NL-08-1758, Submittal of Revision 4 of the Pressure Temperature Limits Report

Submittal of Revision 4 of the Pressure Temperature Limits Report

ML083220145
Person / Time
Site:	Farley
Issue date:	11/14/2008
From:	Ajluni M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-08-1758
Download: ML083220145 (30)

Text

Southern Nuclear Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201-1295 Tel 205.992.5000 November 14, 2008 SOUTHERN A COMPANY Energy to Serve Your World'"

Docket No.:

50-348 NL-08-1758 U. S. Nuclear Regulatory Commission AnN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit 1 Submittal of Revision 4 of the Pressure Temperature Limits Report Ladies and Gentlemen:

In accordance with Section 5.6.6 of the Joseph M. Farley Nuclear Plant (FNP)

Unit 1 Technical Specifications, Southern Nuclear Operating Company (SNC) hereby submits Revision 4 of the FNP Unit 1 Pressure Temperature Limits Report (PTLR).

Revision 4 of the PTLR updates the surveillance capsule withdrawal schedule, the data credibility analysis and the supplemental data sections to reflect the surveillance capsule analysis report (WCAP-16964-NP, submitted October 8, 2008) for Capsule Z, the sixth and final capsule. In addition, Revision 4 includes new pressure temperature limit curves, which were developed based on the updated capsule analysis using existing NRC-approved methodology.

An ex-vessel neutron dosimetry system has been installed to enable long term monitoring of the reactor vessel following withdrawal of the last capsule.

This letter contains no NRC commitments. If you have any questions, please advise.

Sincerely,

~~~

M. J. Ajluni Manager, Nuclear Licensing MJAlDWD/daj

U. S. Nuclear Regulatory Commission NL-08-1758 Page 2

Enclosure:

Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 1, Revision 4, October 2008 cc:

Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. H. Jones, Vice President - Engineering RTYPE: CFA04.054; LC# 14864 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. K. D. Feintuch, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley

Joseph M. Farley Nuclear Plant - Unit 1 Submittal of Revision 4 of the Pressure Temperature Limits Report Enclosure Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 1, Revision 4, October 2008

SOUIIIERNA COMPANY Joseph M. Farley Nuclear Plant Pressure Temperature Limits Report Unit 1 Revision 4 October 2008

PTLR for FNP Unit 1 Revision 4 Page 1 of 26 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 Pressureffemperature (Pff) Limits (LCO - 3.4.3) 5 2.2 RCP Operation Limits 5

2.3 LTOP Arming Temperature (LCO - 3.4.12) 5 3.0 Reactor Vessel Material Surveillance Program 12 4.0 Reactor Vessel Surveillance Data Credibility 13 5.0 Supplemental Data Tables 17 6.0 References 25

PTLR for FNP Unit 1 Revision 4 Page 2 of 26 List of Tables 2-1 Farley Unit 1 30 EFPY Heatup Curve Data Points 8

2-2 Farley Unit 1 30 EFPY Cooldown Curve Data Points 9

3-1 Surveillance Capsule Withdrawal Schedule 12 4-1 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 15 4-2 Scatter of ~RTNOT Values About a Best-Fit Line for Surveillance Plate Material 16 4-3 Scatter of ~RTNOT Values About a Best-Fit Line for Surveillance Weld Material 16 5-1 Comparison of Surveillance Material 30 ft-Ib Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions 18 5-2 Calculation of Chemistry Factors Using Surveillance Capsule Data 19 5-3 Reactor Vessel Toughness Table (Unirradiated) 20 5-4 Reactor Vessel Fluence Used in PTS Evalluation Projections for 36 EFPY.......... 21 5-5 Summary of Adjusted Reference Temperatures (ARTs) for Reactor Vessel Beltline Materials at the 1/4-T and 3/4-T Locations for 30 EFPY 22 5-6 Calculation of Adjusted Reference Temperature at 30 EFPY for the Limiting Reactor Vessel Material...

23 5-7 Pressurized Thermal Shock (RTPTS) Values for 36 EFPY 24

PTLR for FNP Unit 1 Revision 4 Page 3 of 26 List of Figures 2-1 Farley Unit 1 Reactor Coolant System Heatup Limitations 6

2-2 Farley Unit 1 Reactor Coolant System Cooldown Limitations 7

PTLR for FNP Unit 1 Revision 4 Page 4 of 26 This page intentionally blank.

PTLR for FNP Unit 1 Revision 4 Page 5 of 26 1.0 ReS 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 PressurelTemperature Limits (PIT) Limits. All TS requirements associated with low temperature overpressure protection (LTOP) are contained in TS 3.4.12, 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 operability requirements associated with LTOP are specified in TS LCO 3.4.12 and were determined to adequately protect the RCS against brittle fracture in the event of an LTOP transient in accordance with the methodology specified in TS 5.6.6. The limitation on the number of operating reactor coolant pumps (RCPs) is necessary to assure operation consistent with the pressure corrections incorporated in the PIT limits for flow losses associated with the RCPs.

2.1 RCS PressurelTemperature (PIT) 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.

b.

c.

A maximum heatup of 100°F in anyone hour period.

A maximum cooldown of 100°F in anyone hour period.

A maximum temperature chal1ge of less than or equal to 10°F in anyone hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 The RCS PIT 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 is limited to one at RCS temperatures less than 110°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 Arming Temperature (LCO - 3.4.12) 2.3.1 The Low Temperature Overpressure Protection (LTOP) system arming temperature is 325°F.

PTlR for FNP Unit 1 Revision 4 Page 6 of 26 Figure 2*1 Farley Unit 1 Reactor Coolant System Heatup limitations[11 (Heatup Rates up to 1OO°F/hr) Applicable to 30 EFPY (adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 110°F). Includes vessel flange requirements of 180°F and 561 psig per 10 CFR 50, Appendix G[21.

2500 2250 2000 1750 8'

~ 1500

s en! 1250 Q.

"S

~ 1000

~

750 500 250 o

o 50 100 150 200 250 300 Oper1im Version:5.2 Run:6373 Oper1i~.x1S Version: 5.2 IJ I I I I I

",m...,",<<..**,,_....... ~.*~._.....-.'=.......... Leak Test LimitL_-J.....~...

i I

IN J

I HUnacceptable I I

/ J Acceptable ~

Operation I

/

..--\\ Operation I

I Limiting Materlal-1/4 T at 30 EFPY Lower Shell Plate B6919*1 I

Critical Limit I ART =177°F r

60 Oeg. F/Hr I Limiting Material* 3/4 T at 30 EFPY Lower Shell Plate B6919*1 ART =150°F I

I.........

I J

Critical Limit,II 100 Oeg. F/Hr e--._._.

-~~~-

i~ -,,+

J~ fl-

....~"....

II Heatup Ra~1 60 Oeg. F/Hr I

~v -

I Criticality Limit based on

~

inservlce hydrostatic test temperature (296°F) for the ---

...,~

j service period up to 30 EFPY I

Heatup Rate I

-- 100 Oeg. F/Hr C---'"

1~~lInimumBoltup I I

TemDerature =60°F 350 400 450 500 550 Indicated Temperature (Oeg. F)

PTLR for FNP Unit 1 Revision 4 Page 7 of 26 Figure 2-2 Farley Unit 1 Reactor Coolant System Cooldown Limitations[1]

(Cooldown Rates up to 100°F/hr) Applicable to 30 EFPY (adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 110°F). Includes vessel flange requirements of 180°F and 561 psig per 10 CFR 50, Appendix G[2].

2500 I

2250 2000 1750.

6' If 1500

-+--

750 500 250 o

Oper1im Version:5.2 Run:6373 Oper1im.x1s Version: 5.2 I Ii I

HUnacceptable I I

Operation I I

I I

J Limiting Material* 1/4 T at 30 EFPY /

Lower Shell Plate 86919-1 ART =177°F Limiting Material* 314 T at 30 EFPV / I Lower Shell Plate 86919*1 ART =150°F I

I I

I I J I

Cooldown Rates FlHr steady-state

• 20

-40 fII

~

-100 I...~~~ ':/

---.-~/

~ V I

+1,Minimum 80ltup

.\\

Temoerature =60°F Acceptable ~

Operation

~

n

~

~-- -~

~

I

---~-_.

i I

I I

I 400 450 500 550 o

50 100 150 200 250 300 350 Indicated Temperature (Deg. F)

PTLR for FNP Unit 1 Revision 4 Page 8 of 26 Table 2*1 Farley Unit 1 - 30 EFPY Heatup Curve Data Pointsl1]

(adjusted to include 60 psi AP at RCS temperatures ~ 110°F and 27 psi AP at RCS temperatures < 11 0°Fi2]

Leak Test Limit 60°F/hr.

Heatup 60°F/hr.

Criticality 100°F/hr.

Heatup 100°F/hr.

Criticality T

(OF) p (psig)

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig) 275 2000 60 0

296 0

60 0

296 0

275 2000 60 504 296 471 60 461 296 428 296 2485 65 504 296 471 65 461 296 429 296 2485 70 504 296 472 70 461 296 429 75 504 296 473 75 461 296 430 80 504 296 475 80 461 296 431 85 504 296 477 85 461 296 433 90 504 296 479 90 461 296 434 95 504 296 483 95 461 296 437 100 505 296 483 100 461 296 439 105 508 296 489 105 461 296 441 110 479 296 493 110 428 296 445 115 483 296 496 115 428 296 447 120 489 296 503 120 429 296 453 125 496 296 506 125 430 296 453 130 503 296 511 130 433 296 460 135 511 296 520 135 437 296 463 140 520 296 530 140 441 296 469 145 530 296 541 145 447 296 476 150 541 296 553 150 453 296 478 155 553 296 561 155 460 296 488 160 561 296 561 160 469 296 491 165 561 296 561 165 478 296 500 170 561 296 561 170 488 296 509 175 561 296 561 175 500 296 512 180 561 296 629 180 512 296 526 180 561 296 648 180 512 296 541 180 629 296 669 180 512 296 557 185 648 296 691 185 526 296 574 190 669 296 715 190 541 296 593 195 691 296 740 195 557 296 614

PTLR for FNP Unit 1 Revision 4 Page 9 of 26 Table 2*1 (continued)

Farley Unit 1 - 30 EFPY Heatup Curve Data Points[1]

(adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 11 0°F),2]

Leak Test Limit 60°F/hr.

Heatup 60°F/hr.

Criticality 109°F/hr.

Heatup 100°F/hr.

Criticality T

(OF) p (psig)

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig) 200 715 296 767 200 574 296 636 205 740 296 797 205 593 296 659 210 767 296 829 210 614 296 685 215 797 296 863 215 636 296 713 220 829 296 899 220 659 296 743 225 863 296 939 225 685 296 775 230 899 296 981 230 713 296 809 235 939 296 1026 235 743 296 846 240 981 296 1075 240 775 296 886 245 1026 296 1127 245 809 300 929 250 1075 300 1184 250 846 305 975 255 1127 305 1244 255 886 310 1025 260 1184 310 1309 260 929 315 1078 265 1244 315 1379 265 975 320 1135 270 1309 320 1454 270 1025 325 1197 275 1379 325 1535 275 1078 330 1263 280 1454 330 1622 280 1135 335 1334 285 1535 335 1703 285 1197 340 1410 290 1622 340 1785 290 1263 345 1491 295 1703 345 1874 295 1334 350 1579 300 1785 350 1968 300 1410 355 1673 305 1874 355 2070 305 1491 360 1775 310 1968 360 2179 310 1579 365 1883 315 2070 365 2296 315 1673 370 2000 320 2179 370 2422 320 1775 375 2125 325 2296 325 1883 380 2259 330 2422 330 2000 385 2403 335 2125 340 2259 345 2403

PTLR for FNP Unit 1 Revision 4 Page 10 of 26 Table 2*2 Farley Unit 1 - 30 EFPY Cooldown Curve Data Pointsl1 ]

(adjusted to include 60 psi AP at RCS temperatures ~ 110°F and 27 psi AP at RCS temperatures < 11 0°F)12]

Steady State 20°F/hr.

4O°F/hr.

60°F/hr.

100°F/hr.

T(OF)

P (psig)

T(OF)

P (psig)

T(OF)

P (psig)

T(OF)

P (psig)

T(OF)

P (pslg) 60 0

60 0

60 0

60 0

60 0

60 533 60 492 60 450 60 407 60 319 65 536 65 495 65 453 65 410 65 323 70 540 70 499 70 457 70 414 70 326 75 543 75 502 75 461 75 418 75 331 80 548 80 507 80 465 80 422 80 335 85 552 85 511 85 469 85 427 85 340 90 557 90 516 90 474 90 432 90 346 95 561 95 521 95 480 95 438 95 352 100 561 100 527 100 486 100 444 100 358 105 561 105 533 105 492 105 450 105 365 110 547 110 506 110 466 110 424 110 340 115 553 115 513 115 473 115 432 115 348 120 561 120 521 120 481 120 440 120 357 125 561 125 529 125 489 125 449 125 367 130 561 130 538 130 499 130 459 130 378 135 561 135 548 135 509 135 469 135 389 140 561 140 558 140 519 140 480 140 402 145 561 145 561 145 531 145 492 145 415 150 561 150 561 150 543 150 506 150 430 155 561 155 561 155 557 155 520 155 446 160 561 160 561 160 561 160 535 160 463 165 561 165 561 165 561 165 551 165 481 170 561 170 561 170 561 170 561 170 501 175 561 175 561 175 561 175 561 175 523 180 561 180 561 180 561 180 561 180 546 180 706 180 673 180 641 180 609 185 571 185 725 185 693 185 662 185 631 190 598 190 745 190 715 190 685 190 655 195 628 195 767 195 738 195 709 195 681 200 659 200 790 200 762 200 735 200 709 205 693 205 815 205 789 205 763 205 739 210 730 210 842 210 818 210 794 210 771 215 770 215 871 215 848 215 827 215 806 220 813 220 902 220 881 220 862 220 844 225 859 225 936 225 917 225 900 225 884 230 909

PTLR for FNP Unit 1 Revision 4 Page 11 of 26 Table 2*2 (continued)

Farley Unit 1 - 30 EFPY Cooldown Curve Data Points[1j (adjusted to include 60 psi ~P at RCS temperatures ~ 110°F and 27 psi ~P at RCS temperatures < 11 0°F)[2)

Steady State 20°F/hr.

40°Flhr.

60°F/hr.

100°F/hr.

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig)

T (OF)

P (psig) 230 972 230 956 230 941 230 928 235 963 235 1011 235 997 235 985 235 975 240 1020 240 1053 240 1041 240 1032 240 1025 245 1079 245 1097 245 1089 245 1083 245 1079 250 1138 250 1146 250 1140 250 1138 250 1138 255 1196 255 1198 255 1196 255 1196 255 1196 260 1253 260 1253 260 1253 260 1253 260 1253 265 1313 265 1313 265 1313 265 1313 265 1313 270 1378 270 1378 270 1378 270 1378 270 1378 275 1447 275 1447 275 1447 275 1447 275 1447 280 1521 280 1521 280 1521 280 1521 280 1521 285 1601 285 1601 285 1601 285 1601 285 1601 290 1688 290 1688 290 1688 290 1688 290 1688 295 1780 295 1780 295 1780 295 1780 295 1780 300 1880 300 1880 300 1880 300 1880 300 1880 305 1987 305 1987 305 1987 305 1987 305 1987 310 2102 310 2102 310 2102 310 2102 310 2102 315 2225 315 2225 315 2225 315 2225 315 2225 320 2358 320 2358 320 2358 320 2358 320 2358

PTLR for FNP Unit 1 Revision 4 Page 12 of 26 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with 10 CFR 50, Appendix H131, 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 E185-82141. The removal schedule is provided in Table 3-1. The neutron transport and dosimetry evaluation methodologies used follow the guidance and meet the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence"151. The results of the Capsule Z examination (WCAP-16964-NP, Revision 0[61 were used to produce Figures 2-1 and 2-2.

Table 3*1 Surveillance Capsule Withdrawal Schedule (al Capsule Capsule Location (Degree)

Lead Factor Removal EFPY (bl Fluence (n/cm2 )

Y IC) 343 3.24 1.15 6.12x10 111 U IC) 107 3.34 3.08 1.73x10 111 X IC) 287 3.35 6.11 3.06 x 10 111 WIC) 110 3.01 12.43 4.75 x 1019(d)

VC) 290 3.04 20.16 7.14 X 10111(8)

Z5C) 340 3.04 24.26 8.47 x 10111ll)

Notes:

a) Data from Table 7-1, WCAP-16964-NP, Revision 0 [6]

b) Effective Full Power Years (EFPY) from plant startup.

c) Plant-specific evaluation.

d) This f1uence is not less than once or greater than twice the peak EOl f1uence for the initial40-year license term.

e) This f1uence is not less than once or greater than twice the peak EOl f1uence for a 20-year license renewal term to 60 years.

f) This f1uence is not less than once or greater than twice the peak EOl f1uence for an additional 20-year license renewal term to 80 years.

PTLR for FNP Unit 1 Revision 4 Page 13 of 26 4.0 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1.99, Revision 2[7], 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.

All six surveillance capsules 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 purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Farley Unit 1 reactor vessel surveillance data and determine if the Farley Unit 1 surveillance data is credible.

Criterion 1:

Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR 50, Fracture Toughness Requirements, December 19, 1995, to be:

the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

The Farley Unit 1 reactor vessel consists of the following beltline region materials:

Intermediate shell plates B6903-2 and B6903-3; Lower shell plates B6919-1 and B6919-2; Intermediate shell longitudinal weld seams19-894 A & B, heat number 33A277, Linde 1092 flux, flux lot 3889; Lower shell longitudinal weld seams20-894 A & B, heat number 90099, Linde 0091 flux, flux lot 3977; and Circumferential weld 11-894, heat number 6329637, Linde 0091 flux, flux lot 3999.

Per WCAP-881 0181, the Unit 1 surveillance program was based on ASTM E185-73, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vesselsl91* Per Section 4.1 of ASTM E185-73, the base metal and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime. The test material should be selected on the basis of initial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper and phosphorus) and neutron fluence.

Therefore, at the time the Farley Unit 1 surveillance capsule program was developed, lower shell plate B6919-1 was judged to be most limiting based on the above recommendations and was utilized in the surveillance program.

PTLR for FNP Unit 1 Revision 4 Page 14 of 26 The surveillance program weld for Farley Unit 1 was fabricated using the same heat of weld wire used to fabricate the middle shell axial seams19-894 A & B (heat 33A277). The results of mechanical property tests performed on the surveillance weld are considered to be representative of the property changes expected in the reactor vessel beltline seams.

Therefore, the materials selected for use in the Farley Unit 1 surveillance program were those judged to be most likely controlling with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was developed. Based on the above, the Farley Unit 1 surveillance program meets the requirements of Criterion 1.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-Ib temperature and upper shelf energy, unambiguously.

Plots of Charpy energy versus temperature for the unirradiated condition are presented in the Unit 1 reactor vessel surveillance program description contained in WCAP-881 0[8l.

Plots of Charpy energy versus temperature for the irradiated conditions are presented in the reactor vessel surveillance capsule reports for capsules y[lOl, U [111, X [121, W [131, V [14], and Z [61.

Based on engineering jUdgment, the scatter in the data presented in these plots is small enough to determine the 30 ft-Ib temperature and upper shelf energy of the Farley Unit 1 surveillance materials unambiguously. Therefore, the Farley Unit 1 surveillance program meets the requirements of Criterion

2.

Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of

~RTNOT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17°F for base metal. Even if the f1uence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82.

The least squares method, as described in Regulatory Position 2.1, will be utilized in determining a best-fit line for this data to determine if this criterion is met.

PTLR for FNP Unit 1 Revision 4 Page 15 of 26 Table 4-1 Surveillance Capsule Data Calculation of Best-Fit Line as Described in Position 2.1 of Regulatory Guide 1.99, Revision 2 (e)

Material Capsule f (a)

FF (b)

(X)

ARTNDT (y)

FF *ARTNDT (xy)

FF2 (X2)

Y 0.612 0.862 64.6 55.7 0.744 Lower Shell Plate 86919-1 (Longitudinal)

U X

W 1.73 3.06 4.75 1.151 1.295 1.392 110.0 129.2 145.3 126.6 167.1 202.3 1.324 1.678 1.938 V

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 Lower Shell Plate 86919-1 (Transverse)

U X

W 1.73 3.06 4.75 1.151 1.295 1.392 100.4 110.8 150.5 115.5 143.5 209.5 1.324 1.678 1.938 V

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 CFB691 9-1 = ~(FF * ~RTNDT) + ~(FF2) = (2146.20) + (20.118) =106.7°F Y

0.612 0.862 66.9 57.7 0.744 U

1.73 1.151 75.1 86.4 1.324 Weld Metal X

W 3.06 4.75 1.295 1.392 87.4 98.3 113.2 136.8 1.678 1.938 V

7.14 1.466 117.5 172.3 2.149 Z

8.47 1.492 113.5 169.3 2.225 SUM:

735.57 10.059 CFsUIV.Weld = ~(FF * ~RTNDT) + ~(FF2) = (735.57) + (10.059) = 73.1 OF NOTES:

(a) f = Fluence (1019 n/cm2, E > 1.0 MeV).

(b) FF = Fluence Factor = F(O.28-0.1Iog(f).

(c) Data from Table 0-1 ofWCAP-16964-NP, Revision 0[61.

PTLR for FNP Unit 1 Revision 4 Page 16 of 26 Table 4-2 Scatter of ARTNOT Values about a Best-Fit Line for Surveillance Plate Material (a)

Lower Shell Plate B6919-1 Orientation Capsule CF (Best Fit Slope)

FF ARTNoT (30 ft-Ib)

(OF)

Best Fit ARTNoT (OF)

Scatter of ARTNDT (OF)

Y 106.7 0.862 64.6 92.0 27.4 U

106.7 1.151 110.0 122.8 12.8 X

106.7 1.298 129.2 138.2 9.0 Longitudinal W

106.7 1.392 145.3 148.5 3.2 V

106.7 1.466 177.7 156.4 21.3 Z

106.7 1.492 202.2 159.1 43.1 Y

106.7 0.862 70.1 92.0 21.9 U

106.7 1.151 100.4 122.8 22.4 X

106.7 1.298 110.8 138.2 27.4 Transverse W

106.7 1.392 150.5 148.5 2.0 V

106.7 1.466 161.7 156.4 5.3 Z

106.7 1.492 178.3 159.1 19.2 NOTE:

(a) Data from Table 0-2 ofWCAP-16964-NP, Revision 0161*

The scatter of ARTNOT values about a best-fit line drawn with the y-intercept equal to zero, as described in Regulatory Position 2.1, should be less than 17°F for base metal. As shown above, the scatter of seven of the data points is not within 17°F of the best-fit line. Therefore, Criterion 3 is not met for the Farley Unit 1 surveillance plate material. Since not all of the data is within 17°F of the best fit line. SNC has chosen to use the CF from this surveillance data along with a a.... of 17°F when predicting the Farley Unit 1 vessel properties.

Table 4-3 Scatter of ARTNOT Values about a Best-Fit Line for Surveillance Weld Material (a)

Material Capsule CF (Best Fit Slope)

FF ARTNDT (30 ft-Ib)

(OF)

Best Fit ARTNDT (OF)

Scatter of ARTNDT (OF)

Y 73.1 0.862 66.9 63.1 3.8 U

73.1 1.151 75.1 84.3 9.1 X

73.1 1.298 87.4 94.6 7.4 Weld Metal W

73.1 1.392 98.3 101.9 3.5 V

73.1 1.466 117.5 107.3 10.3 Z

73.1 1.492 113.5 109.2 4.4 NOTE:

(a) Data from Table 0-2 ofWCAP-16964-NP, Revision 0161*

PTLR for FNP Unit 1 Revision 4 Page 17 of 26 The scatter of ~RTNOT values about a best-fit line drawn with the y-intercept equal to zero, as described in Regulatory Position 2.1, is less than 28°F as shown above. Therefore, Criterion 3 is met for the Farley Unit 1 surveillance weld material.

Criterion 4:

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

The Farley Unit 1 capsule specimens are located in the reactor between the neutron shielding pads and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the neutron shielding pads. 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 and will not differ by more than 25°F. Therefore, the Farley surveillance program meets the requirements of Criterion 4.

Criterion 5:

The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.

The Farley Unit 1 surveillance program does not include correlation monitor material. Therefore, Criterion 5 is not applicable to Farley Unit 1.

CONCLUSION:

Based on the preceding responses to the criteria of RegUlatory Guide 1.99, Revision 2, section B, and the application of engineering jUdgment, the Farley Unit 1 surveillance plate material data is not credible and the Farley Unit 1 surveillance weld data is credible.

5.0 Supplemental Data Tables Table 5-1 contains a comparison of measured surveillance material 30 ft-Ib 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 f1uences used in the PTS evaluation.

Table 5-5 provides a summary of the adjusted reference temperatures (ARTs) of the Farley Unit 1 reactor vessel beltline materials at the 1/4-T and 3/4-T locations for 30 EFPY.

Table 5-6 shows the calculation of the ART at 30 EFPY for the limiting Farley Unit 1 reactor vessel material (lower shell plate B6919-1).

Table 5-7 provides RTPTS values for Farley Unit 1 for 36 EFPY.

PTLR for FNP Unit 1 Revision 4 Page 18 of 26 Table 5*1 Comparison of Surveillance Material 30 ft-Ib Transition Temperature Shift and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions (a) 30 ft-Ib Transition Upper Shelf Energy Temperature Shift Decrease Fluence Predicted Measured Predicted Measured Material Capsule (1019 n/cm2, (OF)

(OF)

(0/0)

(0/0)

E> 1.0 MeV) y 7

0.612 84.3 64.6 21 1.73 112.5 110.0 27 21 U

Lower Shell X

3.06 126.7 129.2 30 17 Plate 86919-1 W

4.75 136.2 145.7 35 20 (Longitudinal)

V 7.14 177.7 21 39 143.4 Z

8.47 40 145.9 202.2 29 Y

0.612 84.3 70.1 1

21 1.73 112.5 27 U

100.4 9

Lower Shell 12 X

3.06 126.7 110.8 30 Plate 86919-1 W

4.75 150.5 17 136.2 35 (Transverse)

V 7.14 143.4 161.7 21 39 Z

145.9 178.3 40 8.47 23 Y

0.612 67.4 66.9 24 3

1.73 75.1 31 22 U

89.9 Surveillance 3.06 101.2 87.4 15 X

36 Program W

4.75 40 18 108.7 98.3 Weld Metal 7.14 117.5 44 18 V

114.5 46 Z

8.47 116.5 25 113.5 Y

0.612 29.2 7

21 1.73 U

155.3 15 X

3.06 132.9 Heat Affected Zone Material W

4.75 121.7 11 7.14 V

169.7 20 Z

170.7 20 8.47 NOTE:

(a)

Data from Table 5-10 ofWCAP-16964-NP, Revision 016l.

PTLR for FNP Unit 1 Revision 4 Page 19 of 26 Table 5*2 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f (a)

FF(b)

ARTNDT(C)

FF*ARTNDT FF2 y

0.612 0.862 64.6 55.7 0.744 Lower Shell Plate B6919-1 (Longitudinal)

U X

W V

1.73 3.06 4.75 7.14 1.151 1.295 1.392 1.466 110.0 129.2 145.3 177.7 126.6 167.1 202.3 260.5 1.324 1.678 1.938 2.149 Z

8.47 1.492 202.2 301.6 2.225 Y

0.612 0.862 70.1 60.5 0.744 Lower Shell Plate B6919-1 (Transverse)

U X

W V

1.73 3.06 4.75 7.14 1.151 1.295 1.392 1.466 100.4 110.8 150.5 161.7 115.5 143.5 209.5 237.1 1.324 1.678 1.938 2.149 Z

8.47 1.492 178.3 266.0 2.225 SUM:

2146.20 20.118 CFB6919-1 = I(FF

• RTNOT) + I( FF2) = (2146.20) + (20.118) = 106.7°F Y

0.612 0.862 108.4 (66.9)(d) 93.5 0.744 Surveillance Weld Material U

X W

V 1.73 3.06 4.75 7.14 1.151 1.295 1.392 1.466 121.7 (75.1 )(d) 141.6 (87.4)(d) 159.2 (98.3)(d) 190.4 (117.5)(d) 140.0 183.4 221.7 279.1 1.324 1.678 1.938 2.149 Z

8.47 1.492 183.9 (113.5)(d) 274.3 2.225 SUM:

1191.95 10.059 CF Surv. Weld = I(FF

• RTNOT) + I( FF 2

) = (1191.95) + (10.059) = 118.5°F NOTES:

(a) f =fJuence (1019 n/cm2, E> 1.0 MeV); from Table 4-1.

(b) FF =fJuence factor = f(0.28-0.1I09(I>>.

(c) ARTNOT values from Table 5-1.

(d) To calculate the best fit chemistry factor (CF) as provided by Reg. Guide 1.99 Rev. 2, Position 2.1, the surveillance weld aRTNOT values have been adjusted by a ratio of 1.62 (CFvesselweld + CFsurvweld = 126.3 +

78.1 = 1.62). Pre-adjusted values are in parentheses.

PTLR for FNP Unit 1 Revision 4 Page 20 of 26 Table 5*3 Reactor Vessel Toughness Table (Unirradiated) (a)

Beltline Material Cu Weight Ni Weight IRTNDT (OF)

Closure Head Flange

_50(d)

Vessel Flange 60 Intermediate Shell Plate B6903-2 0.13 0.60 0

Intermediate Shell Plate B6903-3 0.12 0.56 10 Lower Shell Plate B6919-1 0.14 0.55 15 Lower Shell Plate B6919-2 0.14 0.56 5

Intermediate Shell Longitudinal Weld Seams19-894 A &B (Heat # 33A277) (b) 0.258 0.165

-56(9)

Surveillance Weld (c) 0.14 0.19 Circumferential Weld Seam 11-894 (Heat # 6329637) (b) 0.205 0.105

-56(9)

Lower Shell Longitudinal Weld Seams20-894 A & B (Heat # 90099) (b) 0.197 0.060

-56(9)

Nozzle Shell Forging B6914 0.16(8) 0.684(f) 30(h)

Nozzle Shell to Intermediate Shell Circumferential Weld Seam 10-894 (Heat # 90099) (b) 0.197 0.060

-56(9)

NOTES:

(a) From Table 1 ofWCAP-14689, Revision 4 [18J.

(b) Best-estimate copper and nickel from CE NPSO-1039 [19].

(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 in WCAP-8810, Revision 0 [81.

(d) Replacement closure head initial RTNOT value was taken from MHI_SNC-Q195[231*

(e) MISC-PENG-ER-Q11[24] provides chemistry measurements for the upper shell forging, none of which contain Cu wt%

values. Regulatory Guide 1.99, Revision 2, Section 1.1 indicates a standard deviations analysis may be used to determine a generic value of wt% Cu when no information is otherwise available. This estimated (generic) wt% Cu value and the corresponding standard deviations analysis are discussed in Appendix G of ORNLfTM-2006/530[25],

which tabulates the mean values, standard deViation, and maximum values for the various material specifications.

broken out as both low Cu << 0.072 wt%) and high Cu (> 0.072 wt%). For A508 Class 2 forging material, there were only a limited number of heats (24) available for analysis relative to the more common A533 Grade B Class 1 vessel material, which had a total of 73 heats for analysis. Since there were fewer heats available, a slightly more conservative approach (than the approach that utilizes the mean + one standard deviation) is prudent to estimate the generic Cu content of A508 Class 2 forging material. The maximum copper content measured in ORNLlTM 2006/530[25] for A508 Class 2 material is 0.16 wt%, and is utilized for the Fariey Unit 1 upper shell forging.

(f) Best-estimate nickel determined based on the five chemistry measurements provided in MISC-PENG-ER-011[24j for the upper shell forging; Ni wt% values reported were 0.70, 0.68. 0.68, 0.68, and 0.68 for a mean value of 0.684.

(g) All weld initial RTNOT values are generic and taken from 10 CFR 50.61 paragraph (c)(1)(ii) of the 1-1-07 edition.

(h) The upper shell forging NOn of 30°F is a bounding value for determining RTNOT considering the Charpy V-Notch data in MISC-PENG-ER-011[24], and the tabulated values of CVN energy and lateral expansion at 90°F (= NOn + 60°F) reduced to 65% of their measured values in the strong orientation of the specimen (per the guidance in U.S. NRC Branch Technical Position MTEB 5-2, Paragraph B-1.1(3)[261.

PTLR for FNP Unit 1 Revision 4 Page 21 of 26 Table 5-4 Reactor Vessel Fluence Projections Used in PTS Evaluation for 36 EFPY (a,b)

(1019 n/cm2, E> 1.0 MeV)

EFPY 36 4.14 2.41 1.73 1.24 NOTES:

(a) From Table 6-2 of WCAP-16221-NP, Revision 0[141, (b) These f1uence projections remain bounding with respect to the updated 36 EFPY f1uence projections in Table 6-2, WCAP-16964-NP, Revision 016].

PTLR for FNP Unit 1 Revision 4 Page 22 of 26 Table 5-5 Summary of Adjusted Reference Temperatures (ARTs) for Reactor Vessel Beltline Materials at the 1/4T and 3/4T Locations for 30 EFPY (a)

Material 1/4 T (OF) 3/4 T (OF)

Intermediate Shell Plate 86903-2 144 120 Intermediate Shell Plate 86903-3 143 122 Lower Shell Plate 86919-1 167 141 Lower Shell Plate 86919-1 Using SIC Data 177(d) 150(d)

Lower Shell Plate 86919-2 157 132 Intermediate Shell Longitudinal Weld Seams19-894 A &8 (Heat # 33A277) 120(b) 88(b)

Intermediate Shell Longitudinal Weld Seams19-894 A & 8 (Heat # 33A277)

Using SIC Data 92(b) 62(b)

Circumferential Weld Seam 11-894 (Heat # 6329637) 128 102 Lower Shell Longitudinal Weld Seams20-894 A & 8 (Heat # 90099) 90(b) 67(b)

Nozzle Shell Forging 86914 148(C) 121(C)

Nozzle Shell to Intermediate Shell Circumferential Weld Seam 10-894 (Heat # 90099) 74(C) 42(C)

NOTES:

(a) The ARTs presented here are based on the peak reactor vessel surface f1uence of 3.39 x 1019 n/cm2 (E > 1.0 MeV) from Table 6-2 of WCAP-16964-NP, Revision 0[6], unless otherwise noted.

(b) 8ased on reactor vessel clad/base metal interface 45° f1uence value of 1.03 x 1019 n/cm2 (E > 1.0 MeV) from Table 6-2 ofWCAP-16964-NP, Revision 0161.

(c) 8ased on Nozzle to Intermediate Shell Circumferential Weld 0° f1uence value of 0.542 x 1019 n/cm2 (E > 1.0 MeV) from Table 6-3 of WCAP-16964-NP, Revision 0161.

(d) Limiting 1/4T and 3/4T ART values. The PIT limit curves are based on these limiting ART values of 177°F and 150°F.

PTLR for FNP Unit 1 Revision 4 Page 23 of 26 Table 5-6 Calculation of Adjusted Reference Temperature at 30 EFPY for the Limiting Reactor Vessel Material - Lower Shell Plate 86919-1 Parameter Value Operating Period 30 EFPY Location 1/4 T 3/4 T Chemistry Factor, CF (OF) (e) 106.7 106.7 Fluence, f (1019 n/cm2) (b) 2.11 0.82 Fluence Factor, FF =f (0.28-0.1"109(1))

1.20 0.94 dRTNOT =CF x FF (OF) 128.4 100.8 Initial RTNOT, I (OF) (e) 15 15 Margin, M (OF) (d) 34 34 Adjusted Reference Temperature (ART), (OF) per Regulatory Guide 1.99, Revision 2 (e) 177 150 NOTES:

(a) Chemistry factor is taken from Table 4-1.

(b) Fluence is based on fsurf =3.39 X 1019 n/cm2 (E > 1.0 MeV), from Table 6-2 of WCAP-16964-NP, Revision 0[61* Farley Unit 1 reactor vessel wall thickness is 7.875 inches in the beltline region.

(c) Initial RTNOT value is taken from Table 5-3.

(d) Margin =2..Ju/ +U6?,(OF); for the lower shell plate 86919-1, aj= O°F and ali = 17°F.

(e) Per Regulatory Guide 1.99, Revision 2: ART (OF) =aRTNOT + I + M.

PTLR for FNP Unit 1 Revision 4 Page 24 of 26 Table 5-7 Pressurized Thermal Shock (RTPTs) Values for 36 EFPY (a.c)

Material CF Surface Fluence(c)

(1019 n/cm2,

E> 1.0 MeV)

FF ARTNDT (CF x FF)

(OF)

I (OF)

M (OF)

RTpTS (OF)

Intermediate Shell Plate 86903-2 91.0 4.14 1.36 123.8 0

34 158 Intermediate Shell Plate 86903-3 82.2 4.14 1.36 111.8 10 34 156 Lower Shell Plate 86919-1 97.8 4.14 1.36 133.0 15 34 182 Lower Shell Plate 86919-1 Using SIC Data 100.9 4.14 1.36 137.2 15 34 186(b)

Lower Shell Plate 86919-2 98.2 4.14 1.36 133.6 5

34 173 Intermediate Shell Longitudinal Welds19-894 A & 8 (Heat # 33A277) 126.3 1.24 1.06 133.9

-56 66 144 Intermediate Shell Longitudinal Welds19-894 A & 8 (Heat # 33A277)

Using SIC Data 124.6 1.24 1.06 132.1

-56 44 120 Circumferential Weld 11-894 (Heat # 6329637) 98.4 4.14 1.36 133.8

-56 66 144 Lower Shell Longitudinal Welds20-894 A &8 (Heat # 90099) 91.4 1.24 1.06 96.9

-56 66 107 NOTES:

(a) From Westinghouse letter ALA-04-170[15l.

(b) This RTPTS value was calculated using the CF from the surveillance data and a full all. margin of 17°F, since this surveillance data is not credible.

(c) These f1uence projections remain bounding with respect to the updated 36 EFPY f1uence projections in Table 6-2, WCAP-16964-NP, Revision 0161*

PTLR for FNP Unit 1 Revision 4 Page 25 of 26 6.0 References

1. ALA-08-68, "Farley Unit 1 Heatup I Cooldown Pressure-Temperature Limit Curves to 30 EFPY,"

G. Gontis, October 1, 2008.

2. 10 CFR 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19,1995.

3. 10 CFR 50 Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.

4. ASTM E185-82, "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.

5. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

6. WCAP-16964-NP, Revision 0, "Analysis of Capsule Z from the Southern Nuclear Operating Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program,"

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

7. Regulator Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

May 1988.

8. WCAP-8810, Revision 0, "Southern Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program," J. A. Davidson, et. aI.,

December 1976.

9. ASTM E185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials.

10. WCAP-9717, Revision 0, "Analysis of Capsule Y from the Alabama Power Company Farley Unit NO.1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, et. aI., June 1980.

11. WCAP-10474, 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. aI.,

February 1984.

12. WCAP-11563, 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. aI.,

September 1987.

13. WCAP-14196, Revision O,"Analysis of Capsule W from the Alabama Power Company Farley Unit 1 Reactor Vessel Radiation Surveillance Program," P. A. Peter, et. aI., February 1995.

14. WCAP-16221-NP, Revision 0, "Analysis of Capsule V from the Southern Nuclear Operating Company, Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, K. G. Knight, et. aI., March 2004.

15. Westinghouse letter ALA-04-170, "Review of Revision 2 Draft of Pressure-Temperature Limits Report," E. C. Arnold to L. M. Stinson, October 25,2004.

16. SNC letter NL-04-0372, "Reactor Material Surveillance Program Specimen Capsule Withdrawal Schedule Revisions - Additional Information," March 5, 2004.

17. NRC letter (SNC LC #14001), "Joseph M. Farley Nuclear Plant, Units 1 and 2 Re: Specimen Capsule Withdrawal Schedule Revisions," March 15,2004.

PTLR for FNP Unit 1 Revision 4 Page 26 of 26

18. WCAP-14689, Revision 4, "Farley Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," E. Terek, April 1998.

19. CE NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," Combustion Engineering Owners Group, June 1997.

20. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.

21. NRC letter (SNC LC #11969), "Joseph M. Farley Nuclear Plant, Units 1 and 2 Acceptance for Referencing of Pressure Temperature Limits Report," March 31, 1998.

22. NRC letter (SNC LC #11972), "Joseph M. Farley Nuclear Plant, Units 1 and 2 Correction to Acceptance Letter for Referencing of the Pressure Temperature Limits Report," April 3, 1998.

23. Mitsubishi Heavy Industries, LTD, Kobe Shipyard & Machinery Works (MHI). MHI-SNC-0195, Reactor Vessel Closure Head for Farley-1, "Certified Material Test Report," August 22, 2003.

24. MISC-PENG-ER-011, Revision 00, "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the Farley Unit 1 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," Combustion Engineering Owner's Group, October 1995.

25. ORNUfM-2006/530, fAA Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels," Oak Ridge National Laboratory, November 2007.

26. NUREG-0800, Revision 1, Section 5.3.2, Branch Technical Position MTEB 5-2 "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., July 1981.