closure Head Flange -- --

-5o<d>

Vessel Flange -- -- 60 Inlet Nozzle 86917-1 0.16 0.83 60 Inlet Nozzle 86917-2 0.16 0.80 60 Inlet Nozzle 86917-3 0.16 0.87 60 Outlet Nozzle 86916-1 0.16 0.77 60 Outlet Nozzle 86916-2 0.16 0.78 60 Outlet Nozzle 86916-3 0.16 0.78 60 Upper Shell Forging 86914 0.16 0.684 30 Intermediate Shell Plate 86903-2 0.13 0.60 0 Intermediate Shell Plate 86903-3 0.12 0.56 10 Lower Shell Plate 86919-1 0.14 0.55 15 Lower Shell Plate 86919-2 0.14 0.56 5 lnleUOutlet Nozzle to Upper Shell Girth Seams 0.04 1.08 10(e) 1-897 A-F Upper Shell to Intermediate Shell Circumferential 0.197 0.06 -56(1)

Weld Seam 10-894 (Heat # 90099) (bl Intermediate Shell Longitudinal Weld Seams19-894 A & 8 (Heat # 33A277) (bl 0.258 0.165 -56(1)

Surveillance Weld <c> 0.14 0.19 - -

I ntermediate Shell to Lower Shell Circumferential 0.205 0.105 -56(1)

Weld Seam 11-894 (Heat # 6329637) (b)

Lower Shell Longitudinal Weld Seams20-894 A & 8 (Heat # 90099) < bl 0.197 0.060 -56(1)

NOTES:

1171 (a) From Table 4.1-1 ofWCAP-17506-NP .

16 (b) Best-estimate copper and nickel from CE NPSD-10391 1.

(c) The surveillance weld is representative of intermediate shell longitudinal welds19-894 A & B. Best-estimate copper and nickel values represent a single chemical analysis documented inWCAP-8810, Revision o191.

1191 (d) Replacement closure head initial RT NOT value was taken from MHI-SNC-0195 .

(e) An estimation method using measured data was used to determine this initial RTNoT value. Therefore, a conservative value of 17"F is used for au and 01 in margin calculations.

(f) These initial RT NOT values are generic and taken from 10 CFR 50.61 paragraph (c)(1)(ii) of the 1-1-07 edition.

PTLR for FNP Unit 1 Revision 6 Page 1 7 of 22 Table 5-4 Reactor Vessel Fluence Projections at 54 EFPY <a>

(1 019 n/cm2, E > 1 .0 MeV) 54 EFPY Reactor Vess el Location Material Neutron Fluence I nlet Nozzle 8691 7- 1 0.0349 I n let Nozzle 8691 7-2 0.01 90 I nlet Nozzle 8691 7-3 0.01 39 Outlet Nozzle 86916-1 0. 00922 Outlet Nozzle 8691 6-2 0 . 0 1 26 Outlet Nozzle 8691 6-3 0.0231 Upper Shell Forging 869 1 4 1 .02 86903-2

& 86903-3 Intermediate Shell Plates 5.93 8691 9- 1

& 8691 9-2 Lower Shell Plates 5.81 I nlet/Outlet Nozzle to Upper Shell 1 -897 A-+F 0.0349 Girth Seams Upper Shell to I ntermediate Shell 1 0-894 1 .02 Circumferential Weld Seam (Heat # 90099)

I ntermediate Shell Longitudinal Weld 1 9-894 A & 8 1 .83 Seams (Heat # 33A277)

I ntermediate Shell to Lower Shell 1 1 -894 5.81 Circumferential Weld Seam (Heat # 6329637)

Lower Shell Longitudinal Weld 20-894 A & 8 1 . 79 Seams (Heat # 90099)

NOTE:

I171 (a) From Table 5.1-1 ofWCAP-17506-NP . These values are also summarized in Table 2-1 of Attachment A of ALA 20 116! 1.

PTLR for FNP Unit 1 Revision 6 Page 1 8 of 22 Table 5-5 Summary of ART Values for the Reactor Vessel Materials at the 1 /4T and 3/4T Locations for 54 EFPY (a) 1/4 T 3/4 T Material

( o F) (o F)

I nlet Nozzle 86917-1 1 05 84 I nlet Nozzle 86917-2 90 76 I nlet Nozzle 86917-3 85 72 O utlet Nozzle 86916-1 78 69 O utlet Nozzle 86916-2 83 71 Outlet Nozzle 86916-3 94 78 Upper Shell Forging 86914 1 69 1 39 I ntermediate Shell Plate 86903-2 1 56 1 34 Intermediate Shell Plate 86903-3 1 54 1 34 Lower Shell Plate 86919-1 1 79 1 56 Lower Shell Plate 86919-1 b 1 66 (b)

Using non-credible SIC Data 1 91 ( )

Lower Shell Plate 86919-2 1 70 1 47 l nleUOutlet Nozzle to Upper Shell Girth Seams 55 50 1-897 A-+F Upper Shell to Intermediate Shell Circumferential 89 66 Weld Seam 10-894 (Heat # 90099)

Intermediate Shell Longitudinal Weld Seams19-894 A & 8 (Heat # 33A277) 1 40 1 07 Intermediate Shell Longitudinal Weld Seams19-894 A & 8 (Heat # 33A277) 111 80 Using credible S/C Data Intermediate Shell to Lower Shell Circumferential Weld Seam 11-894 (Heat # 6329637) 1 41 1 17 Lower Shell Longitudinal Weld Seams20-894 A & 8 (Heat # 90099) 1 04 80 NOTES:

(a) The ART values presented here are based on the reactor vessel surface fluence values summarized in Table 5-4.

12 The values for the beltline materials are from Tables 4-10 and 4-11 ofWCAP-17122-NP 1 . The values for the extended beltline materials are summarized along with the values for the beltline materials in Tables 3-3 and 3-4 of 1201 Attachment A of AL.A-09-116 .

(b) Limiting 1/4T and 3/4T ART values. The PIT limit curves are based on these limiting ART values of 191*F and 166*F.

PTLR for FNP Unit 1 Revision 6 Page 1 9 of 22 Table 5-6 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Reactor Vessel Material - Lower Shell Plate 869 1 9-1 Parameter Value Operating Period 54 EFPY Location 1/4 T 3/4 T Chemistry Factor, CF (°F) (a) 106.7 106 .7 2 (b)

Fluence, f (1 019 n/cm ) 3.622 1.408 0.2e.o.1"1og<fll Fluence Factor, FF = f( 1.3343 1 .0949 RT NOT = CF X FF (°F) 142.4 116.8 I nitial RTNOT. I (°F) (c) 15 15 Margin, M (°F) (d) 34 34 Adjusted Reference Temperature (ART), (°F) per 191 166 Regulatory Guide 1.99, Revision 2 <el NOTES:

(a) Chemistry factor is taken from Table 5-2.

(b) Fluence is based on fsurt = 5.81 x 1 019 n/cm (E 2

> 1.0 MeV), from Table 4-1 ofWCAP-17122-NP, Revision 0121. Farley Unit 1 reactor vessel wall thickness is 7.875 inches in the beltline region.

(c) Initial RT NOT value is taken from Table 5-3.

(d) Margin= 2( ol + o}) 0*5, ("F); for the lower shell plate 86919-1, 0'; = O"F and cr6 = 17"F.

(e) Per Regulatory Guide 1.99, Revision 2: ART ("F) = flRT NOT + I + M.

PTLR for FNP Unit 1 Revision 6 Page 2 0 of 22 Table 5-7 Pressurized Thermal Shock (RTPrs) Values for 54 E FPY (aJ Surface aRTNoT fluence I M RTPTS Material Cf ff (Cf x ff)

(10 1 9 n/cm 2, (Of) (Of) (Of)

(Of)

E > 1 .0 MeV)

I n let Nozzle 869 1 7-1 1 23.3 0.0349 0.2397 29.6 60 29.6 119 I nlet Nozzle 8691 7-2 1 23 0.01 90 0. 1 666 20.5 60 20. 5 1 01 I n let Nozzle 8691 7-3 1 23.7 0.01 39 0.1 365 1 6. 9 60 1 6. 9 94 Outlet Nozzle 8691 6-1 1 22.3 0.00922 0 . 1 037 1 2.7 60 1 2.7 85 O utlet Nozzle 8691 6-2 1 22.5 0.0126 0 . 1 280 1 5. 7 60 1 5. 7 91 Outlet Nozzle 8691 6-3 1 22.5 0.0231 0.1 879 23.0 60 23.0 1 06 Upper Shell Forging 869 1 4 1 20.1 1 .02 1 . 0055 1 20.8 30 34.0 1 85 Intermediate Shell Plate 86903-2 91 .0 5.93 1 .4345 1 30.5 0 34.0 1 65 Intermediate Shell Plate 86903-3 82.2 5.93 1 .4345 1 1 7. 9 10 34.0 1 62 Lower Shell Plate 8691 9-1 97.8 5.81 1 .4308 1 39.9 15 34. 0 1 89 202(b)

Lower Shell Plate 8691 9-1 1 06.7 5.81 1 .4308 1 52.7 15 34.0 Using non-credible SIC Data Lower Shell Plate 8691 9-2 98.2 5.81 1 .4308 1 40.5 5 34.0 1 80 Inlet/Outlet Nozzle to Upper Shell 54 0. 0349 0.2397 1 2. 9 10 36.4 59 G irth Seams 1 -897 A- F Upper Shell to I ntermediate Shell Circumferential Weld Seam 1 0-894 91 .4 1 .02 1 . 0055 91 .9 -56 65.5 1 01 (Heat # 90099)

I ntermediate Shell Longitudinal Weld Seams 1 9-894 A & 8 1 26.3 1 .83 1 . 1 657 1 47.2 -56 65.5 1 57 (Heat # 33A277)

Intermediate Shell Longitudinal Weld Seams 1 9-894 A & 8 1 1 6.8 1 .83 1 . 1 657 1 36 .2 -56 44.0 1 24 (Heat # 33A277)

Using credible SIC Data Intermediate Shell to Lower Shell Circumferential Weld Seam 1 1-894 98.4 5.81 1 .4308 1 40.8 -56 65.5 1 50 (Heat # 6329637)

Lower Shell Longitudinal Weld Seams20-894 A & 8 9 1 .4 1 .79 1 . 1 599 1 06.0 -56 65.5 1 16 (Heat # 90099)

NOTES:

(a) From Table 7.1-1 ofWCAP-17506-NP[17J.

(b) This limiting RTprs value was calculated using the CF from the surveillance data and a full Ot. margin of 17"F, since this surveillance data is not credible.

PTLR for FNP Unit 1 Revision 6 Page 2 1 of 22 6.0 References 1 . WCAP-1 4040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

2. WCAP- 1 7 1 22-NP, Revision 0, "J. M. Farley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," A. E. Leicht and C. C. Heinecke, October 2009.

3. 1 0 CFR 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 1 9 , 1 995.

4. 1 0 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Federal Register, Volume 60, No. 243, December 1 9, 1 995.

5. ASTM E1 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," in ASTM Standards, Section 3, American Society for Testing and Materials, 1 982.

6. NRC Regulatory Guide 1 . 1 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 200 1 .

7. WCAP-1 6964-NP, Revision 0, "Analysis of Capsule Z from the Southern N uclear Operating Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program,"

J . M. Conermann and M. A. Hunter, October 2008.

8. Regulatory Guide 1 .99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

May 1 988.

9. WCAP-88 1 0, Revision 0, "Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program," J . A. Davidson, et. al. ,

December 1 976.

1 0. ASTM E 1 85-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, 1 973.

1 1 . WCAP-971 7, Revision 0, "Analysis of Capsule Y from the Alabama Power Company Farley Unit No. 1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, et. al., June 1 980.

1 2. WCAP-1 0474, Revision 0, "Analysis of Capsule U from the Alabama Power Company Joseph M . Farley Unit 1 Reactor Vessel Radiation Surveillance Program," R. S. Boggs, et. al. ,

February 1 984.

1 3. WCAP- 1 1 563, Revision 1 , "Analysis of Capsule X from the Alabama Power Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program," R. P. Shogan, et. al. ,

September 1 987.

1 4. WCAP- 1 4 1 96, Revision O,"Analysis of Capsule W from the Alabama Power Company Farley Unit 1 Reactor Vessel Radiation Surveillance Program," P. A. Peter, et. al., February 1 995.

1 5. WCAP-1 622 1 -NP, Revision 0, "Analysis of Capsule V from the Southern N uclear Operating Company, Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, K. G. Knight, et. al., March 2004.

1 6. WCAP-1 7365-NP, Revision 0, "Analysis of Capsule 284° from the Calvert Cliffs Unit No. 1 Reactor Vessel Radiation Surveillance Program," E. J. Long and J . I . Duo, March 201 1 .

1 7. WCAP-1 7506-NP, Revision 0, "Farley Units 1 and 2 Pressurized Thermal Shock Evaluations,"

B. A. Rosier, December 201 1 .

PTLR for FNP Unit 1 Revision 6 Page 22 of 22 1 8. CE NPSD-1 039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," Combustion Engineering Owners Group, June 1 997.

1 9. M itsubishi Heavy Industries , LTD, Kobe Shipyard & Machinery Works (MHI). M HI-SNC-01 95, Reactor Vessel Closure Head for Farley-1 , "Certified Material Test Report," August 22, 2003.

20. Westinghouse Letter ALA-09-1 1 6, "P-T Limit Curves with Margins for Instrumentation Errors and Extended Beltline Material Information," John M. Robinson, October 20, 2009.

2 1 . U . S . NRC Letter "Joseph M. Farley Nuclear Plant, Units 1 and 2, Issuance of Amendments Regarding Technical Specifications Revisions Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report (TAC Nos.

M E9256 and ME9257) (NL-1 2-0868)," ADAMS Accession # M L 1 3249A386, October 2, 2 0 1 3.

Joseph M. Farley Nuclear Plant Unit 1 Revision 6 Pressure Temperature Limits Report Unit 2 Revision 6 Pressure Temperature Limits Report Enclosure 2 Unit 2 PTLR, Revision 6

SOUTHERN COMPANY

..\

Energy to Se1'111! lOur WorlJ*

Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 2 Revision 6 November 2014

PTLR for FNP Unit 2 Revision 6 Page 1 of 22 Table of Contents List of Tables ......................................... .................. . ................... . . .................... . ............... 2 List of Figures . . . . . . . . . . . . .................... . . . ............................................. . ................. . . ................ 3 1.0 RCS Pressure Temperature Limits Report (PTLR) ........ . ....... . . ....... . . . ..... . . . . . . ..... . . .. 5 2.0 Operating Limits .......... . . . ............................................... . . . . ..................................... 5 2.1 RCS PressurefTemperature (PfT) Limits (LCO- 3.4.3) . . .................. . ............... 5 2 .2 RCP Operation Limits ............. . . . . . . . . . . . . . . ................. . .. . . . . . . . .............................. . . . . . 5 2.3 LTOP System Applicability Temperature (LCO - 3.4. 1 2) .................... ............... 5 3.0 Reactor Vessel Material Surveillance Program ................. . .

. .................. .............. 12 4.0 Reactor Vessel Surveillance Data Credibility . . . . . ........ . . . . . .................. ................... 13 5.0 Supplemental Data Tables ................................................................................... 13 6.0 References ......... ........................... .......... . .... . .............. . . . . . . . . ................. ..... . .......... 21

PTLR for FNP Unit 2 Revision 6 Page 2 of 22 Li st of Tables 2-1 Farley Unit 2 54 EFPY Heatup Curve Data Points . . . . . . . . . . . . . . . . . . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2-2 Farley Unit 2 54 EFPY Cooldown Curve Data Points . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 1 0 3-1 Surveillance Capsule Withdrawal Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . 1 2 5-1 Comparison of Surveillance Material 30 Ft-Lb Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1 .99, Revision 2, Predictions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 14 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data . . . . . . . . . .. . . . . . . . . . 1 5 5-3 Reactor Vessel Toughness Table (Unirradiated) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 6 5-4 Reactor Vessel Fluence Projections at 54 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 7 5-5 Summary of ART Values for the Reactor Vessel Materials at the 1 /4-T and 3/4-T Locations for 54 EFPY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 8 5-6 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Reactor Vessel Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 9 5-7 Pressurized Thermal Shock (RTPrs) Values for 54 EFPY . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 20

PTLR for FNP Unit 2 Revision 6 Page 3 of 22 List of Fi g ures 2-1 Farley Unit 2 Reactor Coolant System Heatup Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2-2 Farley Unit 2 Reactor Coolant System Cooldown Limitations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

PTLR for FNP Unit 2 Revision 6 Page 4 of 22 This page intentionally blank.

PTLR for FNP Unit 2 Revision 6 Page 5 of 22 1 .0 RCS Pressure Temperature Limits Report (PTLR)

This PTLR for Farley Nuclear Plant - Unit 2 has been prepared in accordance with the requirement of Technical Specification (TS) 5.6.6. Revisions to the PTLR shall be provided to the NRC after issuance.

This report affects TS 3.4.3, RCS PressurefTemperature (PfT) Limits. All TS requirements associated with low temperature overpressure protection (LTOP) are contained in TS 3.4 . 1 2, RCS Overpressure Protection Systems.

2.0 Operating Limits The limits for TS 3.4.3 are presented in the subsection which follows and were developed using the NRC-approved methodologies specified in TS 5.6.6. The methodologies are contained in WCAP-1 4040-A, Revision 4111 . The operability requirements associated with LTOP are specified in TS LCO 3.4. 1 2 and were determined to adequately protect the RCS against brittle fracture in the event of an LTOP transient. The limitation on the number of operating reactor coolant pumps (RCPs) is necessary to assure operation consistent with the pressure corrections incorporated in the PfT limits for flow losses associated with the RCPs.

2.1 RCS PressurefTemperature (PfT) Limits (LCO - 3 .4.3)

2. 1 .1 The minimum boltup temperature is 60°F.

2 . 1 .2 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 1 00°F in any one hour period.

b. A maximum cooldown of 1 00°F in any one hour period.

c. A maximum temperature change of Jess than or equal to 1 ooF in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2 . 1 .3 The RCS PfT limits for heatup and cooldown are specified by Figures 2-1 and 2-2, respectively.

2.2 RCP Operation Limits 2.2.1 The number of operating RCPs i s limited t o one a t RCS temperatures Jess than 1 1 0°F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

2.3 LTOP System Applicability Temperature (LCO- 3.4. 1 2) 2.3. 1 The Low Temperature Overpressure Protection (LTOP) System applicability temperature is 275°F.

The Farley Nuclear Plant - Unit 2 pressure-temperature limits for the reactor vessel inlet/outlet nozzle corner region and balance of reactor coolant system ferritic components were evaluated and found to be non-limiting with respect to the reactor vessel beltline pressure-temperature limits (Figures 2-1 and 2-2 below). This technical assessment was confirmed by the NRC via Safety Evaluation Reporti20J.

PTLR for FNP Unit 2 Revision 6 Page 6 of 22 Figure 2-1 Farley Unit 2 Reactor Coolant System Heatup Limitations121 (Heatup Rates up to 1 00°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi LlP at RCS temperatures ;::: 1 1 0°F and 27 psi .!lP at RCS temperatures < 1 1 0°F). Includes vessel flange requirements per 1 0 CFR 50, Appendix G.

Limiting Material: I ntermediate Shell Plate 8721 2-1 with credible surveillance data Limiting ART Values at 54 EFPY: 1 /4T = 200°F 3/4T = 1 65°F 2500 ---- ------ -- ------ ---- ----

-. .- -r. r------.--------.------.--r---1 I

2280 --r- *- -r----1-- -!--- - -

2000 - - -- ----1----l---L - - ... - --

Unacceptable Ope ra tio n I ---J --

1750 _ .

l-------1

---t-t-+---

/J I Acc:epta ble Operation J 1500

( 12so e J I -

0..

1 1000 u

750 --1-----+--- - -*-+---+ -

250 . -- ... -*--;:t::-.=-:;j**t:;-*-t-* Crtdcallly Lim H-CI on ln*ervk)e hydroatatlc teat t.mpereture (258*F) for ttl*

J :-:ntt;. l **Nio* eriod up to 54 1iFPY

.IC"'""I eo*,. 'I 0 --

• Y---

J l Th* Jower llmtt for Rcsl p rw**uro I* -14.7 P*'D J 1-- 1---1 o 50 100 180 200 250

- - *-- ... -*-* -*-*- ---1-.--1-. ...J---- ,......_._---+-- -t 300 Moderator Temperature (Deg. F) 350 400 4&0 800 SGO

PTLR for FNP U nit 2 Revision 6 Page 7 of 22 Figure 2-2 Farley Unit 2 Reactor Coolant System Cooldown Limitations121 (Cooldown Rates up to 1 00°F/hr) Applicable to 54 EFPY (adjusted to include 60 psi LlP at RCS temperatures :<:: 1 1 0°F and 27 psi LlP at RCS temperatures < 1 1 0°F). Includes vessel flange requirements per 1 0 CFR 50, Appendix G.

Limiting Material: I ntermediate Shell Plate 8721 2- 1 with credible surveillance data Limiting ART Values at 54 EFPY: 1 /4T = 200°F 3/4T = 1 65°F 500 I

I Unacceptable Operation 1

i500 . *- - --

I I Ac ceptable Operation l -*-- -

750 ---

- -- f.-*

k: - -AO;oady-tate -

Ratea, *F/Hr 1)8o""pl "

To"'P*

- eo*F i', *1 00 j

0 J:::i:

The -r lhn" lor FtC$

1- * -L;_I_ ... - . . .. 1- ..-.-..-h-.- - ....f-.r-.

0 SO 100 1 00 200 2110 300 380 400 450 500 550 Moderator 'l'lltmperature (Deg. F)

PTLR for FNP Unit 2 Revision 6 Page 8 of 22 Table 2-1 Farley Unit 2 54 EFPY Heatup Curve Data Pointsl21 (adjusted to include 60 psi .:1P at RCS temperatures 1 1 0°F and 27 psi .:1P at RCS temperatures < 11 0°F) 60°F/hr. 60°F/hr. 1 00°F/hr. 1 00°F/hr.

Leak Tes t Limit Heatu p Criticality Heatup C riticality T P {psig) o T { F) p T p T p T p

{o F) { psig) (o F) (psig) {o F} (psig } { o F) (psig }

238 2000 60 -14.7 256 -14.7 60 - 1 4.7 256 -1 4.7 256 2485 60 594 256 561 60 575 256 542 65 594 256 561 65 575 256 542 70 594 256 561 70 575 256 544 75 594 256 561 75 575 256 544 80 594 256 561 80 575 256 546 85 594 256 561 85 575 256 547 90 594 256 561 90 575 256 551 95 594 256 561 95 575 256 551 1 00 594 256 561 100 575 256 556 105 594 256 561 1 05 575 256 557 11 0 594 256 561 110 575 256 561 1 10 561 256 561 110 542 256 561 1 15 561 256 561 115 542 256 561 120 561 256 561 1 20 544 256 561 125 56 1 256 561 125 547 256 561 130 561 256 561 130 551 256 561 1 35 561 256 561 1 35 556 256 561 140 561 256 561 140 561 256 561 145 561 256 561 145 561 256 561 1 50 561 256 561 1 50 561 256 561 1 55 561 256 561 1 55 561 256 561 160 56 1 256 561 160 561 256 561 165 561 256 561 165 561 256 561 1 70 561 256 561 1 70 561 256 561 1 75 561 256 561 175 561 256 561 180 561 256 828 180 561 256 675 180 561 256 862 1 80 561 256 698 180 828 256 900 180 675 256 724 185 862 256 941 185 698 256 752 1 90 900 256 987 190 724 256 784

PTLR for FNP Unit 2 Revision 6 Page 9 of 22 Table 2-1 (contin ued )

Farley Unit 2 - 54 EFPY Heatup Curve Data Points!21 (adjusted to include 60 psi LlP at RCS temperatures 1 1 0°F and 27 psi LlP at RCS temperatures < 1 1 0°F)

Leak Test G0°F/hr. 1 00°F/hr.

G0°F/hr. Heatup 1 00°F/hr. Heatu p Limit Criticality Criticality p p

( oT T P (psig)

T P (ps ig)

T P (psig)

T F) (ps ig) ( o F) (o F) (o F) (o F) (psig) 1 95 941 256 1 038 1 95 752 256 81 9 200 979 256 1 094 200 784 256 858 205 1 021 256 1 1 30 205 81 9 256 910 210 1 067 260 1 1 75 210 858 260 949 21 5 1 1 19 265 1 237 21 5 901 265 1 00 1 220 1 1 75 270 1 29 1 220 949 270 1 05 9 225 1 237 275 1 350 225 1 001 275 1 1 23 230 1 29 1 280 1 41 6 230 1 059 280 1 1 94 235 1 350 285 1 488 235 1 1 23 285 1 272 240 1 41 6 290 1 568 240 1 1 94 290 1 359 245 1 488 295 1 656 245 1 272 295 1 454 250 1 568 300 1 753 250 1 359 300 1 559 255 1 656 305 1 861 255 1 454 305 1 67 5 260 1 753 31 0 1 979 260 1 559 31 0 1 803 265 1 861 315 21 1 0 265 1 675 315 1 944 270 1 979 320 2254 270 1 803 320 2099 275 21 1 0 325 241 4 275 1 944 325 2250 280 2254 280 2099 330 2399 285 241 4 285 2250 290 2399

PTLR for FNP Unit 2 Revision 6 Page 1 0 of 22 Table 2-2 Farley Unit 2 - 54 EFPY Cooldown Curve Data Points[2J (adjusted to include 60 psi ilP at RCS temperatures ;::: 11 0°F and 27 psi ilP at RCS temperatures < 1 1 0°F)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. 1 00°F/hr.

T(°F) P (psig) T(°F) P (psi g) T(°F) P (psig) T(°F) P (psig) T(°F) P (psig) 60 -14.7 60 -14.7 60 -14.7 60 -1 4.7 60 -14. 7 60 594 60 594 60 553 60 509 60 420 65 594 65 594 65 555 65 512 65 423 70 594 70 594 70 558 70 515 70 426 75 594 75 594 75 561 75 518 75 430 80 594 80 594 80 565 80 522 80 434 85 594 85 594 85 569 85 527 85 439 90 594 90 594 90 574 90 531 90 445 95 594 95 594 95 579 95 537 95 451 100 594 1 00 594 1 00 585 1 00 543 1 00 458 1 05 594 105 594 1 05 591 105 550 1 05 465 1 10 594 11 0 594 110 594 11 0 557 110 474 110 561 1 10 561 1 10 561 1 10 524 110 441 11 5 561 1 15 561 115 561 115 532 115 451 1 20 561 120 561 120 561 120 542 120 461 1 25 561 1 25 561 1 25 561 1 25 552 125 473 130 561 130 561 130 561 130 561 130 487 135 561 135 561 135 561 135 561 1 35 502 1 40 561 140 561 140 561 1 40 561 140 51 9 1 45 561 1 45 561 1 45 561 145 561 1 45 537 150 561 1 50 561 1 50 561 150 561 150 558 155 561 155 561 155 561 1 55 561 155 561 1 60 561 160 561 1 60 561 1 60 561 160 561 1 65 561 1 65 561 1 65 561 165 561 1 65 561 170 561 1 70 561 170 561 170 561 170 561 175 561 175 561 175 561 1 75 561 175 561 1 80 561 180 561 1 80 561 180 561 1 80 561 1 80 561 1 80 561 180 561 180 561 180 561 1 80 847 1 80 823 180 799 180 778 180 741 185 875 185 853 185 832 185 814 1 85 785 1 90 906 190 887 1 90 869 190 854 1 90 833 1 95 941 1 95 924 195 910 1 95 898 195 886 200 979 200 966 200 955 200 948 200 945 205 1021 205 1011 205 1005 205 1 002 205 1002 21 0 1067 21 0 1062 21 0 1060 21 0 1 060 210 1060 215 1119 215 1 1 18 215 1 118 215 1118 215 1 1 18

PTLR for FNP Unit 2 Revision 6 Page 11 of 22 Table 2-2 (continued)

Farley Unit 2 - 54 EFPY Cooldown Curve Data Pointsl21 (adjusted to include 60 psi L1P at RCS temperatures ;::: 11 ooF and 27 psi

.:\P at RCS temperatures < 110°F)

Steady State 20°F/hr. 40°F/hr. 60°F/hr. 1 00° F/hr.

T (°F ) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (ps ig) T (°F) P ( psig) 220 1 1 75 220 1 1 75 220 1 1 75 220 1 1 75 220 1 1 75 225 1 238 225 1 238 225 1 238 225 1 238 225 1 238 230 1 307 230 1 307 230 1 307 230 1 307 230 1307 235 1 384 235 1 384 235 1 384 235 1 384 235 1 384 240 1 468 240 1 468 240 1 468 240 1 468 240 1 468 245 1 562 245 1 562 245 1 562 245 1 562 245 1 562 250 1 665 250 1 665 250 1 665 250 1 665 250 1 665 255 1 779 255 1 779 255 1 779 255 1 779 255 1 779 260 1 906 260 1906 260 1 906 260 1 906 260 1 906 265 2045 265 2045 265 2045 265 2045 265 2045 270 21 99 270 2 1 99 270 2 1 99 270 21 99 270 2 1 99 275 2370 275 2370 275 2370 275 2370 275 2370

PTLR for FNP Unit 2 Revision 6 Page 1 2 of 22 3.0 Reactor Vessel Material Surveillance P rogram The reactor vessel material surveillance program is in compliance with 1 0 CFR 50, Appendix H, and is described in Section 5.4.3.6 of the Farley FSAR. Surveillance capsules are tested and the results reported in accordance with ASTM E1 85-82131.

The removal schedule is provided in Table 3-1 . Consistent with specific requirements for Farley Unit 2 associated with the granting of an exemption to Appendix H of 1 0 CFR 50 documented in NUREG-0 1 1 7 141 , Figures 2-1 and 2-2 are based on the greater, or limiting value, of the following: ( 1 ) the actual shift in reference temperature for plate 8721 2-1 as determined by impact testing, or (2) the predicted shift in reference temperature for weld seam 1 1 -923 as determined by Regulatory Guide 1 .99, Revision 2151 . The neutron transport and dosimetry evaluation methodologies used follow the guidance and meet the requirements of Regulatory Guide 1 . 1 90 , "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"161 . Results from the reactor vessel surveillance program will be used to update Figures 2-1 and 2-2 if the results indicate that the adjusted reference temperature (ART) for the limiting beltline material exceeds the ART used to generate the PIT limits shown in Figures 2-1 and 2-2 for the specified fluence period.

Table 3-1 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULE <a>

Capsule Capsule Location Lead Removal Fluence (c)

EFPY !bl 2

(Degree) Factor (n/cm )

u 343 3.26 1.11 6.05 X 10 18 w 110 2.84 3.96 1.73 X 10 19 X 287 3.38 6.43 2.98 X 10 19 z

(d) 340 2.98 1 3.85 4.92 X 1 0 19 y 290 3. 1 2 19.01 6.79 X 1019 (e) v 107 3.58 2 1 .82 8.73 X 10 19 (I) 171 Notes:

a) Data from Table 7-1, WCAP-16918-NP, Revision 1 b) Effective Full Power Years (EFPY) from plant startup.

c) Plant-specific evaluation.

d) This f luence is not less than once or greater than twice the peak EOL fluence for the initial 40-year license term.

e) This f luence is not less than once or greater than twice the peak EOL f luence for a 20-year license renewal term to 60 years.

f) This fluence is not less than once or greater than twice the peak EOL f luence for an additional 20-year license renewal term to 80 years.

PTLR for FNP Unit 2 Revision 6 Page 1 3 of 22 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1 .99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1 .99, Revision 2, describes the methodology for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in q uestion .

Per WCAP-8956[81 , the Unit 2 surveillance program was based o n ASTM E1 85-73[91 .

All six surveillance capsules (U [101 , W [1 1 1, X [1 21 , Z [1 31 , Y [141 , and V [71) have been removed from the Farley Unit 2 reactor vessel and analyzed. In accordance with the discussion of Regulatory Guide 1 .99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The credibility conclusions for the Farley Unit 2 surveillance plate and weld are described below:

The credibility evaluation of the Farley Unit 2 surveillance materials is documented in WCAP-1 691 8-NP, Revision 1 [71 . The credibility evaluation concluded that the surveillance data for Intermediate Shell Plate 872 1 2-1 is credible. The credibility evaluation concluded that the surveillance data for weld Heat # BOLA is non credible.

5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and u pper shelf energy decreases with Regulatory Guide 1 .99, Revision 2, predictions.

Table 5-2 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5-3 provides the unirradiated Farley Unit 2 reactor vessel toughness data.

Table 5-4 provides a summary of the reactor vessel fluence values at 54 EFPY.

Table 5-5 provides a summary of the ART values of the Farley Unit 2 reactor vessel materials at the 1 /4-T and 3/4-T locations for 54 EFPY.

Table 5-6 shows the calculation of the ART values at 54 EFPY for the limiting Farley Unit 2 reactor vessel material.

Table 5-7 provides RTPTs values for Farley Unit 2 for 54 EFPY.

PTLR for FNP Unit 2 Revision 6 Page 14 of 22 Table 5-1 Comparison of Surveillance Material 30 ft-lb Transition Temperature Shift and Upper Shelf Energy Decrease with Regulatory Guide 1 .99, Revision 2, Predictions <al 30 ft-lb Transition Upper Shelf Energy Temperature Shift Decrease Material Ca ps ule Fluence 181 Predicted Meas ured Predicted Meas u red (x 1 0 1 9 (o F) (b) ( o F) (c) (%) (b) (%) (d) n/cm 2, E >

1 .0 MeV)

I ntermediate Shell u 0.605 128.0 105.5 26 27 Plate 87212-1 w 1.73 171.5 167.7 33 22 (Longitudinal) X 2.98 192.1 1 64.8 37 26 z 4.92 208.5 200 . 1 42 28 y 6.79 217.2 214.2 45 36 v 8.73 222.9 218.3 48 34 I ntermediate Shell u 0.605 1 28.0 1 24.0 26 27 Plate 8721 2-1 w 1.73 171 .5 1 68.5 33 21 (Transverse) X 2.98 192.1 200 . 1 37 28 z 4.92 208.5 195.8 42 29 y 6.79 21 7.2 231 .0 45 42 v 8.73 222.9 215. 3 48 27 Surveillance u 0.605 32.8 -28.4 17 8 Program w 1.73 44.0 7.0 22 0 Weld Metal X 2.98 49.2 -1 5.6 24 0 z 4.92 53.4 1 0. 2 27 8 y 6.79 55.7 69. 1 30 5 v 8.73 57. 1 56.5 32 14 Heat Affected Zone u 0.605 --- 219.8 - - 30 w

Material 1 .73 --- 268.8 - 20 X

2.98 --- 230.5 - - 19 z

4.92 -- 263.8 -- 20 y

6.79 --- 269.6 - - 35 v

8.73 -- - 322.4 - - - 25 NOTES :

(a ) Data from Table 5-10, WCAP-16918-NP, Revision 1 [7]

(b ) Based on Reg. Guide 1.99, Rev. 2 methodology using the mean weight percent values of copper and nickel of the surveillance material.

(c) Calculated using measured Charpy data plotted using CVGRAPH, Version 5.3.

(d) Values are based on the definition of upper shelf energy given in ASTM E185-82.

(e) The fluence values presented here are the calculated values, not the best estimate values.

PTLR for FNP Unit 2 Revision 6 Page 1 5 of 22 Table 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data Ca ps ule f FF*ARTNor

{ 1 01 9 n/cm 2, FF lal ARTNoT FF2 Material Ca ps ule { o F) { o F)

E > 1 .0 MeV) u 0.605 0.859 1 05.5 90.7 0.738 I ntermediate Shell Plate 8721 2-1 w 1 .73 1 . 1 51 1 67.7 1 93.0 1 .324 (Longitudinal) X 2.98 1 .289 1 64.8 2 1 2.4 1 .662 z 4.92 1 .399 200. 1 280.0 1 .958 y 6.79 1 .458 21 4.2 3 1 2.3 2 . 1 25 v 8.73 1 .496 21 8.3 326.6 2.238 I ntermediate Shell u 0.605 0.859 1 24.0 1 06.5 0.738 Plate 8721 2-1 w 1 .73 1 .1 51 1 68.5 1 93.9 1 .324 (Transverse) X 2.98 1 .289 200 . 1 258.0 1 .662 z 4.92 1 . 399 1 95.8 274.0 1 . 958 y 6.79 1 .458 231 .0 336.8 2 . 1 25 v 8.73 1 .496 2 1 5.3 322 . 1 2.238 SUM: 2906 . 1 3 20.091 CF = :E(FF* RTNOT ) + l:(FF2) = 1 44.6 °F Weld Metal u 0.605 0.859 0.0 (c) 0.0 0.738 w 1 .73 1 . 1 51 7.0

( b) 8.1 1 .324 X 2.98 1 .289 0.0 (c) 0.0 1 .662 z

b 4.92 1 . 399 1 0.2 ( ) 1 4.3 1 .958 y 6.79 1 .458 69. 1 (b) 1 00.7 2 . 1 25 v 8.73 1 .496 56. 5 (b) 84.5 2.238 SUM: 207.59 1 0.046 CF = l:{FF* RTNOT ) + l:{FF2) = 20.7 °F 0 28 1 1og fl NOTES : (a) FF = Fiuence Factor = F ( * " 0

[15 1 (b) dRTNoT values from Table 4-1 were not multiplied by the ratio of 0.96 (from WCAP-14689, Rev. 6 Table 4, CFvesser + CFsurv weld = 36.8 + 38.2 = 0.96) to calculate the best fit chemistry factor (CF) as provided by Reg. Guide 1.99, Rev. 2, Position 2.1, since the ratio is less than one. This is a conservative approach.

(c) Actual measured dRT NOT values are less than zero. Since physically a reduction should not occur, a value of zero is conservatively used.

PTLR for FNP Unit 2 Revision 6 Page 1 6 of 22 Table 5-3 Reactor Vessel Toughness Table (Unirradiated) <a>

Beltline Material Cu Wei g ht Ni Wei g ht % IRTNDT (°F)

-- (d)

Closure Head Flange -

-60 Vessel Flange - -

- 60 Inlet Nozzle 8721 8-1 0.16 0.71 32 Inlet Nozzle 8721 8-2 0. 1 6 0.68 50 I nlet Nozzle 8721 8-3 0. 1 6 0.72 60 Outlet Nozzle 872 1 7-1 0.16 0.73 60 Outlet Nozzle 8721 7-2 0. 1 6 0.72 6 O utlet Nozzle 8721 7-3 0. 1 6 0.72 48 Upper Shell Forging 8721 6-1 0. 1 6 0.724 30 I ntermediate Shell Plate 87203-1 0.14 0.60 15 Intermediate Shell P late 8721 2-1 0.20 0.60 -1 0 Lower Shell Plate 8721 0-1 0. 1 3 0.56 18 Lower Shell Plate 8721 0-2 0.14 0.57 10 I nlet/Outlet Nozzle to Upper Shell Girth Seams (e) 0.07 1 .04 10 1 -926 A-+F Upper Shell to Intermediate Shell Circumferential

0. 1 53 0.077 -40 Weld Seam 1 0-923 (bl (Heat # 5P5622)

Upper Shell to Intermediate Shell Circumferential (1) 0.05 1 .0 -56 Weld Seam 1 0-923 (Heat # 51 922)

Upper Shell to Intermediate Shell Circumferential (1) 0.09 0.06 -56 Weld Seam 1 0-923 (Heat # 3P4767)

Intermediate Shell Longitudinal Weld Seam <0 0.027 0.947 -s6 1 9-923 A (bl (Heat # HODA)

I ntermediate Shell Longitudinal Weld Seam 0.027 0.9 1 3 -60 1 9-923 B (bl (Heat # BOLA)

Surveillance Weld (cl 0.028 0.89 -

Intermediate Shell to Lower Shell Circumferential

0. 1 53 0.077 -40 Weld Seam 1 1 -923 (bl (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams20-923 A & B (bl (Heat # 83640) 0.051 0.096 -70 NOTES:

I161 (a) F rom Table 4.2-1 ofWCAP- 1 7506-NP .

1 (b) Best-estimate copper and nickel f rom CE NPSD- 1039 17J.

(c) The best-estimate copper and n ickel value rep resents the average of two chemistry measurements performed 11 on the surveillance weld and documented inWCAP-8956 lgl andWCAP-1 1438 1 1

• The surveillance weld is representative of intermediate shell longitudinal weld 1 9-9238.

81 (d) Replacement closure head initial RT NOT value was taken from MH I-SNC-0455F211 *

(e) An estimation method using measured data was used to determine this initial RT NOT value. Therefore, a conservative value of 1 7"F is used for au and a, in margin calculations.

(f) These initial RT NOT values a re generic and taken from 10 CFR 50.61 paragraph (c)( 1 )(ii) of the 1 07 edition.

PTLR for FNP Unit 2 Revision 6 Page 1 7 of 22 Table 5-4 Reactor Vessel Fluence Projections at 54 EFPY <a>

19 2

( 1 0 n/cm , E > 1 .0 MeV)

Reactor Vess el Location Material 54 EFPY Neutron Fluence Inlet Nozzle 8721 8-1 0.0449 Inlet Nozzle 8721 8-2 0.0254 Inlet Nozzle 8721 8-3 0.01 86 Outlet Nozzle 872 1 7-1 0.01 26 Outlet Nozzle 8721 7-2 0.01 72 Outlet Nozzle 8721 7-3 0.0304 Upper Shell Forging 8721 6-1 1 . 09 87203-1 &

I ntermediate Shell Plates 5.76 8721 2-1 8721 0-1 &

Lower Shell Plates 5.75 8721 0-2 I nlet/Outlet Nozzle to Upper Shell 1 -926 A-+F 0.0449 Girth Seams Upper Shell to I ntermediate Shell 1 0-923 1 .09 Circumferential Weld Seam 1 9-923 A & 8 I ntermediate Shell Longitudinal 1 .83 Weld Seams I ntermediate Shell to Lower Shell 1 1 -923 5.75 Circumferential Weld Seam 20-923 A & 8 Lower Shell Longitudinal Weld 1 .83 Seams (a) From Table 5.2-1 ofWCAP- 1 7506-NP!161. These values are also summarized in Table 2-1 of Attachment B of NOTE:

1 ALA 1 1 61 91.

PTLR for FNP Unit 2 Revision 6 Page 1 8 of 22 Table 5-5 Summary of ART Values for the Reactor Vessel Materials at the 1 /4-T and 3/4-T Locations for 54 EFPY cal Material 1/4-T (0f) 3/4-T (0f)

I nlet Nozzle B721 8-1 83 60 I n let Nozzle B721 8-2 86 69 Inlet Nozzle B721 8-3 89 75 O utlet Nozzle B721 7-1 83 71 O utlet Nozzle B721 7-2 34 20 O utlet Nozzle 8721 7-3 88 69 Upper Shell Forging B721 6-1 1 72 141 I ntermediate Shell Plate B7203-1 1 82 1 58 I ntermediate Shell Plate B72 1 2-1 223 1 87 200(b) 165(b)

Intermediate Shell Plate B72 1 2-1 Using credible S/C Data Lower Shell Plate B721 0-1 1 72 1 50 Lower Shell Plate B721 0-2 1 75 1 52 I nlet/Outlet Nozzle to Upper Shell Girth Seams 69 57 1 -926 A-->F Upper Shell to Intermediate Shell Circumferential 82 55 Weld Seam 1 0-923 (Heat # 5P5622)

Upper Shell to I ntermediate Shell Circumferential 70 42 Weld Seam 1 0-923 (Heat # 5 1 922)

Upper Shell to I ntermediate Shell Circumferential 39 19 Weld Seam 1 0-923 (Heat # 3P4767)

I ntermediate Shell Longitudinal Weld Seam 33 17 1 9-923 A (Heat # HODA)

I ntermediate Shell Longitudinal Weld Seam 16 -3 1 9-923 B (Heat # BOLA)

Intermediate Shell Longitudinal Weld Seam 1 9-923 B (Heat # BOLA) -1 7 -28 Using non-credible S/C Data I ntermediate Shell to Lower Shell Circumferential 1 15 97 Weld Seam 1 1 -923 (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams20-923 A & B (Heat # 83640) 7 -1 2 NOTES:

(a) The ART values presented here are based on the reactor vessel surface f luence values summarized in Table 5-4.

121 The values for the beltline materials are from Tables 4-10 and 4-11 of WCAP-17123-NP, Revision 1 . The values for the extended beltline materials are summarized along with the values for the beltline materials in Tables 3-3 1191 and 3-4 of Attachment 8 of ALA-09-116 .

(b) Limiting 1/4-T and 3/4-T ART values. The PIT limit curves are based on these limiting ART values of 200°F and 165°F.

PTLR for FNP U nit 2 Revision 6 Page 1 9 of 22 Table 5-6 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Reactor Vessel Material <a>

Parameter I ntermediate Shell Plate 8721 2-1 Operating Period 54 EFPY Location -T  %-T Chemistry Factor, CF (°F) 1 44.6 1 44.6 (bl Fluence, f (1 0 1 9 n/cm2) 3 . 59 1 1 .396 Fluence Factor, FF 1 .3324 1 .0926 ilRTNOT = CF X FF (°F) 1 92 .7 1 58 . 1 Initial RTNoT. I (°F) -10 -1 0 Margin, M (°F) <c> 17 17 Adjusted Reference Temperature d 200( ) 1 65 (d)

(ART), (°F) per Regulatory Guide 1 .99, Revision 2 NOTES :

(a)

From Tables 4-10 and 4-11 using credible surveillance capsule data) of WCAP-17123-NP, Revision 1 r21.

2 (b) Fluence is based on fsurt (10 9 n/cm , E > 1.0 MeV) = 5.76. The Farley Unit 2 reactor vessel wall thickness is l + c1t,2) 0 5.

7.875 inches in the beltline region.

(c) Margin is calculated as M = 2( o The standard deviation for the initial RT NOT margin term, O'i, is O"F since the initial RT NOT is a measured value. The standard deviation for the RTNOT term, o6, is 1 7" F for the plate, except that o6 need not exceed 0.5 times the mean value of RT NOT* In accordance with Regulatory Guide 1.99, Revision 2, Position 2.1, values of o6 may be cut in half when based on credible surveillance data.

(d) Limiting Y.-T and -T ART values.

PTLR for FNP Unit 2 Revision 6 Page 20 of 22 Table 5-7 Pressurized Thermal Shock (RTpts) Values for 54 EFPY <aJ Surface Flue nee ARTNoT Material FF (CF x FF) ( I M RTPTS CF (101 9 n/cm 2 , (o oF (o F) ) ( oF) F)

E > 1 .0 MeV)

Inlet Nozzle B72 1 8-1 1 20.75 0.0449 0.2760 33.3 32 33.3 99 I nlet Nozzle B721 8-2 1 20 0.0254 0 . 1 990 23.9 50 23.9 98 Inlet Nozzle B721 8-3 121 0.01 86 0.1 644 1 9. 9 60 1 9. 9 1 00 O utlet Nozzle B72 1 7-1 1 2 1 .25 0.0126 0 . 1 280 1 5.5 60 1 5. 5 91 O utlet Nozzle B721 7-2 1 21 0.01 72 0.1 565 1 8. 9 6 1 8. 9 44 O utlet Nozzle B721 7-3 121 0. 0304 0.22 1 3 26.8 48 26.8 1 02 Upper Shell Forging B72 1 6-1 121 . 1 1 .09 1 . 0241 1 24.0 30 34.0 1 88 Intermediate Shell Plate B7203-1 1 00.0 5.76 1 .4292 1 42.9 15 34.0 1 92 Intermediate Shell Plate B721 2-1 1 49.0 5.76 1 .4292 21 3.0 -1 0 34.0 237 b) 2 1 4(

Intermediate Shell Plate B72 1 2-1 1 44.6 5.76 1 .4292 206.7 -10 1 7.0 Using credible S/C Data Lower Shell Plate B721 0-1 89.8 5.75 1 .4289 1 28.3 18 34.0 1 80 Lower Shell P late B721 0-2 98.7 5.75 1 .4289 1 41 . 0 10 34.0 1 85 Inlet/Outlet Nozzle to Upper Shell 95 0. 0449 0.2760 26.2 10 42. 9 79 Girth Seams 1 -926 A-+F Upper Shell to Intermediate Shell Circumferential Weld Seam 1 0-923 74. 1 1 . 09 1 .0241 75.9 -40 56.0 92 (Heat # 5P5622)

Upper Shell to Intermediate Shell Circumferential Weld Seam 1 0-923 68 1 .09 1 .0241 69.6 -56 65.5 79 (Heat # 51 922)

Upper Shell to Intermediate Shell Circumferential Weld Seam 1 0-923 46. 3 1 .09 1 . 0241 47.4 -56 58.3 50 (Heat # 3P4767)

I ntermediate Shell Longitudinal Weld

36. 8 1 . 83 1 . 1 657 42.9 -56 54.7 42 Seam 1 9-923 A (Heat # HODA)

Intermediate Shell Longitudinal Weld

36. 8 1 . 83 1 . 1 657 42.9 -60 42. 9 26 Seam 1 9-923 B (Heat # BOLA)

I ntermediate Shell Longitudinal Weld Seam 1 9-923 B (Heat # BOLA) 20.7 1 . 83 1 . 1 657 24. 1 -60 24. 1 -12 Using non-credible S/C Data I ntermediate Shell to Lower Shell Circumferential Weld Seam 1 1 -923 74.1 5.75 1 .4289 1 05.9 -40 56.0 1 22 (Heat # 5P5622)

Lower Shell Longitudinal Weld Seams20-923 A & B (Heat # 83640) 37.3 1 . 83 1 . 1 657 43.5 -70 43.5 17 NOTES :

I161 (a) From Table 7.2-1 ofWCAP- 1 7506-NP .

(b) This limiting RTPrs value was calculated using the CF from the surveillance data and a reduced o8 margin of 8.5*F, since this surveillance data is credible.

PTLR for FNP Unit 2 Revision 6 Page 2 1 of 22 6.0 References 1 . WCAP-1 4040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

2. WCAP-1 71 23-NP, Revision 1 , J. M. Farley U nit 2 Heatup and Cooldown Limit Curves for Normal Operation, July 201 1 .

3. ASTM E 1 85-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled N uclear Power Reactor Vessels," in ASTM Standards, Section 3, American Society for Testing and Materials, 1 982.

4. N UREG-0 1 1 7, Supplement 5 to the Safety Evaluation Report (NUREG-75/034) ,

Office of Nuclear Reactor Regulation, U . S. Nuclear Regulatory Commission in the matter of Alabama Power Company Joseph M . Farley N uclear Plant Unit 2, Docket No. 50-364., March 1 9, 1 981 .

5. N RC Regulatory Guide 1 .99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1 988.

6. N RC Regulatory Guide 1 . 1 90, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March, 200 1 .

7 . WCAP- 1 691 8-NP, Revision 1 , Analysis of Capsule V from the Southern N uclear Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, April 2008.

8. WCAP-8956, Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program, J. A. Davidson, et al. , August 1 977.

9. ASTM E1 85-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, American Society for Testing and Materials, 1 973.

1 0. WCAP-1 0425, Analysis of Capsule U from the Alabama Power Company Joseph M .

Farley Unit 2 Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et a l . ,

October 1 983.

1 1 . WCAP-1 1 438, Analysis of Capsule W from the Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program, S. E. Yanichko, et al., April 1 987.

1 2. WCAP-1 247 1 , Analysis of Capsule X from the Alabama Power Company Joseph M .

Farley Unit 2 Reactor Vessel Radiation Surveillance Program, E. Terek, et al.,

December 1 989.

1 3. WCAP-1 5 1 7 1 , Revision 1 , Analysis of Capsule Z from the Alabama Power Company Joseph M . Farley Unit 2 Reactor Vessel Radiation Surveillance Program, February 2000.

1 4. WCAP- 1 6351 -NP, Revision 1 , Analysis of Capsule Y from the Southern Nuclear O perating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Progra m , June 2008.

PTLR for FNP Unit 2 Revision 6 Page 22 of 22 1 5. WCAP-1 4689, Revision 6, Farley Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR S upport Documentation, T. J . Laubham, April 200 1 .

1 6. WCAP-1 7506-NP, Revision 0, Farley Units 1 and 2 Pressurized Thermal Shock Evaluations, B. A. Rosier, December 201 1 .

1 7. CE NPSD- 1 039, Revision 2, Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, Combustion Engineering Owners Group, June 1 997.

1 8. M itsubishi Heavy Industries, L TO, Kobe Shipyard & Machinery Works (MHI).

MHI-SNC-0455F2, Reactor Vessel Closure Head for Farley-2, Certified Material Test Report, June 2004.

1 9. Westinghouse Letter ALA-09-1 1 6, P-T Limit Curves with Margins for Instrumentation Errors and Extended Beltline Material Information, John M .

Robinson, October 20, 2009.

20. U.S. NRC Letter, "Joseph M . Farley Nuclear Plant, Units 1 and 2 , Issuance of Amendments Regarding Techn ical Specifications Revisions Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report (TAC Nos. ME9256 and ME9257) (NL-1 2-0868)."

ADAMS Accession # ML13249A386, October 2, 201 3.