L-2011-029, Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1

From kanterella
(Redirected from ML110700068)
Jump to navigation Jump to search
Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1
ML110700068
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 03/09/2011
From: Kiley M
Florida Power & Light Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2011-029
Download: ML110700068 (39)
v • d • e


Text

0 MAR- 92011 FPL. L-2011 -029 POWERING TODAY. 10 CFR 50.90 EMPOWERING TOMORROW.

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001 Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Response to NRC Request for Additional Information Regarding Extended Power Uprate License Amendment Request No. 205 and Reactor Materials Issues - Round 1

References:

(1) M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-113), "License Amendment Request No. 205: Extended Power Uprate (EPU)," (TAC Nos. ME4907 and ME4908), Accession No. ML103560169, October 21, 2010.

(2) Email from J. Paige (NRC) to S. Franzone (FPL), "EPU Acceptance Review Question Re:

Equivalent Margin Analysis," Accession No. ML103070063, November 1, 2010.

(3) M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-268), "Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU)

License Amendment Request (LAR) No. 205 and Equivalent Margin Analysis (EMA),"

November 12, 2010.

(4) M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-303), Supplemental Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU) License Amendment Request (LAR) No. 205 and Equivalent Margin Analysis (EMA), Accession No. ML103610321, December 21, 2010.

(5) Email from J. Paige (NRC) to T. Abbatiello (FPL), "Turkey Point EPU - Vessels and Internals Integrity (CVIB) Requests for Additional Information - Round 1", Accession No. ML110420241, February 11, 2011 By letter L-2010-113 dated October 21, 2010 [Reference 1], Florida Power and Light Company (FPL) requested to amend Renewed Facility Operating Licenses DPR-31 and DPR-41 and revise the Turkey Point Units 3 and 4 Technical Specifications (TS). The proposed amendment will increase each unit's licensed core power level from 2300 megawatts thermal (MWt) to 2644 MWt and revise the Renewed Facility Operating Licenses and TS to support operation at this increased core thermal power level. This represents an approximate increase of 15% and is therefore considered an extended power uprate (EPU).

By email from the U.S. Nuclear Regulatory Commission (NRC) Project Manager (PM) dated November 1, 2010 [Reference 2], additional information regarding the Equivalent Margin Analysis (EMA) was requested by the NRC staff in the Vessels and Internals Integrity (CVIB) to support their acceptance review of the EPU License Amendment Request (LAR) [Reference 1]. FPL provided its responses to the NRC request by letters L-2010-268 and L-2010-303 dated November 12, 2010 and December 21, 2010, respectively [References 3 and 4]. The responses included AREVA NP Inc proprietary copies of Turkey Point EMA Reconciliation Report and ANP-2312P, Rev 3, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Turkey Point Units 3 and 4 For Extended Life Through 48 Effective Full Power Years," January 2010.

(

an FPL Group company

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Page 2 of 2 By email from the NRC PM dated February 11, 2011 [Reference 5], additional information regarding reactor materials issues was requested by the NRC staff in CVIB to support their review of Reference 1. The Request for Additional Information (RAI) consisted of six (6) questions regarding reactor vessel materials issues. These six RAI questions and the applicable FPL responses are documented in the Attachment to this letter.

In accordance with 10 CFR 50.91(b)(1), a copy of this letter is being forwarded to the State Designee of Florida.

This submittal does not alter the significant hazards consideration or environmental assessment previously submitted by FPL letter L-2010-113 [Reference 1].

This submittal contains no new commitments and no revisions to existing commitments.

Should you have any questions regarding this submittal, please contact Mr. Robert J.

Tomonto, Licensing Manager, at (305) 246-7327.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on March *' , 2011.

Very truly yours, Michael Kiley Site Vice President Turkey Point Nuclear Plant Attachment cc: USNRC Regional Administrator, Region II USNRC Project Manager, Turkey Point Nuclear Plant USNRC Resident Inspector, Turkey Point Nuclear Plant Mr. W. A. Passetti, Florida Department of Health

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 1 of 37 Turkey Point Units 3 and 4 RESPONSE TO NRC RAI REGARDING EPU LAR NO. 205 AND CVIB REACTOR MATERIALS ISSUES - ROUND 1 ATTACHMENT

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 2 of 37 Response to Request for Additional Information The following information is provided by Florida Power and Light Company (FPL) in response to the U. S. Nuclear Regulatory Commission's (NRC) Request for Additional Information (RAI).

This information was requested to support License Amendment Request (LAR) 205, Extended Power Uprate (EPU), for Turkey Point Nuclear Plant (PTN) Units 3 and 4 that was submitted to the NRC by FPL via letter (L-2010-113) dated October 21, 2010 [Reference 1].

In an email dated November 1, 2010 [Reference 2], additional information regarding the PTN Equivalent Margin Analysis (EMA) was requested by the NRC's Vessels and Internals Integrity Branch (CVIB) to support their acceptance review of the EPU LAR. FPL provided responses to the NRC request by letters L-2010-268 and L-2010-303 dated November 12, 2010 and December 21, 2010, respectively [References 3 and 4]. The responses included AREVA NP Inc proprietary copies of Turkey Point EMA Reconciliation Report and ANP-2312P, Rev 3, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Turkey Point Units 3 and 4 for Extended Life Through 48 Effective Full Power Years," January 2010 [References 5 and 6].

In an email dated February 11, 2011 [Reference 7], the NRC staff requested additional information regarding FPL's request to implement the Extended Power Uprate. The RAI consisted of six (6) questions from CVIB regarding reactor materials issues. These six RAI questions and the applicable FPL responses are documented below.

CVIB-1.1 The revised surveillance capsule withdrawal schedule for Turkey Point, Units 3 and 4 allows the last capsule, X4, to be withdrawn between 31.4 and 47.8 effective full power years (EFPY). This schedule does meet the recommendation of American Society for Testing and Materials (ASTM) E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," that for a reactor with five surveillance capsules installed, the last capsule should be withdrawn at a fluence greater than once but less than-t*ice the peak end-of-life (EOL) vessel fluence. However, the staff requests the licensee provide a single estimated EFPY value at which the capsule will be withdrawn rather than a range, or commit to providing this value later.

Surveillance Capsule X4 will be withdrawn when it reaches a fluence that is approximately equivalent to the 80-year (67 EFPY) peak reactor vessel fluence of 8.14 x 1019 n/cm2 (E >1.0MeV). Therefore, accounting for EPU conditions, Capsule X 4 will be withdrawn at the vessel refueling date that is nearest to 35.8 EFPY. This withdrawal date of 35.8 EFPY will be specified in the Turkey Point Units 3 and 4 surveillance capsule withdrawal schedule. It should be noted that the withdrawal fluence is consistent with the fluence and the intent of the "Coordinated U. S. PWR Reactor Vessel Surveillance Program." However, the withdrawal EFPY recommended in it differs from the withdrawal EFPY listed above because the report did not consider the effects of EPU.

CVIB-1.2 The revised equivalent margins analysis (EMA) forwarded by letter dated December 21, 2010 (Reference 1), stated that the low-upper shelf fracture mechanics evaluation is performed according to the acceptance criteria and evaluation procedures contained in Appendix K to Section XI of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code),

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 3 of 37 and references the ASME Code,Section XI, 1998 Edition through 2000 Addenda. Title 10 of Code of Federal Regulations (10 CFR) Part 50, Appendix G, IV.A.I.a, requires that such analyses use the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a(b)(2) at the time the analysis is submitted. The latest edition of the ASME Code,Section XI (Division 1) incorporated by reference into 10 CFR 50.55a at the time of the submittal is the 2004 edition. The staff therefore requests the licensee reconcile the differences between the 1998 through 2000 Addenda, and 2004 editions of the ASME Code,Section XI, specifically as the differences affect the low-upper shelf toughness evaluation.

With respect to a low upper-shelf toughness evaluation of reactor vessel steels,,

there are minor differences between the 1998 Edition through 2000 Addendum and the 2004 Edition of the ASME Code. The two areas of the ASME Code which affect the low-upper shelf toughness evaluation performed for the Turkey Point vessels are the material properties in Section II, Part D, and the acceptance criteria and evaluation procedures of Section XI, Appendix K.

The material properties obtained from the ASME Code for use in the Turkey Point low-upper shelf toughness evaluation are listed in the following tables for the two versions of the Code.

1998 Edition through 2000 Addendum Base Metal Cladding Material Material Young' s Coeff.

T lof ASME Yed Young' s Coeff.

Thermal of Temp Modulus Thermal Yield Modulus Tem.a Expansion Strength Expansion (F) (ksi) (in/in/F) (ksi) (ksi) (in/in/F) 70 27800 6.40E-06 50.0 28300 8.50E-06 200 27100 6.70E-06 47.0 2571 8.90E-06 300 26700 6.90E-06 45.5 27000 9.20E-06 400 226100, 7.1OE-06 44.2 265j,0 9.50E-06 500 25700 7.30E-06 43.2 8 9.70E-06 600 250Q I 7.40E-06 42.1 25300 9.80E-06 2004 Edition Base Metal Cladding Material Material Young' s Coeff.

Teml of ASME Yld Young' s Coeff.

Thra of Temp Modulus Thermal Yield Modulus Thermal Expansion Strength Expansion (F) (ksi) (in/in/F) (ksi) (ksi) (in/in/F) 70 27800 6.40E-06 50.0 28300 8.50E-06 200 27100 6.70E-06 47.0 ý-75,00 8.90E-06 300 26700 6.90E-06 45.5 27000 9.20E-06 400 MI 0Q 7.1OE-06 44.2 r2640 9.50E-06 500 25700 7.30E-06 43.2 r 9.70E-06 600 5 10 , 7.40E-06 42.1 25300 9.80E-06

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 4 of 37 The 2004 changes to Code material properties, highlighted in gray, are less than 0.5% and only affect the Young's modulus. Using the values in the 2004 Edition would not significantly affect the results of the Turkey Point low-upper shelf toughness evaluation.

Regarding Appendix K to Section XI, the only difference between the two versions of the Code is the addition of SI Units in the 2004 Edition. This change would have no effect on the results of the Turkey Point low-upper shelf toughness evaluation.

CVIB-1.3 In Section 7 of Reference 1, the licensee indicated that the applied J-integral was calculated using the following equation:

Japplied(a) = 1000K 2 1total(a)(1-v 2)/E This is essentially the same as the ASME Code,Section XI K-5210 equation:

2 J = 1000(K',) /E' where:

E' = E/(1-v2)

K'i = stress intensity factor adjusted for small scale yielding.

Article K-5000, subparagraph K-5210 of the ASME Code,Section XI, Appendix K (2004 Edition), provides an adjustment of the effective flaw depth for small-scale yielding as follows:

2 ae = a + [1/(67r)](Kl/y) where:

a actual flaw depth, ae = effective flaw depth, K, = applied stress intensity, ry- yield strength.

Paragraph K-5210 further states that the stress intensity factor for small scale yielding, K' 1, shall be calculated by substituting ae for a.

The licensee did not discuss whether the effects of small scale yielding were included in the Kitotal term. The staff therefore requests that the licensee discuss how the effects of small scale yielding were accounted for in the KItotai term.

Small-scale yielding is addressed in Appendix K to Section XI through use of a plastic zone correction to the postulated flaw depth, such that the effective flaw depth is expressed as a, = a + [ 1/(67r)](Kl/Uy) 2 Equation [1]

This effective flaw depth is explicitly cited in Section 4 of ANP-2312P [Reference 6]

for the prescriptive Appendix K flaw evaluation procedure for Levels A and B Service Loadings. Article K-5210 of Appendix K presents an overall procedure for calculating applied J-integrals for Levels C and D Service Loadings. The evaluation for Levels C and D Service Loadings requires plant specific transient analysis to determine pressure and thermal loads and stress intensity factors as a function of time. The PCRIT computer code used by AREVA to determine time-varying stress

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 5 of 37 intensity factors has a built-in feature to calculate the effective flaw depth described by Equation [1]. This option of the code was used in the Turkey Point low-upper shelf toughness evaluation for Levels C and D Service Loadings.

CVIB-1.4 Provide the basis, such as a report or calculation, for the pressure-temperature (P-T) limits for Turkey Point, Units 3 and 4 that are given in proposed revised Technical Specification Figures 3.4-2 and 3.4-3. If the report or calculation does not contain the following items, then the following items should be provided separately:

a. Provide a tabulation of the thermal stress intensity factors (Kit) used to generate the heatup and cooldown curves for each coolant temperature for heatup and cooldown.

Per the agreement reached during the telephone conference call on February 3, 2011 involving the NRC, FPL, and Westinghouse, the thermal stress intensity factors are being provided for only the most limiting heatup and cooldown rates (100°F/hr). The K1, values for the 100°F/hr heatup rate are presented in Table 1.

The K1 t values for the I00°F/hr cooldown rate are presented in Table 2.

Table 1 Kit Values for 100lF/hr Heatup Curve for 48 EFPY Water 1/4T Thermal 3/4T Thermal Temp. Stress Stress (OF) Intensity Factor Intensity Factor (ksi~in.) (ksi~in.)

70 -0.9847 0.4966 75 -2.3616 1.4564 80 -3.4847 2.3644 85 -4.5262 3.1774 90 -5.3926 3.8779 95 -6.1705 4.4891 100 -6.8263 5.0171 105 -7.4130 5.4784 110 -7.9115 5.8789 115 -8.3582 6.2297 120 -8.7400 6.5358 125 -9.0835 6.8051 130 -9.3790 7.0415 135 -9.6465 7.2507 140 -9.8782 7.4356 145 -10.0895 7.6004 150 -10.2739 7.7473 155 -10.4437 7.8792 160 -10.5931 7.9979 165 -10.7322 8.1055 170 -10.8556 8.2032

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 6 of 37 Water 1/4T Thermal 3/4T Thermal Temp. Stress Stress (OF) Intensity Factor Intensity Factor (ksi\in.) (ksihin.)

175 -10.9718 8.2929 180 -11.0761 8.3751 185 -11.1754 8.4513 190 -11.2654 8.5221 195 -11.3520 8.5884 200 -11.4315 8.6505 205 -11.5088 8.7094 210 -11.5803 8.7653 215 -11.6507 8.8187 220 -11.7164 8.8698 225 -11.7816 8.9191 230 -11.8430 8.9667 235 -11.9043 9.0130 240 -11.9626 9.0581 245 -12.0210 9.1021 250 -12.0769 9.1452 255 -12.1332 9.1876 260 -12.1874 9.2293 265 -12.2422 9.2705 270 -12.2952 9.3112 275 -12.3488 9.3515 280 -12.4009 9.3914 285 -12.4537 9.4311 290 -12.5052 9.4705 295 -12.5575. 9.5097 300 -12.6085 9.5487 305 -12.6604 9.5877 310 -12.7112 9.6265 315 -12.7628 9.6652 320 -12.8134 9.7039 325 -12.8649 9.7425 330 -12.9155 9.7811 335 -12.9668 9.8196 340 -13.0174 9.8582 345 -13.0688 9.8967 350 -13.1194 9.9353 355 -13.1708 9.9739 360 -13.2215 10.0125 365 -13.2730 10.0512 370 -13.3239 10.0899

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 7 of 37 Water 1/4T Thermal 3/4T Thermal Temp. Stress Stress (OF) Intensity Factor Intensity Factor (ksi\in.) (ksiin.)

375 -13.3754 10.1286 380 -13.4264 10.1674 385 -13.4781 10.2062 390 -13.5292 10.2451 395 -13.5810 10.2840 400 -13.6323 10.3230 405 -13.6843 10.3621 410 -13.7358 10.4012 415 -13.7879 10.4404 420 -13.8396 10.4796 425 -13.8918 10.5189 430 -13.9437 10.5583 435 -13.9961 10.5977 440 -14.0482 10.6373 445 -14.1008 10.6768 450 -14.1531 10.7165 455 -14.2059 10.7562 460 -14.2584 10.7960 465 -14.3114 10.8359 470 -14.3640 10.8758 475 -14.4172 10.9159 480 -14.4701 10.9559 485 -14.5234 10.9961 490 -14.5765 11.0363 495 -14.6301 11.0767 500 -14.6833 11.1170 505 -14.7371 11.1575 510 -14.7906 11.1980 515 -14.8445 11.2387 520 -14.8982 11.2793 525 -14.9524 11.3201 530 -15.0063 11.3609 535 -15.0606 11.4019 540 -15.1147 11.4428 545 -15.1692 11.4839 550 -15.2236 11.5250

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 8 of 37 Table 2 Kjt Values for 100°F/hr Cooldown Curve for 48 EFPY Water 1/4T Thermal Stress Temp. Intensity Factor (OF) (ksi'in.)

545 0.9598 540 2.4111 535 3.7032 530 4.9429 525 6.0336 520 7.0381 515 7.9218 510 8.7274 505 9.4366 500 10.0798 495 10.6456 490 11.1567 485 11.6052 480 12.0089 475 12.3616 470 12.6778 465 12.9524 460 13.1973 455 13.4083 450 13.5953 445 13.7546 440 13.8945 435 14.0118 430 14.1135 425 14.1966 420 14.2673 415 14.3228 410 14.3684 405 14.4015 400 14.4269 395 14.4419 390 14.4509 385 14.4514 380 14.4472 375 14.4360

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 9 of 37 Water 1/4T Thermal Stress Temp. Intensity Factor (OF) (ksi\Iin.)

370 14.4212 365 14.4007 360 14.3774 355 14.3493 350 14.3193 345 14.2852 340 14.2498 335 14.2110 330 14.1712 325 14.1287 320 14.0855 315 14.0401 310 13.9943 305 13.9465 300 13.8986 295 13.8490 290 13.7995 285 13.7486 280 13.6979 275 13.6459 270 13.5943 265 13.5416 260 13.4892 255 13.4360 250 13.3831 245 13.3295 240 13.2764 235 13.2225 230 13.1691 225 13.1151 220 13.0616 215 13.0075 210 12.9540 205 12.9000 200 12.8465 195 12.7925 190 12.7391 185 12.6852

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 10 of 37 Water 1/4T Thermal Stress Temp. Intensity Factor (OF) (ksi'in.)

180 12.6319 175 12.5782 170 12.5250 165 12.4715 160 12.4185 155 12.3651 150 12.3123 145 12.2591 140 12.2065 135 12.1536 130 12.1012 125 12.0485 120 11.9963 115 11.9438 110 11.8918 105 11.8396 100 11.7879 95 11.7359 90 11.6844 85 11.6326 80 11.5814 75 11.5298 70 11.4788

b. Provide a tabulation or graph of the temperature differential from the coolant to the crack tip used to generate the P-T limits, and describe the methodology used to determine this differential, unless Figure G-2214-1 and Figure G-2214-2 of the ASME Code,Section XI, Appendix G, were used to determine the temperature differential.

Per the agreement reached during the telephone conference call on February 3, 2011 involving the NRC, FPL, and Westinghouse, the coolant and crack tip temperatures will be provided only for the most limiting heatup and cooldown rates (1 00°F/hr). The temperature values for the I 00°F/hr heatup rate are presented in Table 3. The temperature values for the 100°F/hr cooldown rate are presented in Table 4.

Regarding the methodology used in calculating temperature differential, the temperatures are calculated using the one-dimensional transient heat conduction equation that is contained in Section 2.6.1 of WCAP-14040-A, Revision 4

[Reference 8]. A through-wall temperature distribution was calculated for each

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 11 of 37 time step during each cooldown or heatup ramp of interest. These methods are incorporated into the OPERLIM computer code.

Table 3 Temperature Values for 100°F/hr Heatup Curve for 48 EFPY Water 1/4T Crack Tip 3/4T Crack Tip Temperature Temperature Temperature (OF) (OF) (OF) 70 66.156 65.070 75 68.992 65.451 80 72.216 66.385 85 75.670 67.874 90 79.389 69.826 95 83.228 72.190 100 87.263 74.904 105 91.390 77.918 110 95.655 81.191 115 99.995 84.686 120 104.432 88.373 125 108.927 92.223 130 113.491 96.214 135 118.101 100.326 140 122.760 104.543 145 127.455 108.849 150 132.183 113.232 155 136.940 117.681 160 141.721 122.187 165 146.524 126.742 170 151.344 131.340 175 156.181 135.974 180 161.029 140.639 185 165.891 145.331 190 170.761 150.046 195 175.642 154.782 200 180.528 159.534 205 185.422 164.302 210 190.320 169.082 215 195.224 173.873 220 200.131 178.674 225 205.043 183.484 230 209.957 188.300 235 214.874 193.122 240 219.793 197.950

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 12 of 37 Water 1/4T Crack Tip 3/4T Crack Tip Temperature Temperature Temperature (OF) (OF) (OF) 245 224.714 202.782 250 229.637 207.619 255 234.561 212.458 260 239.486 217.300 265 244.413 222.145 270 249.340 226.992 275 254.268 231.841 280 259.197 236.692 285 264.126 241.543 290 269.056 246.396 295 273.986 251.250 300 278.916 256.104 305 283.847 260.959 310 288.778 265.815 315 293.709 270.671 320 298.640 275.528 325 303.572 280.384 330 308.503 285.241 335 313.435 290.098 340 318.366 294.956 345 323.298 299.813 350 328.229 304.670 355 333.161 309.528 360 338.092 314.385 365 343.023 319.242 370 347.955 324.099 375 352.886 328.956 380 357.817 333.813 385 362.749 338.670 390 367.680 343.526 395 372.611 348.383 400 377.542 353.239 405 382.472 358.095 410 387.403 362.951 415 392.334 367.806 420 397.264 372.662 425 402.195 377.517 430 407.125 382.372 435 412.055 387.227 440 416.985 392.081 445 421.915 396.935

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 13 of 37 Water 1/4T Crack Tip 3/4T Crack Tip Temperature Temperature Temperature (OF) (OF) (OF) 450 426.845 401.789 455 431.775 406.643 460 436.704 411.497 465 441.634 416.350 470 446.563 421.203 475 451.492 426.056 480 456.422 430.909 485 461.351 435.761 490 466.279 440.613 495 471.208 445.465 500 476.137 450.316 505 481.065 455.168 510 485.994 460.019 515 490.922 464.869 520 495.850 469.720 525 500.778 474.570 530 505.706 479.420 535 510.633 484.270 540 515.561 489.119 545 520.489 493.968 550 525.416 498.817

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 14 of 37 Table 4 Temperature Values for 100IF/hr Cooldown Curve for 48 EFPY Water 1/4T Crack Tip Temperature Temperature (OF) (OF) 545 549.091 540 546.642 535 543.653 530 540.452 525 536.970 520 533.372 515 529.588 510 525.700 505 521.678 500 517.562 495 513.345 490 509.045 485 504.668 480 500.220 475 495.712 470 491.144 465 486.529 460 481.864 455 477.161 450 472.417 445 467.643 440 462.836 435 458.003 430 453.144 425 448.264 420 443.363 415 438.445 410 433.510 405 428.561 400 423.598 395 418.624 390 413.639 385 408.645 380 403.642 375 398.631 370 393.613 365 388.590 360 383.561

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 15 of 37 Water 1/4T Crack Tip Temperature Temperature (OF) (OF) 355 378.527 350 373.488 345 368.446 340 363.400 335 358.350 330 353.298 325 348.244 320 343.187 315 338.129 310 333.068 305 328.006 300 322.943 295 317.879 290 312.813 285 307.746 280 302.679 275 297.611 270 292.543 265 287.473 260 282.404 255 277.334 250 272.263 245 267.193 240 262.122 235 257.051 230 251.979 225 246.908 220 241.837 215 236.765 210 231.693 205 226.622 200 221.550 195 216.478 190 211.407 185 206.335 180 201.264 175 196.192 170 191.121 165 186.050 160 180.979 155 175.907

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 16 of 37 Water 1/4T Crack Tip Temperature Temperature (OF) (OF) 150 170.836 145 165.765 140 160.695 135 155.624 130 150.553 125 145.483 120 140.413 115 135.342 110 130.272 105 125.202 100 120.132 95 115.062 90 109.993 85 104.923 80 99.854 75 94.785 70 89.716

c. Provide the numerical temperature versus pressure values corresponding to the heatup and cooldown curves, and the hydrotest curve, in Technical Specification Figures 3.4-2 and 3.4-3.

The numerical temperature versus pressure values for the heatup curves and the hydrotest curve are presented in Table 5. The numerical temperature versus pressure values for the cooldown curves are presented in Table 6.

Table 5 Data Points for Heatup P-T Limit Curves Applicable to 48 EFPY with Flange, without Temperature and Pressure Uncertainties, and Using Combined Methodology(a)

Leak Test 6 0 1F/hr 60'F/hr 100 0 F/hr 100IF/hr Limit Heatup Criticality Heatup Criticality T P T P T P T P T P (2F) (psig) (7F) (psig) (0F) (psig) (7F) (psig0 (F) (psig) 238 2000 70 0 262 0 70 0 262 0 238 2000 70 587 262 621 70 552 262 621 262 2485 75 587 262 621 75 552 262 621 262 2485 80 587 262 621 80 552 262 621 85 587 262 621 85 552 262 621 90 587 262 621 90 552 262 621 95 587 262 621 95 552 262 621 100 588 262 621 100 552 262 621 105 591 262 621 105 552 262 621

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 17 of 37 Leak Test 60'F/hr 60 0 F/hr 1001F/hr 100OFahr Limit Heatup Criticality Heatup Criticality T P T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (F) (psig) (OF) (psig) 110 596 262 621 110 552 262 621 115 603 262 621 115 552 262 621 120 612 262 1018 120 553 262 920 120 612 262 1023 120 553 262 921 120 612 262 1029 120 553 262 921 125 621 262 1037 125 557 262 923 130 632 262 1046 130 561 265 946 135 645 262 1056 135 567 270 988 140 658 262 1067 140 575 275 1033 145 673 262 1079 145 583 280 1081 150 690 262 1093 150 594 285 1133 155 708 262 1108 155 605 290 1188 160 727 262 1124 160 618 295 1248 165 748 262 1141 165 632 300 1313 170 770 262 1137 170 648 305 1382 175 795 265 1167 175 666 310 1456 180 821 270 1222 180 685 315 1536 185 849 275 1280 185 705 320 1621 190 880 280 1343 190 728 325 1713 195 912 285 1410 195 752 330 1801 200 947 290 1483 200 779 335 1880 205 985 295 1560 205 808 340 1964 210 1026 300 1644 210 838 345 2055 215 1070 305 1734 215 872 350 2153 220 1117 310 1830 220 908 355 2258 225 1167 315 1934 225 946 360 2371 230 1222 320 2045 230 988 235 1280 325 215.9 235 1033 240 1343 330 2255 240 1081 245 1410 335 2348 245 1133 250 1483 340 2443 250 1188 255 1560 255 1248 260 1644 260 1313 265 1734 265 1382 270 1830 270 1456 275 1934 275 1536 280 2045 280 1621 285 2159 285 1713 290 2255 290 1801 295 2348 295 1880

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 18 of 37 Leak Test 60'F/hr 6 0 °F/hr 100'F/hr 100°F/hr Limit Heatup Criticality Heatup Criticality T P T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 300 2443 300 1964 305 2055 310 2153 315 2258 320 23711 (a) Pressure values in italics resulted from the use of circumferential flaw methodology. All other pressure values resulted from the use of axial flaw methodology.

Table 6 Data Points for Cooldown P-T Limit Curves Applicable to 48 EFPY with Flange, without Temperature and Pressure Uncertainties, and Using Combined Methodology(a)

Steady State 2 0 °F/hr. 4 0 °F/hr. 60°F/hr. 100°F/hr.

T P T P T P T P T P (OF) (psgO°)

pi)(F) (psig) (OF) (psig) (OF) (psig) 70 0 70 0 70 0 70 0 70 0 70 602 70 565 70 527 70 488 70 409 75 609 75 572 75 534 75 495 75 417 80 -616 80 579 80 541 80 503 80 425 85 621 85 586 85 549 85 511 85 434 90 621 90 595 90 558 90 520 90 444 95 621 95 604 95 567 95 530 95 455 100 621 100 613 100 577 100 541 100 466 105 621 105 621 105 588 105 552 105 479 110 621 110 621 110 600 110 564 110 493 115 621 115 621 115 612 115 577 115 507 120 621 120 621 120 621 120 592 120 523 120 621 120 621 120 626 125 607 125 540 120 694 120 660 125 640 130 624 130 559 125 707 125 674 130 656 135 641 135 579 130 721 130 689 135 673 140 661 140 600 135 737 135 705 140 691 145 682 145 624 140 753 140 722 145 711 150 704 150 649 145 771 145 741 150 732 155 728 155 676 150 790 150 761 155 755 160 754 160 705 155 810 155 783 160 780 165 782 165 737 160 833 160 806 165 806 170 812 170 771

Turkey Point Units 3 and 4 L-201 1-029 Docket Nos. 50-250 and 50-251 Attachment Page 19 of 37 Stead, State 20°F/hr. 4 0 IF/hr. 60IF/hr. 100°F/hr.

T P T P T P T P T P (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 165 856 165 831 170 835 175 845 175 808 170 882 170 858 175 865 180 880 180 847 175 909 175 887 180 898 185 918 185 890 180 939 180 918 185 934 190 958 190 936 185 970 185 952 190 972 195 1002 195 970 190 1005 190 988 195 1014 200 1049 200 999 195 1041 195 1027 200 1058 205 1100 205 1030 200 1081 200 1068 205 1106 210 1155 210 1064 205 1123 205 1113 210 1157 215 1212 215 1100 210 1169 210 1162 215 1212 220 1270 220 1140 215 1218 215 1214 220 1270 225 1324 225 1182 220 1270 220 1270 225 1327 230 1365 230 1228 225 1327. 225 1327 230 1388 235 1409 235 1278 230 1388 230 1388 235 1453 240 1456 240 1332 235 1453 235 1453 240 1520 245 1507 245 1389 240 1524 240 1524 245 1568 250 1562 250 1452 245 1599 245 1599 250 1620 255 1621 255 1519 250 1681 250 1680 255 1676 260 1685 260 1592 255 1768 255 1732 260 1736 265 1754 265 1670 260 1843 260 1789 265 1801 270 1828 270 1755 265 1901 265 1850 270 1870 275 1908 275 1846 270 1962 270 1915 275 1945 280 1994 280 1944 275 2029 2175 198-5. 280 2026 285 2087 285 2050 280 2100 280 2061 285 2113 290 2186 290 2164 285 2177 285 2143 290 2206 295 2294. 295 2287 290 2259 290 2230 295 2307 300 2409 300 2409 295 2348 295 2325 300 2415 300 2443 300 2426 305 2545 (a) Pressure values in italics resulted from the use of circumferential flaw methodology. All other pressure values resulted from the use of axial flaw methodology.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 20 of 37

d. The P-T curves provided in EPU Licensing Report Figures 2.1.2-1 and 2.1.2-2 and TS Figures 3.4-2 and 3.4-3 do not indicate whether there is any pressure difference between the reactor vessel (RV) pressure and pressure at the measurement location. If such a pressure difference exists, provide the correction factors used to correct between the actual reactor vessel (RV) pressure and the indicated pressure at the measurement location.

The P-T limit curves do not include margin for the pressure difference between the RV pressure and the pressure at the measurement location. The PTN Overpressure Mitigation System (OMS) power-operated relief valve (PORV) setpoint, which prevents the P-T limits from being exceeded, does however account for a pressure differential of 57.4 psi (see Licensing Report (LR)

Section 2.8.4.3.2.3) between the pressure measurement location and the RV.

Since the OMS setpoint includes the impact of the pressure differential, it is not necessary to include this impact in the P-T limit curves.

e. In the technical specification (TS) bases markups provided with the EPU application, the licensee provided a revised Table B 3/4.4-1 that shows the closure flange RTNDT has been changed from 44 IF to -50 IF. Therefore, the staff requests the licensee provide the basis for changing the highest RTNDT of the closure flange region that is highly stressed by bolt preload from 44 IF to -50 IF.

The closure head for each Unit was replaced. The initial RTNDT values of the new closure heads are -507F. Therefore, the P-T limit curves were developed based on the limiting initial RTNDT in the flange region, which pertains to the Unit 4 vessel flange initial RTNDT of-1 °F.

f. The EPU Licensing Report Figures and the marked up TS bases 3/4.4.9 indicate that the revised P-T limits are based oAhe KMa methodology of the 1996 Edition of ASME Code,Section XI, Appendix G. Since 1996 is an addenda rather than an edition of the ASME Code, the staff requests the licensee confirm that the revised P-T limits are based on the 1995 Edition through 1996 Addenda of the ASME Code,Section XI, Appendix G, and requests the licensee modify the TS bases accordingly.

The TS bases have been modified accordingly to cite that the P-T limit curves were developed based on the 1995 Edition through 1996 Addenda version of the ASME Code,Section XI, Appendix G. See attached Figure 1 for marked up TS bases pages 70, 71, and 75.

g. Provide the following information related to the determination of the adjusted reference temperature (ART) for the limiting RV beltline materials:
1. supporting data for, and the calculation of, the chemistry factors for those reactor vessel (RV) materials that have surveillance data; Tables 7 and 8 provide this information.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 21 of 37 Table 7 Calculation of Chemistry Factors using Turkey Point Unit 3 Surveillance Capsule Data Fluences T FF*

Material Capsule (n/cm 2 , E>1.0 FF ARTNDT ARTNDT FF2 MeV) (0 F) (OF)

T, 0.599 x 10'9 0.856 11.48(c) 9.83 0.734 Intermediate Shell S3 1.272 x 10' 9 1.067 2.83(c) 3.02 1.138 Forging 123P461VA1 Sum: 12.85 1.872 CF123P461vA,= (FF

  • ARTNDT) + 2 Y(FF ) = (12.85) + (1.872) = 6.9 0 F S3 1.272 x 1019 1.067 48.55'c) 51.80 1.138 Lower Shell Forging V3 1.223 x 1019 1.056 42.68(c) 45.08 1.115 123S266VA1 X3 2.897 x 1019 1.282 72.44(c) 92.89 1.644 Sum: 189.77 3.898 CF 12 3 ;26 6VAl= Y(FF
  • ARTNDT) + YX(FF 2) = (189.77) (3.898) = 48.7-F Davis Besse 2.956 x 10'9 1.287 188.1I'-e) (215(0) 242.1 1.657 T- 0.599 x 10' 9 0.856 141.4'el (163.87(c)) 121.1 0.734 Weld Metal Heat V3 1.223 x 10' 9 1.056 156.0(e) (180.77(c)) 164.8 1.115
  1. 71249 T4 0.649 x 10'9 0.879 18 2 .1 ) (2 1 1 (b)) 160.0 0.772 X3 2.897 x l0'9 1.282 164.9(el) (191.06(c)) 211.4 1.644 Sum: 899.5 -5.923 CF 71249= Y(FF
  • ARTNDT) + E(FF 2) = (899.5) + (5.923)= 151.9 0F Notes:

(a) Capsule fluence values were updated as part of the EPU, unless otherwise noted.

(b) Values taken from WCAP-15092, Revision 3 [Reference 9].

(c) Values taken from WCAP-15916 [Reference 10].

(d) A 9°F correction factor was used in the calculation of this value to account for the difference in operating temperatures between Turkey Point and Davis Besse.

(e) Final ARTNDT value has been adjusted using the ratio procedure. For the Davis Besse capsule, the ratio is 0.833.

For the Turkey Point Units 3 and 4 capsules, the ratio is 0.863.

(f) Values taken from WCAP-15885, Revision 0 [Reference 11].

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 22 of 37 Table 8 Calculation of Chemistry Factors Using Turkey Point Unit 4 Surveillance Capsule Data Sum: 110.6 CF122S]8OVAI= I(FF

  • ARTNDT) + X(FF 2) =(10.6)+ (1.919)= 5.5°F Davis Besse 2 .9 5 6 (g) 1.287 18 8 .1{d-f (2 15 (g)) 242.1 1.657 T3 0.599 0.856 14 1 .4(f ( 16 3 .8 7(b)) 121.1 0.734 V3 1.223 1.056 15 6 .0(0 ( 18 0 . 77(b)) 164.8 1.115 Weld Metal Heat
  1. 71249 T4 0.649 0.879 18 2 . 1 f) (2 1 1(c)) 160.0 0.772 X3 2.897 1.282 1 6 4 .9 (f ( 19 1.0 6(b)) 211.4 1.644 Sum: 899.5 5.923 CF 7 1249 = Y(FF
  • ARTNDT) + I(FF2) = (899.5) + (5.923) = 151.9°F Notes:

(a) Capsule fluence values were updated as part of the EPU, unless otherwise noted.

(b) Values taken from WCAP-15916 [Reference 10].

(c) Values taken from WCAP-15092, Revision 3 [Reference 9].

(d) A 9°F correction factor was used in the calculation of this value to account for the difference in operating temperatures between Turkey Point and Davis Besse.

(e) In order to apply this calculation, there must be at least two data points for the material.

(f) Final ARTNDT value has been adjusted using the ratio procedure. For the Davis Besse capsule, the ratio is 0.833.

For the Turkey Point Units 3 and 4 capsules, the ratio is 0.863.

(g) Values taken from WCAP-15885, Revision 0 [Reference 11].

2. the copper and nickel values for the surveillance materials; This information is provided in Table 9.

Table 9 Copper and Nickel Values for Surveillance Weld Metal Heat # 71249 Plant Cu Wt. % Ni Wt. %

Turkey Point Units 3 and 4 0.31 0.57 Davis Besse 0.33 0.57

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 23 of 37

3. the credibility evaluation of the surveillance data; and Introduction Regulatory Guide (RG) 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. Positions 2.1 and 2.2 of RG 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date, there have been four surveillance capsules removed from the Turkey Point Unit 3 reactor vessel and two from the Turkey Point Unit 4 reactor vessel. This capsule data must be shown to be credible. In accordance with the discussion of RG 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The purpose of this evaluation is to document the information provided by FPL in regard to the Turkey Point Units 3 and 4 reactor vessel surveillance data for each of the credibility requirements of RG 1.99, Revision 2.

Evaluation Criterion1: Materials in the capsulesshould be those judged most likely to be controllingwith regardto radiationembrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements," as follows:

"the reactor vessel (shell material including welds, heat affected zones, and plates orforgings) that directly surrounds the effective height of the active core and adjacent regions of the reactorvessel that arepredicted to experience sufficient neutron radiationdamage to be considered in the selection of the most limiting material with regardto radiationdamage."

The forging materials and weld metal contained in the capsules are representative of all of the materials in the Turkey Point Units 3 and 4 reactor vessel beltline regions. Therefore, this criterion is met.

Criterion2: Scatter in the plots of Charpy energy versus temperaturefor the irradiatedand unirradiatedconditions should be small enough to permit the determinationof the 30ft-lb temperature and USE unambiguously.

No surveillance capsule data has been analyzed since the time that Capsule X 3 was analyzed in WCAP-15916 [Reference 10]. Based on the plots contained in WCAP- 15916, this criterion is met.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 24 of 37 Criterion 3: When there are two or more sets of surveillance datafrom one reactor, the scatter of/ARTNDT values about a best-fit line drawn as describedin Regulatory Position2.1 normally should be less than 28°Ffor welds and 17Ffor base metal. Even if the fluence range is large ('two or more orders of magnitude), the scatter should not exceed twice those values.

Even if the datafail this criterionfor use in shift calculations,they may be credible for determining decrease in USE if the upper shelf can be clearly determined,following the definition given in ASTM E]85-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 28'F for welds and less than 17'F for forgings.

The Turkey Point Unit 3 intermediate shell and lower shell forgings and surveillance weld material will be evaluated for credibility. The Turkey Point Unit 4 lower shell forging and surveillance weld material will be evaluated for credibility. Since the plants have an integrated surveillance program, the surveillance weld material evaluation will be identical between plants and thus applicable to both plants. The weld is made from weld wire heat 71249; Turkey Point Units 3 and 4 have a sister plant that shares the same weld wire heat and thus, utilize data from a sister plant (Davis Besse).

The method of RG 1.99, Revision 2 will be followed for determining credibility of the weld as well as the forging material.

Credibility Assessment The chemistry factors for the Turkey Point Units 3 and 4 surveillance forging and weld material contained in the surveillance program were calculated in accordance with RG 1.99, Revision 2, Position 2.1 and presented in Tables 7 and 8 of this letter. A new fitted chemistry factor for the Turkey Point Units 3 and 4 weld material will be calculated only for the purposes of this credibility evaluation. For this evaluation, the adjustment for chemistry differences between the beltline weld and surveillance weld will not be taken into account. The fitted chemistry factor calculation for the weld material is shown in Table 10. The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Tables 11 and 12 for Turkey Point Units 3 and 4, respectively.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 25 of 37 Table 10 Calculated CF for Turkey Point Units 3 and 4 Weld Heat # 71249 Using Turkey Point Units 3 and 4 and Davis Besse Surveillance Capsule Data Measured Adjusted FF

  • Capsule f ARTNDT ARTNDT ARTNDT FF2 Material Capsule (X1019 n/cm 2) FF (OF) (OF) (OF)

Davis Dess 2.956 1.287 215 224 288.3 1.657 BesseI T3 0.599 0.856 163.9 163.9 140.4 0.734 Surveillance V3 1.223 1.056 180.8 180.8 190.9 1.115 Weld Metal T4 0.649 0.879 211 211 185.4 0.772 Heat #71249 X3 2.897 1.282 191.1 191.1 245.0 1.644 Sum: 1050.0 5.923 CF 71 249 = Z(FF

  • ARTNDT) + X(FF 2) = (1050.0) + (5.923) = 177.3°F Table 11 Turkey Point Unit 3 Surveillance Capsule Data Scatter about the Best-Fit Line CF Adjusted Predicted Scatter <17 0 F (SIopebest fit) Capsule f ARTNDT ARTNDT ARTNDT a) (Base Metal)

Material Capsule (OF) (xl019 n/cm 2) FF (OF) (OF) (OF) <28 0 F (Weld)

IS Forging T3 6.9 0.599 0.856 11.5 5.9 5.6 Yes 123P461VA1 S3 6.9 1.272 1.067 2.8 7.3 4.5 Yes S3 48.7 1.272 1.067 48.6 51.9 3.4 Yes 123S 266VAi V3 48.7 1.223 1.056 42.7 51.4 8.7 Yes X3 48.7 2.897 1.282 72.4 62.4 10.0 Yes Davis Bess 177.3 2.956 1.287 224 228.2 4.2 Yes Besse Surveillance T3 177.3 0.599 0.856 163.9 151.9 12.0 Yes Weld Metal V3 177.3 1.223 1.056 180.8 187.2 6.5 Yes Heat #71249 T4 177.3 0.649 0.879 211 155.8 55.2 No X3 177.3 2.897 1.282 191.1 227.4 36.3 No Note:

(a) For the ARTNDT scatter, absolute values are listed.

Turkey Point Unit 3 Table 11 indicates that zero of the surveillance data points fall outside the +/-1 c7 of 17°F scatter band for base metals. Therefore, the intermediate shell forging and lower shell forging data is deemed "credible" per the third criterion. Table 11 indicates that two of the five surveillance data points fall outside the +/-1 of 28°F scatter band for surveillance weld materials. Therefore the surveillance weld data is deemed "not credible" per the third criterion.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attaclunent Page 26 of 37 Table 12 Turkey Point Unit 4 Surveillance Capsule Data Scatter about the Best-Fit Line CF Adjusted Predicted Scatter <17 0 F (Slopebest fit) Capsule f ARTNDT ARTNDT ARTNDT(a (Base Metal)

Material Capsule (OF) (x10 19 n/cm2) FF (OF) (OF) (OF) <280 F (Weld)

IS Forging S4 N/A 1.29 1.071 35 N/A N/A N/A 123P481VA1 LS Forging T4 5.5 0.649 0.879 12 4.8 7.2 Yes 122S180VA1 S4 5.5 1.29 1.071 0 5.9 5.9 Yes Davis Dess 177.3 2.956 1.287 224 228.2 4.2 Yes Besse Surveillance T3 177.3 0.599 0.856 163.9 151.9 12.0 Yes Weld Metal V3 177.3 1.223 1.056 180.8 187.2 6.5 Yes Heat#71249 T4 177.3 0.649 0.879 211 155.8 55.2 No X3 177.3 2.897 1.282 191.1 227.4 36.3 No Note:

(a) For the ARTNDT scatter, absolute values are listed.

Turkey Point Unit 4 Table 12 indicates that zero of the surveillance data points fall outside the +1a of 170 F scatter band for base metals. Therefore, the lower shell forging data is deemed "credible" per the third criterion. Table 12 indicates that two of the five surveillance data points fall outside the +/- 1a of 28°F scatter band for surveillance weld materials. Therefore the surveillance weld data is deemed "not credible" per the third criterion.

Criterion4: The irradiationtemperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25 0F.

The capsule specimens are located in the reactor between the neutron pad and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pad. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F.

Hence, this criterion is met.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 27 of 37 Criterion 5: The surveillance datafor the correlationmonitor materialin the capsule shouldfall within the scatter band of the databasefor that material.

The Turkey Point Unit 3 surveillance program does contain correlation monitor material. This evaluation will be re-evaluated using the updated surveillance capsule fluence values. NUREG/CR-6413, ORNL/TM-13133

[Reference 12], contains a plot of Residual vs. Fast Fluence for the correlation monitor material (Figure 10 of the NUREG Report). The Figure shows a 2cy uncertainty of 50'F. The data used for this plot is contained in Table 15 in the NUREG Report. However, the data in the NUREG report has not been considered for the recalculated fluence values as documented herein. Thus, Table 13 below presents an updated calculation of Residual vs. Fast Fluence.

Table 13 Calculation of Residual vs. Fast Fluence Capsule Fluence Fluence Measured RG 1.99 Shift Residual (x101 9 n/cm 2) Factor Shift (CF*FF(b) (Measured -

(FF) (OF) (OF) RG Shift)

$3 1.272 1.067 106.7(a) 106.7 0 T, 0.5990 0.856 86.66(al 85.6 1.1 V3 1.223 1.056 100.32 105.6 5.3 Notes:

(a) USE T@30 values taken from WCAP-15916 [Reference 10].

(b) Per NUREG/CR-6413, ORNL/TM- 13133, the Cu and Ni values for the correlation monitor material are 0.20 and 0.18, respectively. This equates to a chemistry factor of 100F from RG 1.99, Revision 2.

Table 13 shows a 2y uncertainty of less than 50'F, which is the allowable scatter in NUREG/CR-6413, ORNL/TM- 13133. Hence, this criterion is met.

Conclusion Based on the preceding responses to all five criteria of RG 1.99, Rev 2, Section B, the Turkey Point Unit 3 intermediate shell forging and lower shell forging surveillance data is deemed "credible," but the weld data is deemed "not credible." The Turkey Point Unit 4 lower shell forging surveillance data is deemed "credible," but the weld data is deemed "not credible."

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 28 of 37

4. whether the ratio procedure of Regulatory Guide 1.99, Rev. 2, Position 2.1 was used.

The ratio procedure in RG 1.99, Revision 2, Position 2.1, was used in the chemistry factor (CF) calculations. As described in footnote (e) in Table 7 above, certain ratios were applied for the Turkey Point Units 3 and 4 weld metal and the Davis Besse weld metal. The calculations of these ratios are detailed below.

Turkey Point Units 3 and 4 Reactor Vessel Beltline Weld Heat # 71249 Cu Wt. % = 0.23 Ni Wt. % = 0.59 CFBeltline Weld = 167.6 (using Table 1 of RG, Revision 2)

Turkey Point Units 3 and 4 Surveillance Weld Heat # 71249 Cu Wt. % = 0.31 Ni Wt. % = 0.57 CFsurveillance Weld = 194.1 (using Table 1 of RG 1.99, Revision 2)

Thus, the ratio for the Turkey Point Units 3 and 4 surveillance weld is as follows:

Ratio = CFBeltline Weld / CFsurveillance Weld = 167.6°F/1 94.1 'F = 0.863 Davis Besse Surveillance Weld Heat # 71249 Cu Wt. % 0.33 Ni Wt. % =0.57 CFsurveillance Weld = 201.3 (using Table 1 of RG 1.99, Revision 2)

Thus, the ratio for the Davis Be*sse surveillance weld is as follows:

Ratio = CFBeltline Weld / CFsurveillance Weld = 167.6 0 F/201.3°F = 0.833 CVIB-1.5 Reference 2, Section 2.1.4.2.5 concludes that the new EPU environmental conditions (chemistry, temperature, and neutron fluence) will not introduce any new aging effects on the RVI components during 60 years of operation, nor will the EPU change the manner in which the component aging will be managed by the aging management program credited in the topical report WCAP-14577, Rev. 1-A, "License Renewal Evaluation: Aging Management of Reactor Internals," and accepted by the NRC in the Safety Evaluation Report (SER). The susceptibility of the Turkey Point, Units 3 and 4 RVI components to these aging effects was also assessed for license renewal as documented in the License Renewal Application (LRA) for Turkey Point Units 3 and 4 and the associated SER, NUREG-1759.

Although the licensee stated that there will be no new aging effects, Reference 2 does not address whether particular RVI components will become susceptible to additional aging effects due to higher neutron fluences, temperatures, or stresses introduced by the EPU. The staff therefore requests the following information:

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 29 of 37

a. Describe the method of determining if additional RVI components become susceptible to the aging effects of 1) cracking due to stress corrosion cracking (SCC), irradiation assisted stress corrosion cracking (IASCC), or primary water stress corrosion cracking (PWSCC); 2) reduction of fracture toughness due to irradiation embrittlement (IE); 3) loss of material due to wear; 4) loss of mechanical closure integrity due to IASCC, IE, irradiation creep, or stress relaxation (SR); and 5) loss of preload due to SR, or dimensional change due to void swelling. The discussion should address whether a detailed fluence and temperature map was used, and whether stresses in individual components were reevaluated.

ala: SCC is a synergistic degradation mechanism requiring stress, environment, and a susceptible material. Eliminate any of the required three and SCC will not occur. As identified in License Renewal Application (LRA) Table 3.2-1 all internals components have already been identified as requiring aging management to control SCC. The Turkey Point chemistry controls program maintains rigorous control of reactor coolant chemistry; the increase in temperature or stress due to the EPU therefore will not increase the susceptibility to SCC for the extended license period.

alb: For IASCC to have a potential to occur both sufficient fluence and stress are required; temperature is not included in current industry standard thresholds for evaluating IASCC. In accordance with WCAP-14577 [Reference 13],

lx10 2 1n/cm 2 (E >0.1MeV) and 30 ksi stress are threshold values used to screen in susceptibility to IASCC. For IASCC, the following components were not previously identified in LRA Table 3.2-1 as being susceptible:

radial keys and clevis inserts, upper core plate alignment pins, core barrel outlet nozzle diffusers, upper support plates and colunms, secondary core support, upper core plate, head/vessel alignment pins, guide tubes and guide pins, internals holddown spring, bottom mounted instrumentation (BMI) columns and upper instrumentation columns. An updated fluence map has shown that fluences exceeding lx 1021 n/cm 2 (E>0. 1MeV) extends from the upper core plate down to 9.5" below the lower core plate. The following table shows that two components not currently identified in LRA Table 3.2-1 as requiring aging management for IASCC (upper core plate and portions of the BMI columns) are within this region. The operating stresses in the BMI columns are well below the threshold for IASCC. WCAP-14577 did not originally identify the upper core plate as a component with a fluence greater than 1x10 21 n/cm 2 (E>0.1MeV). However, the updated fluence calculations indicate that the upper core plate fluence at 60 years will exceed this threshold. The higher fluence results from a combination of the plant uprating and updated calculation methodologies.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 30 of 37 Fluence Stress (Pm+Pb+Q) radial keys and clevis inserts < Ix10 21n/cm 2 (E>O.IMeV) upper core plate alignment pins < lx10 2" n/cm 2 (E>O.1MeV) core barrel outlet nozzles < Ix10 2' n/cm 2 (E>O.1MeV) diffusers < ] x1 0 21n/cm 2 (E>0.IMeV) upper support plates < 1x 102' n/cm2 (E>0.IMeV) upper support columns < 1 x10 21 n/cm 2 (E>0. I MeV) upper core plate > lx10 2' n/cm2 (E >0.1MeV) > 30 ksi secondary core support < 1 x 021 n/cm 2 (E>0. I MeV) head/vessel alignment pins < I x 1021 n/cm 2 (E>0.IMeV) guide tubes < Ix10 21 n/cm 2 (E>0.IMeV) guide pins < Ix10 2- n/cm 2 (E>0.1MeV) internals holddown spring <1x10 2' n/cm2 (E>0.IMeV)

BMI columns > lx10 21 n/cm 2 (E>0.IMeV) < 30 ksi upper instrumentation columns < 1x10 21 n/cm 2 (E>0.1MeV)

  • Guide pins were replaced in 2007 & 2008
    • Upper portions closest to the lower core support plate al c: Similar to SCC, PWSCC is a synergistic degradation mechanism requiring stress, environment, and a susceptible material. For nickel-base materials that are susceptible to PWSCC, all internals components made of nickel-base materials have already been identified in LRA Table 3.2-1 as requiring aging management to control PWSCC. The minimal temperature increase due to the EPU, which is the primary influence on PWSCC, is not expected to increase significantly the susceptibility of nickel-base materials during the extended license period.

a2: For IE to have a potential to occur sufficient fluence is required; stress and temperature do not influence IE. In accordance with WCAP-14577, 1x10 2 1 n/cm 2 (E>O. 1MeV) is the threshold value used to screen in susceptibility to IASCC. For IE the following components were not previously identified in LRA Table 3.2-1 as being susceptible: radial keys and clevis inserts, upper core plate alignment pins, core barrel outlet nozzle diffusers, upper support plates and columns, head/vessel alignment pins, guide tubes and guide pins, internals holddown spring, BMI columns and upper instrumentation columns. An updated fluence map has shown that fluences exceeding Ix10 2 1 n/cm 2 (E >0.1MeV) extends from the upper core plate down to 9.5" below the lower core plate. As discussed in the response to CVIB-1.5alb, previous estimates of the upper core plate fluence at 60 years have been below this threshold. The following table shows that two components (upper core plate and portions of the BMI columns) exceed the fluence threshold used in WCAP-14577 to identify components with potential IE concerns.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 31 of37 Fluence radial keys and clevis inserts < x"10 21 n/cm 2 (E>O.lMeV) upper core plate alignment pins < 1x10 2' n/cm 2 (E>O.1MeV) 2 2 core barrel outlet nozzles < 1x10 1 n/cm (E>0.1MeV) diffusers < Ix1 0 n/cm 2 (E>0.1MeV) 2 1 upper support plates < 1x 10 2' n/cm 2 (E>O.1MeV) upper support columns < 1xl0 21 n/cm 2 (E>O. IMeV) upper core plate > x 102 1n/cm2 (E >0.1MeV) secondary core support < lxi 02" n/cm2 (E>O. 1MeV) head/vessel alignment pins < Ix 1021 n/cm2 (E>O.IMeV) guide tubes < lxl02 .(n/cm 2 (E>0.1MeV) guide pins < 1x1021 n/cm 2 (E>0.1MeV) internals holddown spring < 1xI0 2 1 n/cm 2 (E>0. 1MeV)

BMI columns > 1xl0 2 1 n/cm2 (E>0.1MeV) upper instrumentation columns < 1x10 2' n/cm 2 (E>0.1MeV)

  • Guide pins were replaced in 2007 & 2008
    • Upper portions closest to the lower core support plate a3: Loss of material due to wear is a flow dependent phenomenon. Calculations completed for the EPU determined that the EPU will result in a minimal increase in the expected best estimate flow in the reactor coolant system of 0.2%. This was evaluated and it was concluded that this minor increase in flow will not result in any new RVI components being susceptible to the loss of material due to wear during the extended license period.

a4: Loss of mechanical closure integrity, including loss of preload, applies to core support bolting. All RVI bolting has already been identified in LRA Table 3.2-1 as being susceptible to loss of mechanical closure integrity.

There are no chemistry changes due to the EPU and changes in stress or temperature are not expected to change how bolting is managed during the period of extended license.

a5: Besides core support bolting the holddown spring would be the only other RVI component susceptible to loss of preload and it is identified as such in LRA Table 3.2-1. Westinghouse evaluated the performance of the holddown spring with respect to the EPU. It was determined that the reactor internals would remain seated and stable for the EPU conditions for the extended license period.

With respect to void swelling, joint industry testing has been conducted since publication of WCAP-14577. Based upon this testing, the industry is currently using l.3x1022 n/cm2 (E>l.OMeV), as published in MRP-175

[Reference 14], as a threshold for void swelling. While some internals components will exceed this value there have been no indications from the different bolt removal programs that there are any discernable effects attributed to swelling. Turkey Point will continue to participate and follow up industry efforts to investigate swelling effects of the core components.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 32 of 37

b. Confirm whether the design projections of gamma heating bound the projected amount of gamma heating of the RVI under EPU conditions.

Discuss the acceptability of the effects of gamma heating on the RVI components under EPU conditions.

Gamma heating rates for the RVI under EPU conditions were explicitly determined and compared with design values as part of the EPU Program. The heating rates calculated at EPU conditions were all less than the design heating rates for the RVI. Thus, there is no impact on the RVI with respect to gamma heating rates under EPU conditions.

c. Clarify whether any additional RVI components were determined to be susceptible to the aging effects listed in part "a" of this question as a result of EPU, compared to those listed as susceptible to these mechanisms in Table 3.2-1 of the LRA for Turkey Point, Units 3 and 4.

Compared to components listed as susceptible to the mechanisms of Table 3.2-1 of the LRA, the upper core plate may be susceptible to IASCC and the upper core plate and portions of the BMI columns may be susceptible to IE.

CVIB-1.6 Several aging effects identified for RVI in the LRA for Turkey Point, Units 3 and 4, are not evaluated in the EPU evaluation of RVI materials. The SER related to the Turkey Point, Units 3 and 4 LRA, NUREG-1759, concurred with the aging effects requiring management for the RVI. The staff requests the licensee provide an evaluation of the effects of EPU on the following aging effects requiring management, or explain why the aging effect did not require reevaluation.

a. LRA Section 3.2.5.2.3 stated that loss of material due to mechanical wear is an aging effect requiring management for the period of extended operation.

Loss of material due to wear can occur on the lower core plate fuel pins, core barrel flanges, guide tubes and guide pins, upper core plate alignment pins, and radial keys and clevis inserts.

Loss of material due to wear is a flow dependent phenomenon. Calculations completed for the EPU determined that the EPU will result in a minimal increase in the best estimate flow in the reactor coolant system of 0.2%. It was concluded that this minor change would have a negligible impact on the wear of the lower core plate fuel pins, core barrel flanges, guide tubes and guide pins, upper core plate alignment pins, and radial keys and clevis inserts.

b. The LRA indicates loss of mechanical closure integrity due to SCC and SR is an aging effect for upper support column, guide tube, and clevis insert bolting. For baffle-former bolting and barrel-former bolting, loss of mechanical closure integrity can be caused by IASCC, 1E, irradiation creep, and irradiation-assisted SR.

All RVI bolting has already been identified in LRA Table 3.2-1 as being susceptible to loss of mechanical closure integrity. With respect to SCC, there are no chemistry changes due to the EPU and changes in stress or temperature are not expected to change how bolting is managed during the period of

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 33 of 37 extended license. With respect to SR, the minimal changes in temperature and fluence due to the EPU are not expected to change how bolting is managed during the period of extended license.

With respect to baffle-former and barrel-former bolting, these bolts receive the highest internals fluence which is well above known industry thresholds for fluence induced aging mechanisms such as IASCC, IE, irradiation creep, and irradiation-assisted SR (e.g. MRP-175 or WCAP-14577). The minimal increases in temperature and fluence due to the EPU are not expected to change management of such bolting.

c. The LRA indicates loss of preload due to SR can occur for the RVI hold-down spring.

Westinghouse evaluated the performance of the holddown spring with respect to the EPU, considering the effects of SR during the extended license period (60 years). It was determined that there will be no significant impact on the loss of preload during the extended license period (60 years) and the reactor internals will remain seated and stable for the EPU conditions.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 34 of 37 References

1. M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-113), "License Amendment Request No. 205: Extended Power Uprate (EPU)," (TAC Nos. ME4907 and ME4908), Accession No. ML103560169, October 21, 2010.
2. Email from J. Paige (NRC) to S. Franzone (FPL), "EPU Acceptance Review Question Re:

Equivalent Margin Analysis," Accession No. ML103070063, November 1, 2010.

3. M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-268), "Response to NRC Request for Additional Infornmation (RAI) Regarding Extended Power Uprate (EPU) License Amendment Request (LAR) No. 205 and Equivalent Margin Analysis (EMA)," November 12, 2010.
4. M. Kiley (FPL) to U.S. Nuclear Regulatory Commission (L-2010-303), "Supplemental Response to NRC Request for Additional Information (RAI) Regarding Extended Power Uprate (EPU) License Amendment Request (LAR) No. 205 and Equivalent Margin Analysis (EMA)," Accession No. ML103610321, December 21, 2010.
5. AREVA NP Document 51-9133594-000, "Turkey Point Units 3 and 4 EMA Reconciliation,"

March 12, 2010.

6. AREVA NP Document 77-2312-003 (ANP-2312P, Rev 3), "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Turkey Point Units 3 and 4 for Extended Life Through 48 Effective Full Power Years," January 2010.
7. Email from J. Paige (NRC) to T. Abbatiello (FPL), "Turkey Point EPU - Vessels and Internals Integrity (CVIB) Requests for Additional Information - Round 1, Accession No. ML110420241, February 11, 2011.
8. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D. Andrachek et al.,

May 2004.

9. WCAP-15092, Revision 3, "Turkey Point Units 3 and 4 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," T. J.

Laubham and J. H. Ledger, May 2000.

10. WCAP-15916, Revision 0, "Analysis of Capsule X from the Florida Power and Light Company Turkey Point Unit 3 Reactor Vessel Radiation Surveillance Program," J. H. Ledger et al., September 2002.
11. WCAP-15885, Revision 0, "R. E. Ginna Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, July 2002.
12. ORNL/TM- 13133; NUREG/CR-6413, "Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials," J. A. Wang, Oak Ridge National Laboratory, Oak Ridge, TN, April 1996
13. WCAP-14577, Revision 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals," March 2001.
14. MRP-175, Revision 0, "Materials Reliability Program: PWR Internals Aging Degradation Mechanism Screening and Threshold Values," December 2005.

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 35 of 37 Figure 1: Modified P-T Limits TS Bases Procedure No.: Procedure Title' Page: 70 Approval Date:

0-ADM-536 Technical Specification Bases Control Program 1/19/10 ATTACHMENT I (Page 59 of 112)

TECHNICAL SPECIFICATION BASES 3/4.4.9 (Cont'd)

During heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and which are tensile at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the outside surface location. Hlowever. since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiting. Consequently for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.

During cooldown., the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which are compressive at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location. Since the neutron irradiation damage is also greatest at the inside surface location, the inside surface flaw is the limiting location. Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section44-, Appendix G: '- 15 Edtion through 1996 Addenda of the

[ 1. The reactor coolant temperature and pressure and system heatup and cooldown%rates (with the exception of the pressurizer) shall be limited in accordance with Figures ; ; .4e .4-4 for the service period specified thereon: *3.-and3.4-3

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and 3.-24.o3.3
  • ..2an--
b. Figures ;." 4 4m. -.. 4 define limits to assure prevention .of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
2. These limit lines shall be calculated periodically using methods provided below.
3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F.

W2Otl3DPSiln/eisld~s

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 36 of 37 Figure 1: Modified P-T Limits TS Bases (continued)

Procedure No.: Procedure

Title:

Page:

71 J Alpproval Date:

0-ADM-536 Technical Specification Bases Control Program 1/19/10 ATTACHNMENT 1 (Page 60 of 1.1.2)

WCAP-1 4040-NP- TECHNICAL SPECIFICATION BASES A, Revision 2, 3/4.4.9 (Cont'd)

Methodology Used 0

to Develop Cold 4. The pressurizer beatup and cooldown rates shall not exceed 100 F/h and Overpressure 200'F7,1. respectively. The spray shall not be used if the temperature Mitigating System difference between the pressurizer and the spray fluid is greater than.

Setpoints and RCS 3207F, and Heatup and 5. System preservice hydrotests and inservice leak and hydrotests shall be Cooldown Curves. performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XG.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, the version of the ASTM E185 standard required by 10 CFR 50, Appendix H, and in accordance with additional r* r*vessel requirements.'-.-Fk The pr rties are then evaluated in accordance with Appendix G of the 4 Editi, of Section of the ASME Boiler and Pressure Vessel Code and the additional require, ts

- of 10 CFR 50,

, 1.12, Appendix ~.-z Gfor.and the calculation methods described in Westinghouse 'eport Do-,elspn Q -*CSP- ON -1 an-d CO- ldc;-on- G:.-,'. SY 48 throulh 1996 Addenda I Heatup and cooldown limit urves are calcu ated using the ost limiting val.ue of the nil-ductility reference tempelure, ],at effective full power years (EFPY) of service

  • ,at the 1/4Tlife. The incu9 location EFPconre ervice the region life period than is greater is chosen the such *,of that the limiting the limiting prepared by rmining the most conservative case with either the inside or outside wall controlling, for any Lz'a rate u. tz 100 degrees F per hour and cooldown rateso~-e 100 degrees F per hour. The heatup and cooldown curves were prepared bas*.-l6n the most limiting value of predicted adjusted reference temperature at the end olhe appicable service period ( PY... 48040 and*

The reactor vessel materials have been tested to determne their initial I  ; the nesults of these tests are shown in Tables B 3/4.41 and B 3/4.42. Reactor operaoil ,ind resultant fast neutron (r greater than S MeVm irradiation cacause an increase incthe w Therefore, an adjusted reference temperature, based upon the fluence d chempistry fac rs of the material has been predicted using Regulatory Guide 1.99, Revisionr dated w[ay 1988. Radiation Embrittlement of Reactor ad Vessel Materials. The heatup cooldc wn imitocurves of Figuresf. , ". . include predicted adjustments fo r this shic ain sat the end of the applicableIrvice period.

] 13.4an 3.4-3 I

  • Topical Report BAW-2308, Revision 2-A is the source for the initial weld materials properties for Linde 80 welds. I W2OO3:OPSIn/dos'cls

Turkey Point Units 3 and 4 L-2011-029 Docket Nos. 50-250 and 50-251 Attachment Page 37 of 37 Figure 1: Modified P-T Limits TS Bases (continued)

ProcedreN.: Pocedue,, Titl, Page:

Appoovrl Dale:

O-ADM-536 Technical Specification Bases Control Program 1 1/19/10 WCAP-14040-NP-A. Revision 2, "Methodology Used to 1 Develop Cold Overpressure PATTACHMeE6NT 1 Mitigating System Setpoints (Page 64 of 112) and RCS Heatup and TECHNICAL SPECIFICATION BASES Limit Curves.' -Cooldow Section X1 of the 1995 Edition 3/4.4.9 (Cont'd) through 1996 Addenda Allowable pressure-temperature relationships for variousheatup and cooldown rates are calculated using methods derived from Appendix G in geeii.- IR of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and Westinghouse Reportep C A i..1;, P....@ai -_- .. op, i_..... . s8 , ... , S..

Appendix G of the The general method for calculating heatup and cooldown limit curves is based upon the 1995 Edition through principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation 1996 Addenda of procedures a semi-elliptical surface defect with a depth of one-quarter of the wall ASME Section XI thickness, T, and a length of 312T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in A ':::"C .. f A.;M S:  :. Al as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature.. -t ;, is used and this includes the radiation-induced shift, , cornesponding to th nd of the period for which heatup and cooldown curve4are ;enerated. RTND]

The ASME approach for calculating the allowable limit ,-es for various heatup and cooldown rates specifies that the total stress intensity factor, 4d, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, 44, for the metal temperature at that time. is obtained from the reference fractur.Aoughness curve, defined in Appendix G to thqt ASME Code. Tb 444curve is given by th\q uatio: --

KIa 4 26.78 + 1.223 exp [0.0145(T-RTNDr, + 1.60)] T (1)

Where: 4"R-is the reference stress intensity factor as function of the metal temperature T and the metal nil-ductility reference temperature R . Thus, the governing equation p-cooldown analysis is defined in Appendix G of the ASME Code as follows:

C Kim + KITt- 47r (2)

Where: Kni = the stress intensity factor caused by membrane (pressure) stress, Krr = the stress intensity factor caused by the thermal gradients,

= constant provided by the Code as a function of temperature relative to the RTNTDT of the material, C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

Retrieved from "https://kanterella.com/w/index.php?title=L-2011-029,_Response_to_NRC_Request_for_Additional_Information_Regarding_Extended_Power_Uprate_License_Amendment_Request_No._205_and_Reactor_Materials_Issues_-_Round_1&oldid=2551684"