ML032750349

To GE-NE-0000-0013-3193-01a-R1, Pressure-Temperature Curves for TVA Browns Ferry, Unit 2

ML032750349
Person / Time
Site:	Browns Ferry
Issue date:	08/31/2003
From:	Branlund B, Frew B, Tilly L General Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2005-0108 GE-NE-0000-0013-3193-01a, Rev 1
Download: ML032750349 (135)

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GE Nuclear Energy Engineering and Technology General Electric Company 175 Curtner Avenue GE-NE-0000-0013-3193-01a-RI Revision 1 Class I San Jose, CA 95125 August 2003 Non-Propretary Version Pressure-Temperature Curves For TVA Browns Ferry Unit 2 Prepared by:

Verified by:

Approved by:

L lAf-y L.J. Tilly, Senior Engineer Structural Mechanics and Materials (3D (Frew B.D. Frew, Principal Engineer Structural Mechanics and Materials (B7 cBranrund B.J. Branlund, Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version REPORT REVISION STATUS Revision Purpose I

Proprietary notations have been updated to meet current requirements. Figure 4-2 and its associated discussion have been revised. There is no impact to the P-T curves as a result of I ___________ this change.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between TVA and GE, PO O00001704, PVT Curves and Flaw Handbooks, effective 3/25/03, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than TVA, or for any purpose other than that for which it is furnished by GE, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

Copyright, General Electric Company, 2003

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version EXECUTIVE

SUMMARY

This report provides the pressure-temperature curves (P-T curves) developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 1998 [1]; the P-T curves in this report represent 23 and 30 effective full power years (EFPY), where 30 EFPY represents the end of the 40 year license, and 23 EFPY is provided as a midpoint between the current EFPY and 30 EFPY. The P-T curve methodology includes the following: 1) the use of K1c from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in accordance with 10CFR50.55a (4], in effect at the time of this evaluation.

This report incorporates a fluencecalculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190.

This fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MWt.

CONCLUSIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B & C)

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 100F/hr or less for which the curves are applicable.

However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 150F/hr or less must be maintained at all times.

The P-T curves apply for both heatup and cooldown and for both the 114T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kin, at 114T to be less than that at 3/4T for a given metal temperature.

Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 23 and 30 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beltline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beltline (at 23 and 30 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

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GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version TABLE OF CONTENTS

1.0 INTRODUCTION

2.0 SCOPE OF THE ANALYSIS 3.0 ANALYSIS ASSUMPTIONS 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY

5.0 CONCLUSION

S AND RECOMMENDATIONS

6.0 REFERENCES

1 3

5 6

6 14 20 50 73

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE OF APPENDICES APPENDIX A APPENDIX B APPENDIX C APPENDIX D APPENDIX E APPENDIX F DESCRIPTION OF DISCONTINUITIES PRESSURE-TEMPERATURE CURVE DATA TABULATION OPERATING AND TEMPERATURE MONITORING REQUIREMENTS GE SIL 430 DETERMINATION OF BELTUNE REGION AND IMPACT ON FRACTURE TOUGHNESS EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE BROWNS FERRY UNIT 2 RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2.

CR1) PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 31 FIGURE 4-3.

FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 37 FIGURE 5-1: BOTIOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] - 23 EFPY [150F/HR OR LESS COOLANT HEATUP/COOLDOWN]

53 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE Al - 23 EFPY [15 0F/HR OR LESS COOLANT HEATUP/COOLDOWN]

54 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 23 EFPY [15 0F/HR OR LESS COOLANT HEATUP/COOLDOWN]

55 FIGURE 5-4: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 23 EFPY 115 0F/HR OR LESS COOLANT HEATUP/COOLDOWN]

56 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 23 EFPY

[100F/HR OR LESS COOLANT HEATUP/COOLDOWN]

57 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 23 EFPY 1100-F/HR OR LESS COOLANT HEATUP/COOLDOWNJ 58 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 23 EFPY

[100IF/HR OR LESS COOLANT HEATUP/COOLDOWN]

59 FIGURE 5-8: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 23 EFPY 11000FIHR OR LESS COOLANT HEATUP/COOLDOWN]

60 FIGURE 5-9: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 23 EFPY [1000F/HR OR LESS COOLANT HEATUP/COOLDOWN]

61 FIGURE 5-10: COMPOSITE CORE NOT CRITICAL [CURVE B] INCLUDING BOTTOM HEAD AND CORE CRITICAL P-T CURVES [CURVE C] UP TO 23 EFPY [100IFIHR OR LESS COOLANT HEATUP/COOLDOWN]

62 FIGURE 5-11: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE Al - 30 EFPY 1150F/HR OR LESS COOLANT HEATUP/COOLDOWNJ 63 FIGURE 5-12: UPPER VESSEL P-T CURVE FOR PRESSURE TEST tCURVE Aj - 30 EFPY [15OF/HR OR LESS COOLANT HEATUP/COOLDOWN]

64 FIGURE 5-13: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 30 EFPY 1150F/HR OR LESS COOLANT HEATUP/COOLDOWN]

65 FIGURE 5-14: COMPOSITE PRESSURE TEST P-T CURVES [CURVE Al UP TO 30 EFPY [150F/HR OR LESS COOLANT HEATUP/COOLDOWN]

66

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version FIGURE 5-15: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE BJ - 30 EFPY 11000F/HR OR LESS COOLANT HEATUP/COOLDOWN]

67 FIGURE 5-16: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] -30 EFFY 1100-F/HR OR LESS COOLANT HEATUP/COOLDOWN]

68 FIGURE 5-17: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 30 EFPY

[1000F/HR OR LESS COOLANT HEATUP/COOLDOWN]

69 FIGURE 5-18: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE L] UP TO 30 EFPY 1100F/1HR OR LESS COOLANT HEATUP/COOLDOWNJ 70 FIGURE 5-19: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 30 EFPY [1000F/HR OR LESS COOLANT HEATUP/COOLDOWN]

71 FIGURE 5-20: COMPOSITE CORE NOT CRITICAL [CURVE BJ INCLUDING BOTTOM HEAD AND CORE CRITICAL P-T CURVES [CURVE C] UP TO 30 EFPY I100IFIHR OR LESS COOLANT HEATUP/COOLDOWN]

72

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Propfietary Version TABLE OF TABLES TABLE 4-1: RTz VALUES FOR BROWNS FERRY UNIT 2 VESSEL MATERIALS 11 TABLE 4-2: RT4u VALUES FOR BROWNS FERRY UNIT 2 NOZZLE AND WELD MATERIALS 12 TABLE 4-3: RTr VALUES FOR BROWNS FERRY UNIT 2 APPURTENANCE AND BOLTING MATERIALS 13 TABLE 4-4: BROWNS FERRY UNIT 2 BELTLINE ART VALUES (23 EFPY) 18 TABLE 4-5: BROWNS FERRY UNIT 2 BELTLINE ART VALUES (30 EFPY) 19 TABLE 4-6:

SUMMARY

OF THE 10CFR50 APPENDIX G REQUIREMENTS 22 TABLE 4-7: APPLICABLE BWRI4 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 24 TABLE 4-8: APPLICABLE BWRA4 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 24 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 52

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version

1.0 INTRODUCTION

The pressure-temperature (P-T) curves included in this report have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline.

Complete P-T curves were developed for 23 and 30 effective full power years (EFPY),

where 30 EFPY represents the end of the 40 year license, and 23 EFPY is provided as a midpoint between the current EFPY and 30 EFPY. The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. This report incorporates a fluence calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190. This fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MW,.

The methodology used to generate the P-T curves in this report is presented in Section 4.3 and is similar to the methodology used to generate the P-T curves in 1998 [1]. The P-T curve methodology Includes the following: 1) the use of Kc from Figure A-4200-1 of Appendix A [17] to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in accordance with IOCFR50.55a [4], in effect at the time of this evaluation. P-T curves are developed using geometry of the RPV shells and discontinuities, the initial RTNDT of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.

The initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. The Charpy energy data used to determine the initial RTNDT values are tabulated from the Certified Material Test Report (CMTRs). The data and methodology used to determine initial RTNDT are documented in Section 4.1.

Adjusted Reference Temperature (ART) is the reference temperature when including irradiation shift and a margin term.

Regulatory Guide 1.99, Rev. 2 [7] provides the GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 23 and 30 EFPY are included in Section 4.2.

The peak ID fluence values of 1.Ox1018n/cm2 (23 EFPY) and 1.33x1018n/cm2 (30 EFPY) used in this report are discussed in Section 4.2.1.2. Beltline chemistry values are discussed in Section 4.2.1.1.

Comprehensive documentation of the RPV discontinuities that are considered in this report is included in Appendix A. This appendix also includes a table that documents which non-beltline discontinuity curves are used to protect each discontinuity.

Guidelines and requirements for operating and temperature monitoring are included in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D.

Appendix E demonstrates that all reactor vessel nozzles requiring fracture toughness evaluation are outside the beltline region. Finally, Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Verslon 2.0 SCOPE OF THE ANALYSIS The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 1998 [1]. A detailed description of the P-T curve bases is included in Section 4.3. The P-T curve methodology includes the following: 1) the use of lc from Figure A-420O-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in accordance with 10CFR50.55a [4], in effect at the time of this evaluation. Other features presented are:

Generation of separate curves for the upper vessel in addition to those generated for the beltline, and bottom head.

Comprehensive description of discontinuities used to develop the non-beltline curves (see Appendix A).

The pressure-temperature (P-T) curves are established to the requirements of 10CFR50, Appendix G [8] to assure that brittle fracture of the reactor vessel is prevented. Part of the analysis involved in developing the P-T curves is to account for irradiation embrittlement effects in the core region, or beltline. The method used to account for irradiation embrittlement is described in Regulatory Guide 1.99, Rev. 2 [7].

In addition to beltline considerations, there are non-beltline discontinuity limits such as nozzles, penetrations, and flanges that influence the construction of P-T curves. The non-beltline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable Browns Ferry Unit 2 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].

Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize heating requirements prior to pressure testing. Operating and temperature monitoring requirements are found In Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version Appendix D. Appendix E demonstrates that all reactor vessel nozzles requiring fracture toughness evaluation are outside the beitline region. Finally, Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:

The hydrostatic pressure test will be conducted at or below 1064 psig; the evaluation conservatively uses this maximum pressure.

The shutdown margin, provided in the Definitions Section of the Browns Ferry Unit 2 Technical Specification [51, is calculated for a water temperature of 680F.

The fluence is conservatively calculated using an EPU flux for the entire plant life. The flux Is calculated In accordance with Regulatory Guide 1.190.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits.

The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 and are summarized as follows:

a.

Test specimens shall be longitudinally oriented CVN specimens.

b.

At the qualification test temperature (specified in the vessel purchase specification), no impact test result shall be less than 25 ft-lb, and the average of three test results shall be at least 30 ft-lb.

c.

Pressure tests shall be conducted at a temperature at least 600F above the qualification test temperature for the vessel materials.

The current Nequirements used to establish an initial RTNDT value are significantly different. For plants constructed according to the ASME Code after Summer 1972, the requirements per the ASME Code Section 1I1, Subsection NB-2300 are as follows:

a.

Test specimens shall be transversely oriented (normal to the rolling direction) CVN specimens.

b.

RTMwT is defined as the higher of the dropweight NDT or 60F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion are met.

c.

Bolt-up in preparation for a pressure test or normal operation shall be performed at or above the highest RTNDT of the materials in the closure flange region or lowest service temperature (LST) of the bolting material, whichever is greater.

IOCFR50 Appendix G [81 states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented in an approved manner. GE developed methods for analytically GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Ri Non-Proprietary Version converting fracture toughness data for vessels constructed before 1972 to comply with current requirements.

These methods were developed from data in WRC Bulletin 217 [19] and from data collected to respond to NRC questions on FSAR submittals in the late 1970s.

In 1994, these methods of estimating RTNDT were submitted for generic approval by the BWR Owners' Group 110], and approved by the NRC for generic use 111].

4.1.2 Values of Initial RTNDT and Lowest Service Temperature (LST)

To establish the initial RTNDT temperatures for the Browns Ferry Unit 2 vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT.

Example RTNDT calculations for vessel plate, forging, and for bolting material LST are summarized in the remainder of this section.

The RTNDT values for the vessel weld materials were not calculated; these values were obtained from [13] (see Table 4-2).

For vessel plate material, the first step in calculating RTNDT is to establish the 50 ft-lb transverse test temperature from longitudinal test specimen data (obtained from certified material test reports, CMTRs [121). For Browns Ferry Unit 2 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 20F per ft-lb energy difference from 50 ft-lb.

For example, for the Browns Ferry Unit 2 beltline plate heat C2460-2 in the lower shell course; the lowest Charpy energy and test temperature from the CMTRs is 40 ft-lb at 100F. The estimated 50 ft-lb longitudinal test temperature is:

TsoL = 100F + [ (50 - 40) ft-lb

• 20F/ft-lb ] = 300F The transition from longitudinal data to transverse data Is made by adding 300F to the 50 ft-lb longitudinal test temperature; thus, for this case above, T50T = 300F + 300F = 600F.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version The initial RTN= is the greater of nil-ductility transition temperature (NDT) or (T6or 600F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is -20oF. Thus, the initial RTNDT for plate heat C2460-2 is 00F.

For the vessel HAZ material, the RTNDT is assumed to be the same as for the base material, since ASME Code weld procedure qualification test requirements and post-weld heat treat data indicate this assumption is valid.

For vessel forging material, such as nozzles and closure flanges, the method for establishing RTNDT is the same as for vessel plate material. For the recirculation inlet nozzle at Browns Ferry Unit 2 (Heat E25VW\\,

the NDT is 400F and the lowest CVN data Is 63 ft-lb at 40°F. The corresponding value of (Tsor 600F) is:

(T6OT

- 600F) = [ 40 + 30 ]-600 F = 100F.

Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (Twr 600F), which is 400F.

In the bottom head region of the vessel, the vessel plate method Is applied for estimating RTNDT. For the bottom head dollar plate heat of Browns Ferry Unit 2, Heat C2669-2, the NDT is 400F and the lowest CVN data was 34 ft-lb at 400F. The corresponding value of (T5OT - 600F) was:

(T5ST - 600F) = { l40 + (50 - 34) ft-lb

• 20F/ft-lb ] + 300F }-600F = 420F.

Therefore, the initial RTmT was 420F.

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 600F is the LST for the bolting materials. All Charpy data for the Browns Ferry Unit 2 closure studs did not meet the 45 ft-lb, 25 MLE requirements at 100F. Therefore, the LST for the bolting material is 700F. The highest GE Nuclear Energy GE-NE-0000-0013-3193-Ole-Rl Non-Proprietary Version RTNDT in the closure flange region is 23.10F, for the vertical electroslag weld material in the upper shell. Thus, the higher of the LST and the RTNDT +600F is 83.10F, the bolt-up limit in the closure flange region.

The initial RTNDT values for the Browns Ferry Unit 2 reactor vessel (refer to Figure 4-1 for the Browns Ferry Unit 2 Schematic) materials are listed in Tables 4-1, 4-2, and 4-3. This tabulation includes beltline, closure flange, feedwater nozzle, and bottom head materials that are considered in generating the P-T curves.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version TOP HEAD TOP HEAD FLANGE SH-ELL FLANGE SHELLCOURSE 95 y,.*77r7,77 '77 7/7 ZM=

.7/77.

y77 SHELL COURSE #4

.. A7

/-7-W.......=

SHELL COURSE #3

)

i)

) i TOP OF a; SHELL COUSE2 ACTIVE FUELSHLCORE2 (TAF) 3~~~.3 AXIAL WELDS (TAF)366.3-l

/{ESW) R GIRTH WELD BOTTOM OF SHELLCOURSE #1 ACTIVE FUEL (RAF} 216i3' s BOTTOM HEAD SUPPORT SKIRT Notes: (1) Refer to Tables 4-1, 4-2, and 4-3 for reactor vessel components and their heat identifications.

(2) See Appendix E for the definition of the beliline region.

Figure 4-1: Schematic of the Browns Ferry Unit 2 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-NE-0000-0013-3193-Ole-RI Non-Proprietary Version Table 4-1: RTNDT Values for Browns Ferry Unit 2 Vessel Materials Top Head & Flange Shell Flange (MK 48) 48-127-2 115 e1 ARZ 76 10 F

1105

-20 1o I 1o Top Head Flange (O 209) 209-127-2 AKU75 10 F

108 111 106

-20 10 10 Top Head Dollar (MK201) 201-122-1 85524-2 10 P

42 5D 43

-4 10 10 Top Head Side Plates (W202) 202-127-5 C2426-2 10 P

53 80 79

-20 10 10 202-12746 C2426-2 10 P

80 69 75

-20 10 10

-127-7 C2426G4 10 P

91 71 98

• 20 10 10 202-1274 Ct717-3 10 P

74 69 95

-20 10 10 202-127-10 C1717-3 10 P

93 51 90

-20 10 10 202-127-11 C17224 I o P

81 go 101

-20 10 10 Shell Comses Upper Shel Plates QIAK6a) 6-127-11 C2559-2 10 P

31 60 54 18 10 18 127-21 C2791-1 10 P

83 29 54 22 10 22 127-22 c266o-1 10 P

67 70 67

-20 10 10 ransition Shel Plates (MK 16) 127-1 C2533-1 10 P

58 58 72

-20 10 10 5-127-3 C2533-3 10 P

57 67 47

-14 10 10 15-127-4 B58423 10 P

63 81 46

-12 10 10 Upper Irdernediate Shell Plates (MK 59) 6127-18 C2528-1 10 P

66 79 71

-20 10 10 6-127-23 C24632 10 P

69 60 72

-20 10 10 6-127-24 C2605-2 10 P

56 71 57

-2_

10 10 Lower Intermediate Shell Plates (8) 6-1274 A09s81-I0 P

72 68 65

-20

-10

-10 6127-16 C2467-1 I 0 P

65 81 74

-20

-10

-10 6127-20 C2849-1 10 P

01 59 74

-20

-10

-10 r Shell Plates (MK 57) 6-127-14 C2467-2 10 P

59 78 65

-20

-2

-20 127-15 C24S3-1 10 P

85 74 54

-20

-20

-20 6-127-17 C2480-2 10 P

59 60 40 0

-20 0

Bottom Head Bottom Head Dolar MWI) 1-139-1 C2669-2 40 P

34 40 34 42 40 42 Bottom Fead Upper Torus (W 2) 2-139-1 B8747-1 40 P

78 80 75 10 40 40 2-139-2 B6747-1 40 P

S0 54 77 10 40 40 2-139-3 BS776-2 40 P

90 88 92 10 40 40 2-139-4 B677-2 40 P

53 64 50 10 40 40 2-127-11 C2369-1 40 P

81 80 88 10 40 40 2-127-12 C2369-1 40 P

101 101 100 10 40 40 Bottom Head Lower Torus (MK 4) 4-127-5 C2412-1 40 P

57 66 68 10 40 40 4127-6 C2412-1 40 P

67 45 76 20 40 40 4-127-7 C2412-2 40 P

74 79 60 10 40 40 4-127-8 C2412-2 40 P

60 55 43 24 40 40 NOTE: These are minimum Charpy values.

GE Nudear Energy GE-NE-0000-0013-3193-Ola-R1 Non-Proprietary Version Table 4-2: RTNDT Values for Browns Ferry Unit 2 Nozzle and Weld Materials NI Recrc OutetNozzle MK 8) 8-127-3 8-127-4 E31VW 431H-3 E30VW 431H-4 40 F

AA 86 84 B1 100 10 40 98 1971 10 1

40 40 40 N2 Recrc Inlet Nozzle (MK 7) 7-127-11 E25VW 433H-11 40 F

97 106 B6 10 40 40 7-127-12 E26VW433H-12 40 F

63 106 94 10 40 40 7-127-13 E25VW 4331-13 40 F

95 79 101 10 40 40 7-127-14 E25VW 4331-14 40 F

93 103 107 10 40 40 7-127-15 E25VW433H-15 40 F

105 113 98 10 40 40 7-127-16 E25VW433H-16 40 F

100 97 86 10 40 40 7127-17 E25VW 433H-17 40 F

101 94 99 10 40 40 7.127-18 E25VW433H-18 40 F

112 96 85 10 40 40 7-127-19 E26VW433H-19 40 F

107 87 110 10 40 40 7-127-20 E25VW433H-20 40 F

96 92 116 10 40 40 N3 Steam Outlet Nozzle fdK 14) 14-127-5 E26VW 435H-40 F

106 113 107 10 40 40 14-127-0 E26VW435H-6 40 F

112 99 116 10 40 40 14-127-7 E26VW43SH-7 40 F

87 99 106 10 4D 40 14.12748 E28VW435H-8 40 F

113 112 95 10 40 40 N4 Feedwater Nozzle e4K10) 10-127-7 E25VW436H-7 40 F

108 94 112 10 40 40 10-127-4 E25VW4361-1-8 40 F

112 105 113 10 40 40 10-127-9 E25VW 4361--9 40 F

103 112 93 10 40 40 10-127-10 E25W 436H-10 40 F

B9 78 91 10 40 40 1D-127-11 E25VW 436H-11 40 F

112 104 119 10 40 40 10-127-12 E25VW436H-12 40 F

III 94 94 10 40 40 N5 Core Spray Nozzle (MK 11) 11-139-1 EV9964 N7H6029B 40 F

38 42 44 34 0

34 11-127-3 E26VW437H-3 40 F

73 96 87 10 40 40 NS Top Head Instrumentation Nozzle (MK 206) 206-146-1& 2 BT2615-4 40 F

123 143 144 10 40 40 N7 Top Head Vent Nozzle (MK 204) 204-127-2 ZT3043-3 40 F

102 130 117 10 40 40 NS Jet Pump Instrumentation (MK 19) 19-122-3 a -4 214484 40 F

37 35 23 65 40 65 N9 CRD HYD System Return Nozzle (MK 13) 13-127-2 E23VW 438H 40 F

120 112 114 10 40 40 N10 Core 6P & Uquid Control Nozzle (WM I) 17-139-1 Z73043-1 40 F

155 154 156 10 4O 40 N11 N12. N16 Instrumentation Nozzle IMK 12)

Inconel 12-127-13 through 16 8801 12-139-118 *12 8601 N13.N14 High & Low Pressure Seal Leak QMK139 N15 Drain Nozzle OMK22) 2-139-1 7478 40 F

136 160 172 10 40 40 WELDS_

Cylinddcal SheN Axial Welds Electroslag Welds ESW 3.1 Girh Welds Shell I to Shel 2 (MK57 to MIK58) 066733

_40-

• No CMTR information; obtained from Purchase Specification 21A1 111.

Weld initial RTNDT values obtained from 1 31.

NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-001 3-3193-Ola-RI Non-Proprietary Version Table 4-3: RTNDT Values for Browns Ferry Unit 2 Appurtenance and Bolting Materials Support Skirt (MK 24 )

24-127-5 through -8 A4846-5 40 11091 96 104 10 40 1 40 Shroud Support (MK 51, MK52, M K53)

Alloy 600 51-127-5 through 8 65952-5 52-127-3 and 4 65952-5 53-127-5 332273-4 53-127-10 through -12 & -14 56782-S53-127.13 & -16 567826 53-127-15 56825-1 Steam Dryer Support Bracket (MK131)

Stainless Steel 131 00431 Core Spray Bracket (MK132)

Stainless Steel 132 3342230 Dryer Hold Dovm Bracket (MK133) 133 EV8448 40 94 110 113 10 40 40 Guide Rod Bracket (MKI34)

Stainless Steel 134 00431 Feedwater Sparger Brackets -MK135 Stainless Steel 135 00431 Stabilizer Bracket (IK 196) 196 C6458-1 10 60 59 5s

-20 40 40 Surveillance Brackets (MK199 & MK200)

Stainless Steel 199,200 342633-2 LUfting Lugs (K210) 210 A1210-3 10 98 108 72

-20 0

0 CRD penetrations (MKIO1 - MK1128)

Aloy 600 101 through 128 Refuelhig Containment Skirt (MK71) 71-127-5 through 8 87478-4B 10 110 89 102

-20 10 10 STUDS:

Closure (MK61) 6730502 10 34 52 68 n/a 70 6780382 10 42 42 42 n/a 70 6720443 10 35 38 37 n/a 70 NUTS:

Closure (MK62) 6730502 10 34 52 68 n/a 70 BUSHINGS:

Closure (MK63)

T3798 10 61 68 73 51 10 M2513 10 64 65 67 40 10 IM3232 10 65 65 68 45 10 WASHERS:

losure (MK64) 6790156 10 nWa n/a nta n/a 70 losure (MK64 and MK65) 6730502 10 34 52 68 nW 70 Cbsure (MK64) 6780278 10 40 43 43 n/a 70 NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-0013-3193-OIa-RI Non-Proprietary Version 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG1.99) provides the methods for determining the ART. The RG1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beltline plates and welds was performed and is summarized in Tables 4-4 and 4-5 for 23 and 30 EFPY, respectively.

4.2.1 Regulatory Guide 1.99, Revision 2 (RGI.99) Methods The value of ART is computed by adding the SHIFT term for a given value of effective full power years (EFPY) to the initial RTNDT. For RG1.99, the SHIFT equation consists of two terms:

SHIFT = ARTNDT + Margin

where, ARTNOT = [CF]. f (028-0.0logt Margin = 2(0r12 + oA2) 0.5 CF = chemistry factor from Tables I or 2 of RG1.99 f = Y4T fluence /1019 Margin = 2(Ci 2 + A2) 0.5 ofI

= standard deviation on initial RTNDT, which is taken to be 0F (13IF for electroslag welds).

oA

= standard deviation on ARTNDT, 280F for welds and 17"F for base material, except that c,, need not exceed 0.50 times the ARTND value.

ART = Iniftal RTNDT+ SHIFT The margin term qa has constant values of 170F for plate and 280F for weld as defined in RG1.99. However, A,& need not be greater than 0.5 - ARTNDT. Since the GE/BWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version of a, is taken to be 00F for the vessel plate and most weld materials, except that a, is 130F for the beltline electroslag weld materials [13].

4.2.1.1 Chemistry The vessel beltline chemistries were obtained from [13].

The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG1.99, to determine a chemistry factor (CF) per Paragraph 1.1 of RG1.99 for welds and plates, respectively. Best estimate results are used for the beltline electroslag [13] materials for the initial RTNDT; therefore, the standard deviation (a') is specified.

4.2.1.2 Fluence An EPU (Extended Power Uprate) flux for the vessel ID wall was calculated using methods consistent with Regulatory Guide 1.190. The flux is determined for the EPU rated power of 3952 MWt.

The peak fast flux for the RPV inner surface is 1.4e9 n/cm2-s for EPU conditions. The calculated fast flux at the Browns Ferry Unit 2 Cycle 7 capsule center is 8.85e8 n/cm2-s with a corresponding lead factor of 0.98.

This calculation was performed prior to Regulatory Guide 1.190 (RG1.190), using methodology similar to RG1.190. [

)) fast flux at this capsule Is 9.5e8 n/cm2-s. The flux wire measurement for the Browns Ferry Unit 2 Cycle 7 capsule removed during the Fall 1994 refueling outage at 8.2 EFPY is 5.9e8 nicm2-s [22] (with a lead factor of 0.98), resulting in a calculation-to-measurement ratio of 1.6. The currently licensed Browns Ferry Unit 2 P-T curves are based upon a 32 EFPY fluence of 1.1e18 nlcm2, which was derived from the first cycle dosimetry flux of 1.06e9 n/cm2-s.

30 EFPY Fluence Browns Ferry Unit 2 will begin EPU operation at approximately 18.1 EFPY, thereby operating for 11.9 EFPY at EPU conditions for 30 EFPY. As can be seen above, use of the EPU flux of 1.4e9 nlcm2-s to determine the fluence for the entire 30 EFPY (representing the GE Nuclear Energy I 1, GE-NE-0000-0013-3193-012-Rl Non-Proprietary Version 40 year Browns Ferry Unit 2 license period) is conservative. The RPV peak ID fluence is calculated as follows:

1.4e9 ncm2-s - 9.46e8 s = 1.33e18 n/cm2.

This fluence applies to the lower-intermediate plate and axial weld materials. The fluence is adjusted for the lower shell and axial welds, as well as for the lower to lower-intermediate girth weld based upon a peak / lower shell location ratio of 0.81 for EPU conditions (at an elevation of approximately 258" above vessel 00"); hence the peak ID fluence used for these components is 1.1e18 n/cm2.

It was determined that the EPU axial flux distribution factor bounds the pre-EPU factor calculated during the 1995 capsule evaluation [22].

The fluence at 114T is calculated per Equation 3 of Regulatory Guide 1.99, Revision 2 [7]

using the Browns Ferry Unit 2 plant specific fluence and vessel thickness of 6.13". The 30 EFPY 114T fluence for the lower-intermediate shell plate and axial welds is:

1.33e18 n/cm2

• exp(-0.24 * (6.1314)) = 9.2e17 n/cm2.

The 30 EFPY 114T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

1.1e18 n/Cm2

• exp(-0.24 * (6.1314)) = 7.4e17 n/cm2.

23 EFPY Fluence The RPV peak ID fluence for 23 EFPY is scaled from the 30 EFPY calculation above:

1.33e18 n/cm2 * (23 / 30) = 1.Oel8 n/cm2.

Similarly, this fluence applies to the lower-intermediate plate and axial weld materials. The fluence is adjusted for the lower shell and axial welds, as well as for the lower to lower-intermediate girth weld based upon a peak I lower shell location ratio of 0.81 for EPU conditions; hence the peak ID fluence used for these components is 8.2e17 n/cm2.

The fluence at 1/4T is calculated per Equation 3 of Regulatory Guide 1.99, Revision 2 [7]

using the Browns Ferry Unit 2 plant specific fluence and vessel thickness of 6.13". The 23 EFPY 114T fluence for the lower-intermediate shell plate and axial welds is:

1.0e18 n/cm2

• exp(-0.24 * (6.13 / 4)) = 70e17 n/crr 2.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version The 23 EFPY 114T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

8.2e17 nCM2 - exp(-0.24 * (6.13 /4)) = 5.7e17 ncm2.

4.2.2 Limiting Beltline Material The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTwcT. Using initial RTNDT, chemistry, and fluence as inputs, RG1.99 was applied to compute ART.

Tables 4-4 and 4-5 list values of beltline ART for 23 and 30 EFPY, respectively.

GE Nudear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version Table 4-4: Browns Ferry Unit 2 Beltline ART Values (23 EFPY)

Lowu4ntermedafe Pi. and Adl Welds Thi&n-6.13 hdcs Leow Pigs wid Mal Wrlds wid Lwmr to Lowvr-riermsi*

GMlh Weld Thkk GS13 MM S0E1PYPakMfitw-I.Et1S nkr'2 3OEFPT7.ekU4ftiunc,-

9.217 mbu*2 23 EYhtkV4lincs-7.1E.17 W

ako2 3OE P*akW.i--

l.l3+12 okns2 30113WPYkMju4m-7AE+17 nRAs2 23 WffTkj/4TJoucs-5.7E+17 naW7 hIRl1l 1WT 23 Y

CA 23 EM 23 EFPY Cos&0iM i, HmATI oMT=

16OR O RTPkT Pbnace A RTk mu*

fli ART

_r 1ans2 T

T IF MATIM:

5 t27-14 C2467-2 0.16 0.52 112

-20 53731?

35 0

17 34 69 49

.127-15 C2463-1 0.17 0.48 117

.20 S.7B17 37 0

17 34 71 Si 6127-17 C2460-2 0.13 0.S1 6n 0

S.7E+17 2

0 14 2Y

'5 33 L.intunu.e Duel 6-127-4 A091-1 0.14 0S5 66

.10 7.0£+1' 34 0

1?

14 a6 S

8-127-16 C2407-1 0.10 0.52 112

-10 7.0£ 1?

39 0

17 34 n3 63 6S17-20 C2M9U1 0.11 0.60 73 10 7.0.+17 26 0

13 26 51 41 hJ MU 0.24 0.37 141 23.1 7.01.1 49 13 2I 56 145 121 000h 015733 0.00 OS 117

• 40 5.7z1.?

37 6

1s 37 73 33

• hsm os tined tusar l131 GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version Table 4-5: Browns Ferry Unit 2 Beltline ART Values (30 EFPY)

LenNrbntemnate Riate ad AMal Welds T11kn-4.13 hd 30 771 NFYP*D.aics -

1J.2+18 sks-2 30 EPFYTPak1/4T ce 9.23+17 atm'2 30 3fllakIMT4AT cs-91E117 nhn^12 Lowe Plat a Axdial Welds id Lowr bo Lawer.lihemedite Mt Weld lhkkln 6.13 "elia 3O7YFnekIMlluse-

.IB+13 nksi2 3OWfYlMA4TSumcc-7.H+17 Wan7 30 EW YN~kIMThcc-7AE+17 PIW2

.Ie W 11T 30 3FY a,

_A 30 EFrY 30 EFO Y COUOMtr 29"3OI3IMATA.

%Ci Y%

(C ITAd ftence A RThdk MniJ shift ART E_____________

F Ward.2 T

IF T

IF a.AM 6-127-14 C2467-2 0.1B 0.52 112

-20 7A.417 40 0

17 34 74 54 6-127-15 C2463-1 0.1 0.43 117

-20 7.41E17 42 0

17 34 76 56 8-127-17 C2460-2 0.13 0.51 e6 0

7AE+.1 32 0

16 32 a

43 rtwmdldte 5.

6-1274 MMi-1 0.14 0.55 06

-10 9.33*1' 39 0

17 34 73 3

6127-16 C24B7-1 0.1 0.52 112

-10 9.23317 45 0

17 34 79 69 6127-20 C2949-1 0.11 0.50 73

-10 9.33+1?

29 0

15 29 So 48 Amdl Eh 0.24 037 141 23.1 923.17 56 13 23 62 118 141 camh 5105733 029 0as 117

.40 7.4R17 42 a

21 42 P

44

' ChelWkes theld tam 1131 GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) IOCFR50 Appendix G [8] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions to which a pressure-retaining component may be subjected over its service lifetime.

The ASME Code (Appendix G of Section Xl [6]) forms the basis for the requirements of 10CFR50 Appendix G.

The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B & C)

The closure flange region Includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTNOT. The remaining portions of the vessel (i.e., upper vessel, lower vessel) include shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will also be called the non-beltline region.

For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 100*F/hr or less for which the curves are applicable.

However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version nozzle thermal cycle diagrams P31. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 15'F/hr or less must be maintained at all times.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, It is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest Is in the inner wall during cooldown and is In the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, Klr, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

The applicable temperature Is the greater of the IOCFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits.

A summary of the requirements Is provided in Table 4-6.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version Table 4-6: Summary of the 1 OCFR50 Appendix G Requirements fylUI UbtLIU re VbU I Ubt OL Lei (Core is Not Critical) - Curve A

1. At c 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 600F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNpT + 900F

11. Normal operation (heatup and cooldown),

including anticipated operational occurrences

a. Core not critical - Curve B

1. At < 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region Initial RTNOT + 60F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 1200F

b. Core critical - Curve C

1. At c 20% of preservice hydrotest Larger of ASME Umits + 40*F or of a.1 pressure, with the water level within the normal range for power operation

2. At > 20% of preservice hydrotest Larger of ASME Limits + 400F or of pressure a.2 + 40°F or the minimum permissible temperature for the inservice system hydrostatic pressure test

• 600F adder is included by GE as an additional conservatism as discussed in Section 4.3.2.3.

There are four vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the two regions in the remainder of the vessel (i.e., the upper vessel and lower vessel non-beltline regions). The closure flange region limits are controlling at lower pressures primarily because of 10CFR50 Appendix G [8]

requirements.

The non-beltline and beltline region operating limits are evaluated according to procedures in IOCFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15].

The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version

))

4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beitline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.Oe17 n/cm2) to cause any significant shift of RTNDT. Non-beltline components include nozzles (see Appendix E),

the closure flanges, some shell plates, the top and bottom head plates and the control rod drive (CRD) penetrations.

Detailed stress analyses of the non-beltline components were performed for the BWRI6 specifically for the purpose of fracture toughness analysis. The BWR6 stress analysis bounds for BWRI2 through BWRI5 designs, as will be demonstrated in the following evaluation.

The analyses took into account all mechanical loading and anticipated thermal transients.

Transients considered include 100°F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling Injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWRJ6 components: the feedwater nozzle (FV) and the CRD penetration (bottom head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-7 and 4-8.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-Ri Non-Proprietary Version Table 4-7: Applicable BWRI4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Water Level Instrumentation Nozzie Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Drain Nozzle Table 4-8: Applicable BWR/4 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Dl$¢OntinUittenCtifctOn-;:

EEEE l". l 7

~~CRD and Bottom Head Top Head Nozzles 7

~~Recirculation Outlet Nozzle 7 7

~~~Drain Nozzle Shell**

Support skin**

Shroud Support Attachments~

Core AP and Liquid Control Nozzle**

These discontinuities are added to the bottom head curve discontinuity list to assure that the entire bottom head Is covered, because separate bottom head P-T curves are provided to monitor the bottom head.

The P-T curves for the non-beitline region were conservatively developed for a large BWR/6 (nominal Inside diameter of 251 inches). The analysis is considered appropriate GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version for Browns Ferry Unit 2 as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4. The generic value was adapted to the conditions at Browns Ferry Unit 2 by using plant specific RTNDT values for the reactor pressure vessel (RPV).

The presence of nozzles and CRD penetration holes in the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.

This was the result of the stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.

((

4.3.2.1.1 Pressure Test - Non-Belfine, Curve A (Using Bottom Head)

In a ((

)) finite element analysis ((

)), the CRD penetration region was modeled to compute the local stresses for determination of the stress Intensity factor, K1.

The ((

D evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section Xi Appendix G [6] and shown below. The results of that computation were K = 143.6 ksi-in1'2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel).

The computed value of (T - RTNDT) was 840F. [

1]

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version The limit for the coolant temperature change rate is i 15Fhr or less.

11~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~1 The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on a thickness of 8.0 Inches; hence, tmn = 2.83. The resulting value obtained was:

Mm = 1.85 for -it2 Mm = 0.926 At for 2<t<3.464 = 2.6206 Mm = 3.21 for ti >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Kb is calculated from the equation in Paragraph G-2214.2 [6]:

K. mM.m-Cypm = ((

K~b =(2/3) M.m Cpb ((

J] ksi-inm

)) ksi-in'4 The total K1 is therefore:

K = 1-5 (Krm+ KWb) + Mm - (a,,

+ (213)

• iab)

= 143.6 ksi-in 2 This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific K1 is based on the 4, equation of Paragraph A-4200 in ASME Appendix A [171:

GE Nuclear Energy GE-NE-0000-0013-3193-012-RI Non-Proprietary Version (T - RTDT) = In [(Kg - 33.2) 120.734] / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 84F The generic curve was generated by scaling 143.6 ksi-in"2 by the nominal pressures and calculating the associated (T - RTmDT):

Pressure Test CRD Penetration K and (T - RTNDT) as a Function Of Pressure

-- ;;Nominal Pressure KiITPR~

1563 144 84 1400 129 77 1200 111 66 1000 92 52 800 74 33 600 55 3

400 37

-88 The highest RTNDT for the bottom head plates and welds is 420F, as shown in Tables 4-1 and 4-2. II GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Verslon 11 Second, the P-T curve is dependent on the calculated K1 value, and the K1 value is proportional to the stress and the crack depth as shown below:

K1 cc a (a) 1 2

(4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, K1 is proportional to R/(t) m. The generic curve value of R/(t)'12, based on the generic BWR/6 bottom head dimensions, is:

Generic:

R / (t) "2 = 138 / (8) "' = 49 inch"2 (4-2)

The Browns Ferry Unit 2 specific bottom head dimensions are R = 125.7 inches and t =8 inches minimum [19], resulting in:

Browns Ferry Unit 2 specific:

R / (t)1 2 = 125.7/ (8)12 = 44 inch"'

(4-3)

Since the generic value of R/(t) 2 is larger, the generic P-T curve is conservative when applied to the Browns Ferry Unit 2 bottom head.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version 4.3.2.1.2 Core Not Critical Heatup/Cooldown - Non-Beltine Curve B (Using Bottom Head)

As discussed previously, the CRD penetration region limits were established primarily for consideration of bottom head discontinuity stresses during pressure testing.

Heatup/cooldown limits were calculated by increasing the safety factor in the pressure testing stresses (Section 4.3.2.1.1) from 1.5 to 2.0. ((

1]

The calculated value of K1 for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KIR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical. Therefore, the K, value for the core not critical condition is (143.6 / 1.5)

• 2.0 = 191.5 ksi-inm.

Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the 1,:

equation of Paragraph A4200 in ASME Appendix A [17] for the core not critical curve:

(T - RTNDT) = In [(K - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T-RTNDT) = 102°F The generic curve was generated by scaling 192 ksi-in"' by the nominal pressures and calculating the associated (T - RTNDT):

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version Core Not Critical CRD Penetration K, and (T - RTNIT) as a Function of Pressure Nominal Pressure T- :RTMJ

.........WW.........

.. -....i EEE..................

..i

~~~~~~~~~~~~~~~~~~...... r i w.

...:(sii'}

1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49

-14 The highest RTNDT for the bottom head plates and welds is 420 F, as shown in Tables 4-1 and 4-2. ii D

As discussed in Section 4.3.2.1.1 an evaluation Is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Table 4-8 and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-Rl Non-Proprietary Version aI 11 GE Nuclear Energy GE-NE-0000-0013-3193-012-Rl Non-Proprietary Version 4.3.2.1.3 Pressure Test - Non-Belfline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, K,, for the feedwater nozzle was computed using the methods from WRC 175 [15] together with the nozzle dimension for a generic 251-inch BWR/6 feedwater nozzle. The result of that computation was 1 = 200 ksi-in"2 for an applied pressure of 1563 psig preservice hydrotest pressure.

II 1]

The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 114T through the comer thickness.

To evaluate the results, 14 is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or Xl). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of 14 is shown below using the BWRI6, 251-inch dimensions:

Vessel Radius, R, 126.7 inches Vessel Thickness, t, 6.1875 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psig

• 126.7 inches / (6.1875 inches) = 32,005 psi.

The dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.97 ksi. The factor F (a/rj) from Figure A5-1 of WRC-175 is 1.4 where:

a = % ( tn 2 + tv 2)112

=2.36 inches t, = thickness of nozzle

= 7.125 inches t, = thickness of vessel

= 6.1875 inches r, = apparent radius of nozzle

= r +' 0.29 r,=7.09 inches r = actual inner radius of nozzle

= 6.0 inches rc = nozzle radius (nozzle corner radius)

= 3.75 inches Thus, air0 = 2.36 / 7.09 = 0.33. The value F(a/r 0), taken from Figure A5-1 of WRC Bulletin 175 for an air0 of 0.33, Is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (na) "

• F(a/rn):

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version Nominal K = 1.5

• 34.97 * (r - 2.36) 1/2

• 1.4 = 200 ksi-in'12 The method to solve for (U - RTNDT) for a specific K1 is based on the K, equation of Paragraph A-4200 in ASME Appendix A [171 for the pressure test condition:

(T - RTNDT) = In ((K, - 33.2) / 20.734 /0.02 (T - RTNDT) = In [(200-33.2) / 20.734] / 0.02 (T - RTNDT) = 104.20F 11 The generic pressure test P-T curve was generated by scaling 200 ksi-in'12 by the nominal pressures and calculating the associated (T -

RTNDT)A [

D GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 11 11 The highest RTNDT for the feedwater nozzle materials is 400F as shown in Table 4-2.

However, the RTNDT was increased to 44°F to consider the stresses in the top head nozzle together with the initial RTNDT as described below. The generic pressure test P-T curve is applied to the Browns Ferry Unit 2 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 440F.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version 1]

Second, the P-T curve is dependent on the K value calculated. The Browns Ferry Unit 2 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K are shown below:

Vessel Radius, R, Vessel Thickness, t, Vessel Pressure, P, 125.7 inches 6.125 inches 1563 psig Pressure stress: a = PR / t = 1563 psig

• 125.7 inches I (6.125 inches) = 32,077 psi.

The dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding car

= 35.04 ksi. The factor F (aIr") from Figure A5-1 of WRC-175 is determined where:

a =

1/4 2 + t 2)1/2

=2.32 Inches U = thickness of nozzle

= 6.96 inches t, = thickness of vessel

6.125 inches r,

apparent radius of nozzle

= r, + 0.29 ri=6.9 inches r, = actual inner radius of nozzle

= 6.0 inches r, = nozzle radius (nozzle comer radius)

= 3.0 inches Thus, afrs = 2.32 / 6.96 = 0.33. The value F(a1rQ), taken from Figure AS-1 of WRC Bulletin 175 for an alrs of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K,, is 1.5 a (na) 12

• F(a/r,):

Nominal K = 1.5 - 35.04 * (w

• 2.32) tQ

• 1.4 = 198.7 ksi-in' 2 GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version 11~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~1 4.3.2.1.4 Core Not Critical HeafuplCooldown - Non-Beftline Curve B (Using Feedwater Nozzle/Upper Vessel Region)

The feedwater nozzle was selected to represent non-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences feedwater flow that is colder relative to the vessel coolant.

Stresses were taken from a ((

D finite element analysis done specifically for the purpose of fracture toughness analysis [f

)). Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 40*F feedwater injection, which is equivalent to hot standby, as seen in Figure 4-3.

The non-beltline curves based on feedwater nozzle limits were calculated according to the methods for nozzles in Appendix 5 of the Welding Research Council (WRC)

Bulletin 175 [15].

The stress intensity factor for a nozzle flaw under primary stress conditions (Klp) is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:

Kp = SF

• es (ira)%

• F(a/r1)

(4-4) where SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/r,) is the shape correction factor.

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Verslon 1[

))

Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(aIrQ) for Equation 4-4.

These values are shown in Figure A5-1 of WRC Bulletin 175 [15].

The stresses used in Equation 4-4 were taken from ((

]J design stress reports for the feedwater nozzle. The stresses considered are primary membrane, garp, and primary bending, Upb. Secondary membrane, m,,, and secondary bending, 0 sb, stresses are included in the total KN by using ASME Appendix G [6] methods for secondary portion, K18= Mm (asC + (2/3)

• NO (4-5)

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [15].

However, the correction was not applied to primary membrane stresses because primary stresses satisfy the laws of equilibrium and are not self-limiting. K1p and K, are added to obtain the total value of stress intensity factor, K1. A safety factor of 2.0 is applied to primary stresses for core not critical heatup/cooldown conditions.

Once Ke was calculated, the following relationship was used to determine (T - RTNDT).

The method to solve for (T - RTNDT) for a specific K1 is based on the K, equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

(T - RTNDT) = In [(K -33.2) I 20.734] / 0.02 (4-6)

Example Core Not Critical Heatup/Cooldown Calculation for Feedwater NozzlelUpper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the [l 1D feedwater nozzle [l D analysis, where feedwater injection of 40°F into the vessel while at operating conditions (551.4 0F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle comer stresses were obtained from finite element analysis ((

D. To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 inches was used in the evaluation.

However, a thickness of 7.5 Inches is not conservative for the pressure stress evaluation. Therefore, the pressure stress (acp) was adjusted for the actual ((

3]

vessel thickness of 6.1875 inches (i.e., apm = 20.49 ksi was revised to 20.49 ksi

• 7.5 inches/6.1875 inches = 24.84 ksi). These stresses, and other inputs used in the generic calculations, are shown below:

Opm = 24.84 ksi

a. = 16.19 ksi ay. = 45.0 ksi t, = 6.1875 inches Opb = 0.22 ksi csb = 19.04 ksi a = 2.36 inches r, = 7.09 inches t, = 7.125 inches GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version In this case the total stress, 60.29 ksi, exceeds the yield stress, ac,, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 [15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the inside surface temperature is used.)

R = [ay. - ap. + ((at&ata - ayd / 30)] / (atota - apm)

(4-7)

For the stresses given, the ratio, R = 0.583. Therefore, all the stresses are adjusted by the factor 0.583, except for a1w. The resulting stresses are:

capm = 24.84 ksi asm = 9.44 ksi spb = 0.13ksi Csb 11.10ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on the 4a thickness; hence, ti'2 = 3.072. The resulting value obtained was:

Mm = 1.85 for sit:S2 Mm = 0.926 4i for 2Lfi3.46S4 = 2.845 Mm = 3.21 for ft >3.464 The value F(a1rQ), taken from Figure AS-1 of WRC Bulletin 175 for an a/re of 0.33, is therefore, F (a / rJ) = 1.4 K4p is calculated from Equation 4-4:

Kip = 2.0 * (24.84 + 0.13) * (n

• 2.36) "2

• 1.4 K1 p = 190.4 ksi-in'"2 K1 is calculated from Equation 4-5:

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version K1S = 2.845 - (9.44 +2/3 11.10)

Kes = 47.9 ksi-in'Q2 The total K is, therefore, 238.3 ksi-in12.

The total Ke is substituted into Equation 4-6 to solve for (T - RTNDT):

(T - RTNDT) = In [(238.3-33.2) /20.734] / 0.02 (T - RTNDT) = 115'F The E

)) curve was generated by scaling the stresses used to determine the K; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40°F water injected into the hot reactor vessel nozzle. In the base case that yielded a K1 value of 238 ksi-in', the pressure is 1050 psig and the hot reactor vessel temperature is 551.4 0F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tawaf - 40) / (551.4 - 40). From Kt the associated (T - RTNDT) can be calculated:

Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure NominalPraszure: S:::atuatnTmp

(-

A~~~~~~~~.,..... ^...

..b. A.

,~~~~~~~~~~~~.

m..

1563 604 0.23 303 128 1400 588 0.34 283 124 1200 557 0.48 257 119 1050 551 0.58 238 115 1000 546 0.62 232 113 800 520 0.79 206 106 600 489 1.0 181 98 400 448 1.0 138 81

• Note:

For each change in stress for each pressure and saturation temperature condition, there is a corresponding change to R that influences the determination of K4.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version The highest non-beltline RTNDT for the feedwater nozzle at Browns Ferry Unit 2 is 400F as shown in Table 4-2. However, the RTNDT was increased to 44°F to consider the stresses in the top head nozzle as previously discussed. The generic curve is applied to the Browns Ferry Unit 2 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 44°F as discussed in Section 4.3.2.1.3.

1[

4.3.2.2 CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beeline region are determined according to the ASME Code [6]. As the beltline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature.

The stress intensity factors (I), calculated for the beltline region according to ASME Code Appendix G procedures [6], were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate; values were calculated for 100°F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits.

4.3.2.2.1 Belitine Region - Pressure Test The methods of ASME Code Section Xl, Appendix G [6] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version thickness (to,,) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

a, = PR / oI (4-8)

The stress intensity factor, Km, is calculated using Paragraph G-2214.1 of the ASME Code.

The calculated value of Km for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [61 for comparison with Kc, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

The relationship between Kc and temperature relative to reference temperature (T - RTNDT) is based on the Kc equation of Paragraph A-4200 in ASME Appendix A [17]

for the pressure test condition:

Kqm

• SF = Kc = 20.734 exp[O.02 (T - RTNDT )j + 33.2 (4-9)

This relationship provides values of pressure versus temperature (from KPR and (T-RTNDT), respectively).

GE's current practice for the pressure test curve Is to add a stress Intensity factor, A,, for a coolant heatuplcooldown rate, specified as 150F/hr for Browns Ferry Unit 2, to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatup/cooldown rate of 100iF/hr.

The K, calculation for a coolant heatup/cooldown rate of 1000F/hr is described in Section 4.3.2.2.3 below.

4.3.2.2.2 Calculations for the Beiline Region - Pressure Test This sample calculation is for a pressure test pressure of 1064 psig at 30 EFPY. The following inputs were used in the beltline limit calculation:

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version Adjusted RTNDT = Initial RTNDT + Shift A = 23 + 118 = 141'F (Based on ART values in Table 4-5)

Vessel Height H = 875.13 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to inside of clad)

R = 125.7 inches Minimum Vessel Thickness (without clad) t = 6.13 inches Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1064 psi + (H - B) 0.0361 psi/inch = P psig

= 1064 + (875.13-216.3) 0.0361 = 1088 psig (4-10)

Pressure stress:

a = PR/t

= 1.088

• 125.7 / 6.13 = 22.3 ksi (4-11)

The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on a thickness of 6.13 inches (the minimum thickness without cladding);

hence, t"2 = 2.48. The resulting value obtained was:

Mm = 1.85 for t1<2 Mm = 0.926 4t for 2L<4t.3.464 = 2.29 Mm = 3.21 for 4 >3.464 The stress intensity factor for the pressure stress is K[m = Mm

• A. The stress intensity factor for the thermal stress, K1t, is calculated as described in Section 4.3.2.2.4 except that the value of "G' is 150F/hr instead of 1000F/hr.

Equation 4-9 can be rearranged, and 1.5 Krm substituted for Kc, to solve for (T - RTNDT).

Using the Kc equation of Paragraph A-4200 in ASME Appendix A [17], Ktm = 51.1, and K,= 1.71 for a 15*F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

GE Nuclear Energy GE-NE-OOOD-0013-3193-Ola-Rl Non-Proprietary Version (T - RTNDT)

= In[(1.5 K*m + Kn, -33.2)1 20.734] /0.02 (412)

= ln[(1.5 51.1 + 1.71 - 33.2) / 20.734] / 0.02

= 38.9°F T can be calculated by adding the adjusted RTNDT:

T = 38.9 + 141 = 179.90F for P = 1064 psig at 30 EFPY 4.3.2.2.3 Beltline Region - Core Not Critical Heatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section Xl Appendix G [6]:

Kc = 2.0 - Km +Kt (4-13) where K.m is primary membrane K due to pressure and K, is radial thermal gradient K due to heatup/cooldown.

The pressure stress intensity factor Km is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.

The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes In the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient Mt from Figure G-2214-1 of ASME Appendix G [6] by the through-wall temperature gradient ATw, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G [6].

The relationship used to compute the through-wall AT, is based on one-dimensional heat conduction through an insulated flat plate:

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version a 2T(X,t) / Ox2 = I (OT(Xst) I at)

(4-14) where T(x,t) is temperature of the plate at depth x and time t, and p is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that aT(x,t) / At = dT(t) I dt = G, where G is the coolant heatup/cooldown rate, normally 100OF/hr. The differential equation is integrated over x for the following boundary conditions:

1. Vessel inside surface (x = 0) temperature is the same as coolant temperature, To.

2. Vessel outside surface (x = C) is perfectly insulated; the thermal gradient dT/dx = 0.

The integrated solution results in the following relationship for wall temperature:

T = Gx2/ 2P - GCx / P + To (4-15)

This equation is normalized to plot (T - To) / AT, versus x / C.

The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G [6]. Therefore, aT, calculated from Equation 4-15 is used with the appropriate Mt of Figure G-2214-1 of ASME Appendix G [6] to compute Kt for heatup and cooldown.

The Ml relationships were derived in the Welding Research Council (WRC)

Bulletin 175 [15] for infinitely long cracks of 114T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 4.3.2.2.4 Calculations for the Beldine Region Core Not Critical HeatuplCooldown This Browns Ferry Unit 2 sample calculation is for a pressure of 1064 psig for 30 EFPY.

The core not critical heatup/cooldown curve at 1064 psig uses the same K, as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational condition rather than test condition; the operational condition necessitates the use of a higher safety factor. In addition, there is a Kt term for the thermal stress. The additional inputs used to calculate Kit are:

Coolant heatup/cooldown rate, normally 100"F/hr G = 100 "F/hr Minimum vessel thickness, including clad thickness C = 0.526 ft (6.125" + 0.188" = 6.313")

Thermal diffusivity at 550°F (most conservative value) p = 0.354 fte/ hr [21]

Equation 4-15 can be solved for the through-wall temperature (x = C), resulting in the absolute value of AT for heatup or cooldown of:

AT = GC2/2,2 (4-16)

= 100 - (0.526)2/ (2

• 0.354) = 39"F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.2914) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kmt = Mt - AT = 11.39, can be calculated. KIm has the same value as that calculated in Section 4.3.2.2.2.

The pressure and thermal stress terms are substituted into Equation 4-9 to solve for (T-RTNDT):

(T - RTNDT)

=

ln[((2

• Km + Kit) -33.2) / 20.734] /0.02 (4-17)

= In[(2

• 51.1 + 11.39 - 33.2) /20.734] /0.02

=

67.8 F GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version T can be calculated by adding the adjusted RTNDT:

T = 67.8 + 141 = 208.8 0F for P = 1064 psig at 30 EFPY 4.3.2.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT.

Similar to the evaluations performed for the bottom head and upper vessel, a BWRI6 finite element analysis [18] was used to model the flange region. The local stresses were computed for determination of the stress intensity factor, Ki. Using a 1/4T flaw size and the Kc formulation to determine T - RTNDT, for pressures above 312 psig the P-T limits for all flange regions are bounded by the 10CFR50 Appendix G requirement of RTNDT + 900F (the largest T-RTNDT for the flange at 1563 psig is 730F).

For pressures below 312 psig, the flange curve is bounded by RTNDT + 60 (the largest T -

RTNDT for the flange at 312 psig is 540F); therefore, instead of determining a T (temperature) versus pressure curve for the flange (i.e., T - RTNDT) the value RTNDT + 60 Is used for the closure flange limits.

In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves. However, some closure flange requirements do impact the curves, as is true with Browns Ferry Unit 2 at low pressures.

The approach used for Browns Ferry Unit 2 for the bolt-up temperature was based on the conservative value of (RTNDT+ 60), or the LST of the bolting materials, whichever is greater.

The 600F adder is included by GE for two reasons:

1) the pre-1971 requirements of the ASME Code Section lii, Subsection NA, Appendix G included the 600F adder, and 2) inclusion of the additional 60'F requirement above the RTNDT provides the additional assurance that a 114T flaw size is acceptable. As shown in Tables 4-1, 4-2, and 4-3, the limiting initial RTNDT for the closure flange region is GE Nuclear Energy GE-NE-0000-0013-3193-01la-R1 Non-Proprietary Version represented by the electroslag weld materials in the upper shell at 23.1OF, and the LST of the closure studs is 700F; therefore, the bolt-up temperature value used is the more conservative value of 83*F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

10CFR50 Appendix G, paragraph IV.A.2 [8] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 900F) and Curve B temperature no less than (RTNDT + 120*F).

For pressures below 20% of preservice hydrostatic test pressure (312 psig) and with full bolt preload, the closure flange region metal temperature is required to be at RTNDT or greater as described above. At low pressure, the ASME Code [6] allows the bottom head regions to experience even lower metal temperatures than the flange region RTNDT.

However, temperatures should not be permitted to be lower than 680F for the reason discussed below.

The shutdown margin, provided In the Browns Ferry Unit 2 Technical Specification, is calculated for a water temperature of 680F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 680F limit, further extensive calculations would be required to justify a lower temperature. The 83OF limit for the upper vessel and beltline region and the 680F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel Is in the vessel. When the head is not tensioned and fuel is not in the vessel, the requirements of IOCFR50 Appendix G (8] do not apply, and there are no limits on the vessel temperatures.

GE Nuclear Energy GE-NEb000-0013-3193-01a-R1 Non-Proprietary Version 4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF IOCFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of IOCFR50 Appendix G [8], Table 1. Table 1 of [8] requires that core critical P-T limits be 400F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40'F for pressures above 312 psig.

Table 1 of 10CFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60*F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 830F, based on an RTNDT of 23.10F.

In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 1600F or the temperature required for the hydrostatic pressure test (Curve A at 1064 psig). The requirement of closure region RTNDT + 1600F causes a temperature shift in Curve C at 312 psig.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version

5.0 CONCLUSION

S AND RECOMMENDATIONS The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B & C)

For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 100OF/hr or less for which the curves are applicable.

However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [31. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 150 F/hr or less must be maintained at all times.

The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, K4r, at 1/4T to be less than that at 3/4T for a given metal temperature.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version The following P-T curves were generated for Browns Ferry Unit 2:

Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 23 and 30 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Umits (CRD Nozzle) curve Is also Individually included with the composite curve for the Pressure Test and Core Not Critical condition.

Separate P-T curves were developed for the upper vessel, beltline (at 23 and 30 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

A composite P-T curve was also generated for the Core Critical condition at 23 and 30 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.

Mhile the Bottom Head (CRD Nozzle) and Upper Vessel (FW Nozzle) curves are valid for the entire plant license period (30 EFPY), for clarity and convenience of Browns Ferry Unit 2 personnel, two (2) sets of these curves are provided, each with a designation of EFPY (either 23 or 30) within the title. It should be understood that this designation of EFPY in non-beltline curves does not Imply limitations with regard to EFPY.

Using the fluence from Section 4.2.1.2, the P-T curves are beltline limited above 590 psig for Curve A and above 430 psig for Curve B at 30 EFPY. At 23 EFPY, the P-T curves become beltline limited above 640 psig for Curve A and above 500 psig for Curve B.

Table 5-1 shows the figure numbers for each P-T curve. A tabulation of the curves is presented in Appendix B.

GE Nuclear Energy GE-N E-0000-0013-3193-01a-RI Non-Proprletary Version Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves

=

A A

A A

B B

B B

Bottom Head Limits (CRD Nozzle) - 23 EFP'Y Upper Vessel Limits (FW Nozzle) - 23 EFPY Beltline Limits - 23 EFPY Bottom Head and Composite Curve A - 23 El Bottom Head Umits (CRD Nozzle) - 23 EFPY Upper Vessel Limits (FW Nozzle) - 23 EFPY Bettline Limits - 23 EFPY Bottom Head and Composite Curve B - 23 El I

Figure 5-1 1 Table B-1 Figure 5-2 j

Table B-1 Figure 5-3 Table B-1

'PY*

I Figure 5-4 S

Table B-2 Flaure 5-5 Table B-1

-4 4

Flaure 5-6 Table B-1

_ ~~~~~I Figure 5-7 Table B-1 Figure 5-8 Table B-2 C

coinComosite Curve C-23 EFPY**

Figure 5-9 Table B-2 B & C Composite Curve C** and Curve B* with Bottom Figure 5-10 Tables B-1 & 2 Head Curve - 23 EFPY 30 EFPY Curves A

Bottom Head Limits (CRD Nozzle) -30 EFPY Figure 5-11 Table B-3 A

Upper Vessel Limits (FW Nozzle) - 30 EFPY Figure 5-12 Table B-3 A

Beitline Limits - 30 EFPY Figure 5-13 Table B-3 A

Bottom Head and Composite Curve A-30 EFPY*

Figure 5-14 Table B4 B

Bottom Head Limits (CRD Nozzle) - 30 EFPY Figure 5-15 Table B-3 B

Upper Vessel Limits (FW Nozzle) - 30 EFPY Figure 5-16 Table 5-3 B

Beftline Limits - 30 EFPY Figure 5-17 Table B-3 B

Bottom Head and Composite Curve B-30 EFPY*

Figure 5-18 Table B-4 C,

Composite Curve C-30 EFPY*

Figure 5-19 Table B-4 B & C Composite Curve C*' and Curve B* with Bottom Figure 5-20 Tables B-3 & 4 Head Curve - 30 EFPY The Composite Curve A & i curve Is the more limiting of three limits: IOCFR5O Bolt-up Limits, Upper Vessel Umits (FW Nozzle), and Beltline Umits. A separate Bottom Head Limits (CRD Nozzle) curve is individually included on this figure.

• • The Composite Curve C curve is the more limiting of four limits: IOCFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Umits (FW Nozzle), and Belbine Limits.

GE Nuclear Energy GE-NE-O000-0013-3193-01a-Rl Non-Proprietary Version 1400 1300 1200 1100 L 1000 a

a. 900 mu 0

800 co W 700 Z

600 z

=

500 X

400 mu 300 I

lINITIAL RTnctt VALUE IS I 149'F FOR BOTTOM HEADlI

[HEATUP/COOLDOWN RATE OF COOLANT

<S 151F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ("F)

Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve Al - 23 EFPY

[15 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version 1400 1300 1200 1100 la rL 1000 10 IL 900

-J S3800 o

700 500

'u a,

400 300 INITIALRTndtVALUEISl

_ 44°F FOR UPPER VESSELl HEATUPICOOLDOWN RATE OF COOLANT

[_'j15OF/HR ACCEPTABLE AREA OF

_ OPERATION TO THE RIGHT OF THIS CURVE

-UPPER VESSEL LIMITS (Inducing I

Flange and FW mjNozzle Umb) 200 200 100 0

0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE rF)

Figure 5-2: Upper Vessel P-T Curve for Pressure Test [Curve A] - 23 EFPY

[150Fhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 1400 1300 1200 1100 X 1000 0

900 0

'i w

800 o 700 a:

600 m

3 500 12 400 30 300 BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F) 23 105 HEATUPICOOLDOWN 1 RATE OF COOLANT I I5FIHR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 200 100 0

0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE Irn 200 Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 23 EFPY

[150F~hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version 1400 1300 1200 1100 CL.

1000 0

0.

900 0I-U.'

uw 800 o

700 x

600 z

3 500 w

en 400 300 INITIAL RTndt VALUES ARE 23.1-F FOR BELTLINE, 44F FOR UPPER VESSEL, AND 49'F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (9F) 23 105 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE UPPER VESSEL AND BELTLINE LIMITS BOTTOM HEAD CURVE 200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE CF)

Figure 5-4: Composite Pressure Test P-T Curves (Curve Al up to 23 EFPY

[1 50Fhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-3193-Ola-RI Non-Proprietary Version 1400 1300 1200 1100 a

1000 0

x

0.

900 ar 800 0

700 o

B00 E

=

500 W

0:

0 400 300 200 100 lINITIAL RTndt VALE IS I 149F FOR BOTTOM HEADI HEATUP/COOLDOWN RATE OF COOLANT I 100°F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE

-BOTTOM HEAD UMITS 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-5: Bottom Head P-T Curve for Core Not Critical [Curve B] - 23 EFPY

[1001F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version 1400 1300 1200 1100 D

A 1000 0

R 700 I-a 600 z

p 500 0

400 300

[INITIAL RTndt VALUE ISl 144-F FOR UPPER VESSELl HEATUPICOOLDOWN RATE OF COOLANT cS I 00°F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHTOFTHISCURVE

-UPPER VESSEL LIMITS (Including Flange and FW Nozzle Umits) 200 100 0

I A

4.

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (CF)

Figure 5-6: Upper Vessel P-T Curve for Core Not Critical [Curve B] - 23 EFPY

[1000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 1400 1300 1200 1100 2

o 1000 I-R 90 id co 800 0) o 700 w

600 E

500 mu a

400 co 300

. I I I I I I I B I I I/II 10CFR50 BOLTIP

_~3F

___ J ~~~aT__

I_

I INITUAL RTndt VALUE IS 1 23.1*F FOR BELTUNE BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (OF) 23 105 HEATUP[COOLDOWN RATE OF COOLANT

< 100°FIHR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE BELTUNE LIMITS 200 100 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

('F)

Figure 5-7: Belfine P-T Curve for Core Not Critical [Curve B] up to 23 EFPY

[1000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-R1 Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 23.1PF FOR BELTLINE, 1300 t

44°F FOR UPPER VESSEL, AND 1200 49°F FOR BOTTOM HEAD I BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT ('F) 23 105

~1000

.000 1

HEATUP/COOLDOWN

c.

900

/

RATE OF COOLANT 800

• 1

- ~100-FIHR U) 800__,._

700 I5,.

/

ACCEPTABLE AREA OF t

600 OPERATION TO THE z

/

RIGHT OF THIS CURVE

3 500 PS10PSI 5

PsIG I u

400 mU IY

~~BOTTOCM.

A.

HE~~AD I 300 68'F UPPER VESSEL 200 7

AND BELTLINE

(

FLANGE LIMITS 100 8W-R

.G...BOTTOM HEAD 100 11 aaF

~~~~~~CURVE 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

('F)

Figure 5-8: Composite Core Not Critical P-T Curves [Curve B] up to 23 EFPY

[1000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a-RI Non-Proprietary Version 1400 1300 1200 1100 D

I1000 c

0-900 0

W..

p 800 o 700 W 600 9

II=

M 3

500 gZ 400 300 INITIAL RTndt VALUES ARE 23.1-F FOR BELTUNE, 44°F FOR UPPER

VESSEL, AND 49F FOR BOTTOM HEAD BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 23 105 I

ACCEPTABLE AREA OF 6---500 PSIG; IOPERATION TO THE RIGHT OF THIS CURVE 312 PSI Minimum Crl 0catity 1 Temperature 83 F

.BELTUNE AND NON-BELTUNE 0

25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE 200 100 0

Figure 5-9: Composite Core Critical P-T,Curves [Curve C] up to 23 EFPY

[100°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version B

C 1400 1300 1200 1100 V 1000 a

AL 900 U) 800 p

700 w

600 z

3 500 Nj e

400 30 300 INITIAL RTndt VALUES ARE 23.1F FOR BELTUNE, 44@F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (OF) 23 105 HEATUPICOOLDOWN RATE OF COOLANT 5 100-FIHR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-10: Composite Core Not Critical [Curve B] including Bottom Head and Core Critical P-T Curves [Curve C] up to 23 EFPY 1100°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 1400 1300 1200 1100 0.

va D 1000 L

900 J

E 800 p

700 t

600 2

m 500 IJ V, 400 00 300 lINITIAL RTndtVAL1E IS I 149°F FOR BOTTOM HEAD I lHEATIP/COOLDOWN RATE OF COOLANT I 15°F/HR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 200 100

-BOTTOM HEADL UMITS 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Figure 5-11: Bottom Head P-T Curve for Pressure Test [Curve A] - 30 EFPY

[15°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 1400 1300 1200 1100 la D 1000 0

a. 900 L

800 Ir o

700 i

600 2

I-3 500 U.'

V 400 30 300 lINITIAL RTndt VALUE IS I

_144-F FOR UPPER VESSELl lHEATUPCOOLDOWN RATE OF COOLANT e ~15'F/HR FABCEPTBAREAOF OPERATION TO THE

_RIGHT OF THIS CURVE

-UPPRVSSEL LIMITS (Induding I

Flange and FW 200 Nole Umb) 200 200 100 0

0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE

('F)

Figure 5-12: Upper Vessel P-T Curve for Pressure Test [Curve Al - 30 EFPY

[150F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version 1400 1300 1200 1100 Is

-1000 C.

900

-Iw cO 800 o

700 0

600 t=

3 500 8

400 ow 00 300 BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT ('F) 30 118 HEATUP/COOLDOWN 1 RATE OF COOLANT

_ 15FIHR ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE

-BELTUNE LIMITS 200 100 0

0 25 50 75 100 125 150 175 200 225 MINIMUM REACTOR VESSEL METAL TEMPERATURE

('F)

Figure 5-13: Beltline P-T Curve for Pressure Test [Curve Al up to 30 EFPY

[15°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version 1400 1300 1200 1100 Is 1000 0

IL 900 0I--J u0 800

@2 o

700 Z

600 t

n 500 1U

X 400 300 I

INITIAL RTndt VALUES ARE 23.1°F FOR BELTUNE, 44°F FOR UPPER VESSEL, AND 49OF FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (OF) 30 118 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE UPPER VESSEL AND BELTUNE LIMITS BOTTOM HEAD CURVE 200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (rF 225 Figure 5-14: Composite Pressure Test P-T Curves [Curve A] up to 30 EFPY

[150F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 1400 1300 1200 1100 Ca 1000 a

2 IL 900 0

I-0n 800 0'

o 700 13 a:

600 2

3 500 LJ E

400 300 INITIAL RTndt VALUE ISI

_149-F FOR BOTTOM HEADI HEATUP/COOLDOWN RATE OF COOLANT

~~<

100*FMH ACCETABLEAREA OF OPERATION TO THE

_RIGHT OF THIS CURVE

-l BOTTOM HEAD UIMITSl

-B0 200 200 100 0

0 25 50 75 100 125 150 175 MINIMUM REACTOR VESSEL METAL TEMPERATURE C'FI Figure 5-15: Bottom Head P-T Curve for Core Not Critical [Curve B] - 30 EFPY

[100OF/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-R1 Non-Proprietary Version 1400 1300 12D0 1100 einm

-. 1000 IL 900 UJ 800 o

700 W 600 z

Z3 500 uJ a-30

. __ ~~~ff_

312 PSIG

_FA G

REGIO t

X

~1~QGE REG ION 83 3 INITIAL RTndt VALUE ISl 44F FOR UPPER VESSEL HEATUP/COOLDOWN RATE OF COOLANT I I100°F/HRl ACCEPTAALE AREA OF OPERATION TO THE RIGHT OF THIS CURVE

-UPPER VESSEL LIMITS (Including Flange and FW Nozle Limits) 200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTORVESSEL METAL TEMPERATURE (OF)

Figure 5-16: Upper Vessel P-T Curve for Core Not Critical [Curve B] - 30 EFPY

[100°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version 1400 1300 1200 1100

.0 aZ 1000U Vl a-900 0o 800 o

700 W

600 E

500 m

0o 400 0:

A

)

BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 30 118 HEATUPICOOLDOWN 1 RATE OF COOLANT 5 1000FIHR ACCEPTABLE AREA OFl OPERATION TO THE RIGHT OF THIS CURVE B

BELTLINE LIMITS 300 200 100 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METALTEMPERATURE rF)

Figure 5-17: Beltline P-T Curve for Core Not Critical [Curve B] up to 30 EFPY

[100°Flhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version 1400 1300 1200 1100 a

1000

0.

900

-a In 800 g

700 r

600 2

I 500 mu ala 400 c

300 INITIAL RTndt VALUES ARE 23.1-F FOR BELTLINE, 44°F FOR UPPER VESSEL, AND 49¶F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT OF) 30 118 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE

-UPPER VESSEL AND BELTLINE LIMITS

-BOTTOM HEAD CURVE 200 100 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-18: Composite Core Not Critical P-T Curves [Curve B] up to 30 EFPY

[100°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version 1400 1300 1200 1100 ci L 1000 0

I-9W

&800 o

700 t

600 S

500 mu E400 300 INITIAL RTndt VALUES ARE 23.1PF FOR BELTLINE, 44°F FOR UPPER

VESSEL, AND 49TF FOR BOTTOM HEAD BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F) 30 118 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE

-BELTUNE AND NON-BELTLINE LIMITS 200 100 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(*F)

Figure 5-19: Composite Core Critical P-T Curves [Curve C] up to 30 EFPY

[1000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version B

C 1400 1300 1200 1100 a 1000 A.

900

0) 800 0)

I 700 t

600 2iI:

500 N

)

w 400 NJ A-INITIAL RTndt VALUES ARE 23.1F FOR BELTLINE, 44°F FOR UPPER VESSEL, AND 490F FOR BOTTOM HEAD BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (F) 30 118 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 300 COMPOSITE CURVE B BOTTOM HEAD CURVE B COMPOSITE CURVE C E,

200 100 IIII 0

0 25 50 75 100 125 150 175 200 225 250 275 300 MINIMUM REACTOR VESSEL METAL TEMPERATURE rF)

Figure 5-20: Composite Core Not Critical [Curve B] including Bottom Head and Core Critical P-T Curves [Curve C] up to 30 EFPY [10(0 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version

6.0 REFERENCES

1. RG Carey (GE) to HL Williams (TVA), "New Bounding EFPY for Previously Generated P-T Curves Considering Power Uprate for Browns Ferry Units 2 & 3 Using Calculated Fluence and Estimated ESW Information", GENE, San Jose, CA, December 11, 1998 (RGC-9803).

2.

GE Drawing Number 729E7625, "Reactor Thermal Cycles - Reactor Vessel," GE-APED, San Jose, CA, Revision 0 (GE Proprietary).

3. GE Drawing Number 135B9990, "Nozzle Thermal Cycles - Reactor Vessel,' GE-APED, San Jose, CA, Revision 1 (GE Proprietary).

4.

"Codes and Standards", Part 50.55a of Title 10 of the Code of Federal Regulations, December 2002.

5. Technical Specifications For Browns Ferry Nuclear Plant, Unit 2.

6. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section III or Xl of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

7. "Radiation Embrittlement of Reactor Vessel Materials', USNRC Regulatory Guide 1.99, Revision 2, May 1988.

8. "Fracture Toughness Requirements", Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

9. Hodge, J. M., "Properties of Heavy Section Nuclear Reactor Steels", Welding Research Council Bulletin 217, July 1976.

10. GE Nuclear Energy, NEDC-32399-P, "Basis for GE RTNDT Estimation Method",

Report for BWR Owners' Group, San Jose, California, September 1994 (GE Proprietary).

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version

11. Letter from B. Sheron to R.A. Pinelli, "Safety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation

Method, September 1994",

USNRC, December 16,1994.

12. QA Records & RPV CMTRs Browns Ferry Unit 2 GE PO# 205-55577, Manufactured by B&W, 'General Electric Company Atomic Power Equipment Department (APED)

Quality Control -

Procured Equipment, RPV QC", Mt. Vernon, Indiana, and Madison, Indiana.

13. Letter, TE Abney (TVA) to NRC, "Browns Ferry Nuclear Plant (BFN) - Units 2 and 3-Technical Specification (TS) Change No. 393, Supplement I -

Pressure-Temperature (P-T)

Curve Update",

Docket Nos.

50-260 and 50-296, (TVA-BFN-TS-393, Supplement 1, 10 CFR 50.90 (ROB 981215 742)),

December 15, 1998.

14. Letter, S.A. Richard, USNRC to J.F. Kiapproth, GE-NE, "Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)", MFN 01-050, September 14,2001.

15. "PVRC Recommendations on Toughness Requirements for Ferritic Materials",

Welding Research Council Bulletin 175, August 1972.

16. ((

))

17. "Analysis of Flaws', Appendix A to Section Xl of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version

18. ((

1]

19. Bottom Head and Feedwater Nozzle Dimensions:

a. Babcock & Wlcox Company Drawing 122859E, Revision 10, Lower Head Forming Details- (GE VPF 1805-003).

b. Babcock & Wilcox Company Drawing 94975C, Revision 1, WMK-10 12" Feedwater Nozzle" (GE VPF 1805-035).

20. ((

1]

21.

Materials - Properties", Part D to Section II of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

22. C. Oza, "Browns Ferry Steam Electric Station Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis", GE-NE, San Jose, CA, August 1995 (GENE-B1100639-01, Revision 1).

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a-R1 Non-Proprietary Version APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-NE-0000-0013-3193-01ca-R1 Non-Proprietary Version

4.

I I

I.

I

4.

I 9*

9

4.

4

4.

4 t

I

1.

I t

I

4.

4 t

I I.

I I.

I I*

I I.

I

4.

4

9.

I

4.

4 4.

11 A-2

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Per ASME Code Appendix G, Section G2223 (c), fracture toughness analysis to demonstrate protection against non-ductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5' or less provided the lowest service temperature is not lower than RTFvr plus 60*F. Also Inconel discontinuities require no fracture toughness evaluations.

Nozzle or Appurtenance Material Reference Remarks MK12 - 2' Instrumentation (attached to Alloy 600 1, 6 9, 10. 23 Nozzles made from Alloy 600 and less Shells 58, 59, 60) Shell 58 MK12 nozzle than 2.5 require no fracture toughness Is Within the beltline region (see evaluation.

Appendix E).

MK 71 - Refueling Containment Skirt SA302 GR B 1,24, 25 Not a pressure boundary component; Attachment (to Shell Flange) therefore requires no fracture toughness evaluation.

MK 74, 75, 81, 82 - Insulation Brackets Carbon Steel 1, 26 Not a pressure boundary component; (Shells 57 and 59) therefore requires no fracture toughness evaluation.

MK 85, 86 - Thermocouple Pads (all Carbon Steel 1, 27 Not a pressure boundary component; Shells, Shell Flange, Bottom Head, therefore requires no fracture toughness Feedwater Nozzle) evaluation.

MKIOI - 128 - Control Rod Drive Stub Alloy 600 1,12,15,16 Nozzles made from Alloy 600 require no Tubes (in Bottom Head Dollar Plate) fracture toughness evaluation.

MKI31 - Steam Dryer Support Bracket SA182 F304 1,21,22 Appurtenances made from Stainless (Shell 60)

Steel require no fracture toughness evaluation.

MKI 32 - Core Spray Bracket (Shell 59)

SA276 T304 1,21,22 Appurtenances made from Stainless Steel require no fracture toughness evaluation.

MK133 - Dryer Hold Down Bracket (Top SA508 CL2 1, 22 Not a pressure boundary component; Head Flange) therefore requires no fracture toughness evaluation.

MK134 - Guide Rod Bracket (Shell SA182 F304 1,21, 22 Appurtenances made from Stainless Flange)

Steel require no fracture toughness evaluation.

MK135 - Feedwater Sparger Bracket SA182 F304 1, 21, 22 Appurtenances made from Stainless (Shell 59)

Steel require no fracture toughness I____

________evaluation.

MK 139* - N13 High and N14 Low Carbon Steel 1, 24 Not a pressure boundary component; Pressure Seal Leak Detection therefore requires no fracture toughness Penetration (Shell Flange) evaluation.

MK199, 200 - Surveillance Spedmen SA276 304 1,21,22 Appurtenances made from Stainless Brackets (Shells 58 and 59)

Steel require no fracture toughness I___ ____ ____ ____ ____ ____ ____

________evaluation.

MK 210 - Top Head Lifting Lugs SA302 GR B 1, 17 Loading only occurs during outages. Not a pressure boundary component; therefore requires no fracture toughness evaluation.

• The higMow pressure leak detector, and the seal leak detector are the same nozzle; these nozzles are the closure flange leak detection nozzles.

A-3

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version APPENDIX A

REFERENCES:

1. Vessel Drawings and Materials:

Drawing #24185F, Revision 11, "General Outline", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-042).

Drawing #24186F, Revision 14, "Outline Sections', Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-018).

Drawing #24187F, Revision 11, "Vessel Sub-Assembly", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-059).

Drawing #122855E, Revision 14, "Ust of Materials", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-056).

Drawing 886D499, Revision 12, 'Reactor Vessel", General Electric Company, GENE, San Jose, California.

2. Task Design Input Request (DIR), Pressure-Temperature Curves, Browns Ferry Units 2&3", V. Schiavone (TVA), February 25, 2003.

3. Drawing #122859E, Revision 10, "Lower Head Forming Details", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-003).

4. Drawing #122860E, Revision 8, "Shell Segment Assembly Course #1 and #4",

Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-017).

5. Drawing #122864E, Revision 4, "Recirculation Nozzles", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-041).

6. Drawing #122861E, Revision 8, "Shell Segment Assembly Course #3", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-020).

7. Drawing #94975C, Revision 1, "MK-10 12" Feedwater Nozzle Forgingm, Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-035).

8. Drawing #94976C, Revision 1, "MK-11 Core Spray Nozzle Forging", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-036).

9. Drawing #122868E, Revision 5, "2 Instrument and 4' CRD HYD System Return Nozzles", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-054).

10. Drawing #122862E, Revision 6, "Shell Segment Assembly Course #5", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-019).

11. Drawing #122865E, Revision 4, "26" Steam Outlet Nozzle", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-040).

12. Drawing #122856E, Revision 11, uLower Head Assembly", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-013).

A-4

GE Nuclear Energy GE-NE-0000-0013-3193-018-Ri Non-Proprietary Version

13. Drawing #122858E, Revision 11, "Lower Head Upper Segment Assembly", Babcock

& Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-012).

14. Drawing #122869E, Revision 3, "4" Jet Pump Nozzle", Babcock & WIcox Company, ML Vernon, Indiana (GE VPF #1805-051).

15. Drawing #122857E, Revision 11, "Lower Head Bottom Segment Assembly",

Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-011).

16. Drawing #149938E, Revision 2, "Control Rod Nozzles, Unit #2", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-144).

17. Drawing #122876E, Revision 7, 'Closure Head Assembly", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-049).

18. Drawing #122877E, Revision 5, "Closure Head Nozzles", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-048).

19. Drawing #122872E, Revision 8, "Support Skirt Assembly and Detail", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-108).

20. Drawing #122870E, Revision 6, "Shroud Support, Babcock & Wilcox Company, Mt.

Vernon, Indiana (GE VPF #1805-039).

21. Drawing #122881E, Revision 9, "Vessel Subassembly Details", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-058).

22. Drawing #122871E, Revision 6, 'Vessel Attachment Details", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-057).

23. Drawing #142115E, Revision 3, "Shell Segment Assembly Course #2", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-104).

24. Drawing #122863E, Revision 5, "Shell Flange Details", Babcock & wilcox Company, Mt Vernon, Indiana (GE VPF #1805-107).

25. Drawing #122875E, Revision 2, 'Refueling Containment Skirt, Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-050).

26. Drawing #122873E, Revision 1, Vessel Insulation Support", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-068).

27. Drawing #122874E, Revision 2, "Vessel Thermocouple Pads", Babcock & Wilcox Company, Mt Vernon, Indiana (GE VPF #1805-069).

A-5

GE Nuclear Energy GE-NE-0000-0013-3193-01a-Rl Non-Proprietary Version APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-N Et-O00-0013-3193-01a-RI Non-Proprietary Version TABLE B-1. Browns Ferry Unit 2 P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 °F/hr for Curve A For Figures 5-1. 5-2, 5-3, 5-5, 5-6, 5-7 & 5-10 U

10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 84.2 86.9 89.5 91.9 94.2 96.3 98.3 100.3 102.1 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 B-2

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-1. Browns Ferry Unit 2 P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 15 "F/hr for Curve A For Figures 5-1, 5-2,5-3, 5-5, 5-6, 5-7 & 5-10 j.VE$$

L

~ 3 LIU 240 68.0 83.0 83.0 68.0 103.9 83.0 250 68.0 83.0 83.0 68.0 105.6 83.0 260 68.0 83.0 83.0 68.0 107.2 83.0 270 68.0 83.0 83.0 68.0 108.8 83.0 280 68.0 83.0 83.0 68.0 110.3 83.0 290 68.0 83.0 83.0 68.0 111.8 83.0 300 68.0 83.0 83.0 68.0 113.2 84.0 310 68.0 83.0 83.0 68.0 114.5 89.2 312.5 68.0 83.0 83.0 68.0 114.9 90A 312.5 68.0 113.0 113.0 68.0 143.0 143.0 320 68.0 113.0 113.0 68.0 143.0 143.0 330 6B.0 113.0 113.0 68.0 143.0 143.0 340 68.0 113.0 113.0 68.0 143.0 143.0 350 68.0 113.0 113.0 68.0 143.0 143.0 360 68.0 113.0 113.0 68.0 143.0 143.0 370 68.0 113.0 113.0 68.0 143.0 143.0 380 68.0 113.0 113.0 68.0 143.0 143.0 390 68.0 113.0 113.0 68.0 143.0 143.0 400 68.0 113.0 113.0 68.0 143.0 143.0 410 68.0 113.0 113.0 68.0 143.0 143.0 420 68.0 113.0 113.0 68.0 143.0 143.0 430 68.0 113.0 113.0 68.0 143.0 143.0 440 68.0 113.0 113.0 68.0 143.0 143.0 450 68.0 113.0 113.0 68.0 143.0 143.0 460 68.0 113.0 113.0 68.0 143.0 143.0 470 68.0 113.0 113.0 68.0 143.0 143.0 B-3

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version TABLE B-1. Browns Ferry Unit 2 P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 15 'F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5. 5-6, 5-7 & 5-10

~~~~WEL B~LUINE Vw~

e.3LU 480 68.0 113.0 113.0 68.0 143.0 143.0 490 68.0 113.0 113.0 68.0 143.0 143.0 500 68.0 113.0 113.0 68.0 143.0 143.0 510 68.0 113.0 113.0 68.0 143.0 143.6 520 68.0 113.0 113.0 68.2 143.0 145.2 530 68.0 113.0 113.0 70.2 143.0 146.8 540 68.0 113.0 113.0 72.1 143.0 148.3 550 68.0 113.0 113.0 73.9 143.0 140.8 560 68.0 113.0 113.0 75.7 143.0 151.3 570 68.0 113.0 113.0 77.4 143.0 152.7 580 68.0 113.0 113.0 79.0 143.0 154.0 590 68.0 113.0 113.0 80.6 143.0 155.4 600 68.0 113.0 113.0 82.2 143.0 156.7 610 68.0 113.0 113.0 83.7 143.0 157.9 620 68.0 113.0 113.0 85.1 143.0 159.1 630 68.0 113.0 113.0 66.5 143.4 160.3 640 68.0 113.0 113.0 87.9 143.8 161.5 650 68.0 113.0 114.9 89.2 144.2 162.7 660 68.0 113.0 117.1 90.5 144.7 163.8 670 68.0 113.0 119.2 91.8 145.1 164.9 680 68.0 113.0 121.2 93.1 145.5 165.9 690 68.0 113.0 123.1 94.3 145.9 167.0 700 69.2 113.0 124.9 95.4 146.3 168.0 710 70.7 113.0 126.7 96.6 146.7 169.0 720 72.1 113.0 128.4 97.7 147.1 170.0 730 73.5 113.0 130.1 98.8 147.5 171.0 B-4

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-1. Browns Ferry Unit 2 P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 15 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, 5-7 & 5-10 A

tgg u!~

740 74.8 113.0 131.7 99.9 147.9 171.9 750 76.1 113.0 133.2 101.0 148.2 172.9 760 77.4 113.0 134.7 102.0 148.6 173.8 770 78.6 113.0 136.2 103.0 149.0 174.7 780 79.8 113.0 137.6 104.0 149.4 175.6 790 81.0 113.0 139.0 105.0 149.8 176.4 800 82.2 113.0 140.3 105.9 150.1 177.3 810 83.3 113.0 141.6 106.9 150.5 178.1 820 84.4 113.4 142.9 107.8 150.9 178.9 830 85.5 114.1 144.2 108.7 151.2 179.8 840 86.5 114.8 145A 109.6 151.6 180.6 850 87.6 115.5 146.6 110A 151.9 181.3 860 88.6 116.2 147.7 111.3 152.3 182.1 870 89.6 116.9 148.9 112.1 152.6 182.9 880 90.5 117.6 150.0 113.0 153.0 183.6 890 91.5 118.3 151.0 113.8 153.3 184.4 900 92.4 118.9 152.1 114.6 153.7 185.1 910 93.4 119.6 153.1 115.4 154.0 185.6 920 94.3 120.2 154.2 116.1 154.4 186.5 930 95.1 120.9 155.2 116.9 154.7 187.2 940 96.0 121.5 156.1 117.7 155.0 187.9 950 96.9 122.1 157.1 118.4 155.4 188.6 960 97.7 122.7 158.0 119.1 155.7 189.3 970 98.6 123.3 159.0 119.9 156.0 189.9 980 99.4 123.9 159.9 120.6 156.4 190.6 990 100.2 124.5 160.8 121.3 156.7 191.2 B-5

GE Nuclear Energy GE-WE-0000-0013-3193-01le-R1 Non-Proprietary Version TABLE B-1. Browns Ferry Unit 2 P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 15 "F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, 5-7 & 5-10

..P ~HEA.:.~ws.

~~N 1000 101.0 125.1 161.6 122.0 157.0 191.8 1010 101.7 125.7 162.5 122.6 157.3 192.5 1020 102.5 126.2 163.3 123.3 157.6 193.1 1030 103.3 126.8 164.2 124.0 158.0 193.7 1040 104.0 127.4 165.0 124.6 158.3 194.3 1050 104.7 127.9 165.8 125.3 158.6 194.9 1060 105.4 128.5 166.6 125.9 158.9 195.5 1064 105.7 128.7 166.9 126.2 159.0 195.7 1070 106.2 129.0 167.4 126.5 159.2 196.1 1080 106.9 129.5 168.1 127.2 159.5 198.7 1090 107.6 130.1 168.9 127.8 159.8 197.2 1100 108.2 130.6 169.6 128.4 160.1 197.8 1105 108.6 130.8 170.0 128.7 160.3 198.1 1110 108.9 131.1 170.4 129.0 160.4 198.4 1120 109.6 131.6 171.1 129.6 160.7 198.9 1130 110.2 132.1 171.8 130.2 161.0 199.4 1140 110.9 132.6 172.5 130.7 161.3 200.0 1150 111.5 133.1 173.2 131.3 161.6 200.5 1160 112.1 133.6 173.9 131.9 161.9 201.0 1170 112.8 134.1 174.5 132.4 162.2 201.6 1180 113.4 134.6 175.2 133.0 162.5 202.1 1190 114.0 135.1 175.9 133.5 162.7 202.6 1200 114.6 135.5 176.5 134.1 163.0 203.1 1210 115.2 136.0 177.2 134.6 163.3 203.6 1220 115.8 136.5 177.8 135.2 163.6 204.1 1230 116.3 136.9 178.4 135.7 163.9 204.6 B-6

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-1. Browns Ferry Unit 2 P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 ¶F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, 5-7 & 5-10 1240 116.9 137.4 179.0 136.2 164.2 205.1 1250 117.5 137.8 179.8 136.7 164.4 205.6 1260 118.0 138.3 180.2 137.2 164.7 206.0 1270 118.6 138.7 180.8 137.7 165.0 206.5 1280 119.1 139.2 181.4 138.2 165.2 207.0 1290 119.7 139.6 182.0 138.7 165.5 207.5 1300 120.2 140.0 182.6 139.2 165.8 207.9 1310 120.7 140.5 183.1 139.7 166.1 208.4 1320 121.3 140.9 183.7 140.2 166.3 208.8 1330 121.8 141.3 184.2 140.6 166.6 209.3 1340 122.3 141.7 184.8 141.1 166.8 209.7 1350 122.8 142.1 185.3 141.6 167.1 210.2 1360 123.3 142.6 185.9 142.0 167.4 210.6 1370 123.8 143.0 186.4 142.5 167.6 211.0 1380 124.3 143.4 186.9 142.9 167.9 211.4 1390 124.8 143.8 187.5 143.4 168.1 211.9 1400 125.3 144.2 188.0 143.8 168.4 212.3 B-7

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version TABLE B-2. Browns Ferry Unit 2 Composite P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 15 ¶ F/hr for Curve A For Figures 5-4. 5-8, 5-9 & 5-10 U

10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 0o.u 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 oo.u 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 go.U 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 84.2 86.9 89.5 91.9 94.2 96.3 98.3 100.3 O8.0 83.0 83.0 83.0 83.0 83.0 84.0 91.2 97.2 102.3 106.8 110.9 114.7 118.2 121.4 124.2 126.9 129.5 131.9 1342 136.3 138.3 140.3 B-8

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-R1 Non-Proprietary Version TABLE B-2. Browns Ferry Unit 2 Composite P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-4, 5-8, 5-9 & 5-10 AT Ii~~~~~~~AP~

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230 68.0 83.0 68.0 102.1 142.1 240 68.0 83.0 68.0 103.9 143.9 250 68.0 83.0 68.0 105.6 145.6 260 68.0 83.0 68.0 107.2 147.2 270 68.0 83.0 68.0 108.8 148.8 280 68.0 83.0 68.0 110.3 150.3 290 68.0 83.0 68.0 111.8 151.8 300 68.0 83.0 68.0 113.2 153.2 310 68.0 83.0 68.0 114.5 154.5 312.5 68.0 83.0 68.0 114.9 154.9 312.5 68.0 113.0 68.0 143.0 183.0 320 68.0 113.0 68.0 143.0 183.0 330 68.0 113.0 68.0 143.0 183.0 340 68.0 113.0 68.0 143.0 183.0 350 68.0 113.0 68.0 143.0 183.0 360 68.0 113.0 68.0 143.0 183.0 370 68.0 113.0 68.0 143.0 183.0 380 68.0 113.0 68.0 143.0 183.0 390 68.0 113.0 68.0 143.0 183.0 400 68.0 113.0 68.0 143.0 183.0 410 68.0 113.0 68.0 143.0 183.0 420 68.0 113.0 68.0 143.0 183.0 430 68.0 113.0 68.0 143.0 183.0 440 68.0 113.0 68.0 143.0 183.0 450 68.0 113.0 68.0 143.0 183.0 B-9

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-2. Browns Ferry Unit 2 Composite P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 15 ¶F/hr for Curve A For Figures 5-4, 5-8, 5-9 & 5-10 HEAD ~PE..L; AT UA0~ETNA Lh4AM 460 68.0 113.0 68.0 143.0 183.0 470 68.0 113.0 68.0 143.0 183.0 480 68.0 113.0 68.0 143.0 183.0 490 68.0 113.0 68.0 143.0 183.0 500 68.0 113.0 68.0 143.0 183.0 510 68.0 113.0 68.0 143.6 183.6 520 68.0 113.0 68.2 145.2 185.2 530 68.0 113.0 70.2 146.8 186.8 540 68.0 113.0 72.1 148.3 188.3 550 68.0 113.0 73.9 149.8 189.8 560 68.0 113.0 75.7 151.3 191.3 570 68.0 113.0 77.4 152.7 192.7 580 68.0 113.0 79.0 154.0 194.0 590 68.0 113.0 80.6 155.4 195.4 600 68.0 113.0 82.2 156.7 196.7 610 68.0 113.0 83.7 157.9 197.9 620 68.0 113.0 85.1 159.1 199.1 630 68.0 113.0 86.5 160.3 200.3 640 68.0 113.0 87.9 161.5 201.5 650 68.0 114.9 89.2 162.7 202.7 660 68.0 117.1 90.5 163.8 203.8 670 68.0 119.2 91.8 164.9 204.9 680 68.0 121.2 93.1 165.9 205.9 690 68.0 123.1 94.3 167.0 207.0 700 69.2 124.9 95.4 168.0 208.0 B-10

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-2. Browns Ferry Unit 2 Composite P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 OFihr for Curve A For Figures 5-4, 5-8, 5-9 & 5-10

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710 70.7 126.7 96.6 169.0 209.0 720 72.1 128.4 97.7 170.0 210.0 730 73.5 130.1 98.8 171.0 211.0 740 74.8 131.7 99.9 171.9 211.9 750 76.1 133.2 101.0 172.9 212.9 760 77.4 134.7 102.0 173.8 213.8 770 78.6 136.2 103.0 174.7 214.7 780 79.8 137.6 104.0 175.6 215.6 790 81.0 139.0 105.0 176A 216.4 800 82.2 140.3 105.9 177.3 217.3 810 83.3 141.6 106.9 178.1 218.1 820 84.4 142.9 107.8 178.9 218.9 830 85.5 144.2 108.7 179.8 219.8 840 86.5 145.4 109.6 180.6 220.6 850 87.6 146.6 110.4 181.3 221.3 860 88.6 147.7 111.3 182.1 222.1 870 89.8 148.9 112.1 182.9 222.9 880 90.5 150.0 113.0 183.6 223.6 890 91.5 151.0 113.8 184.4 224.4 g00 92.4 152.1 114.6 185.1 225.1 910 93.4 153.1 115.4 185.8 225.8 920 94.3 154.2 116.1 186.5 226.5 930 95.1 155.2 116.9 187.2 227.2 940 96.0 156.1 117.7 187.9 227.9 950 96.9 157.1 118.4 188.6 228.6 B-11

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Verslon TABLE B-2. Browns Ferry Unit 2 Composite P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-4, 5-8, 5-9 & 5-10 x~

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23pF 960 97.7 158.0 119.1 189.3 229.3 970 98.6 159.0 11990 18990 2299 980 0994 159.9 120.6 190.6 230.6 990 100.2 160.8 121.3 191.2 231.2 1000 101.0 161.6 122.0 191.8 231.8 1010 101.7 162.5 122.6 192.5 232.5 1020 102.5 163.3 123.3 193.1 233.1 1030 103.3 164.2 124.0 193.7 233.7 1040 104.0 165.0 124.6 194.3 234.3 1050 104.7 165.8 125.3 104.9 234.9 1060 105.4 166.6 125.9 195.5 235.5 1064 105.7 166.9 126.2 195.7 235.7 1070 106.2 167.4 126.5 196.1 238.1 1080 106.9 168.1 127.2 196.7 236.7 1090 107.6 168.9 127.8 197.2 237.2 1100 108.2 169.6 128.4 197.8 237.8 1105 108.6 170.0 128.7 198.1 238.1 1110 108.9 170.4 129.0 198.4 238.4 1120 109.6 171.1 129.6 198.9 238.9 1130 110.2 171.8 130.2 199.4 239.4 1140 110.9 172.5 130.7 200.0 240.0 1150 111.5 173.2 131.3 200.5 240.5 1160 112.1 173.9 131.9 201.0 241.0 1170 112.8 174.5 132.4 201.6 241.6 1180 113.4 175.2 133.0 202.1 242.1 B-12

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-2. Browns Ferry Unit 2 Composite P-T Curve Values for 23 EFPY Required Coolant Temperatures at 100 "F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-4, 5-8, 5-9 & 5-10 2MFFY

M3E2 PY.

1190 114.0 175.9 133.5 202.6 242.6 1200 114.6 176.5 134.1 203.1 243.1 1210 115.2 177.2 134.6 203.6 243.6 1220 115.8 177.8 135.2 204.1 244.1 1230 116.3 178.4 135.7 204.6 244.6 1240 116.9 179.0 136.2 205.1 245.1 1250 117.5 179.6 136.7 205.6 245.6 1260 118.0 180.2 137.2 206.0 246.0 1270 118.6 180.8 137.7 206.5 246.5 1280 119.1 181.4 138.2 207.0 247.0 1290 119.7 182.0 138.7 207.5 247.5 1300 120.2 182.6 139.2 207.9 247.9 1310 120.7 183.1 139.7 208.4 248.4 1320 121.3 183.7 140.2 208.8 248.8 1330 121.8 184.2 140.6 209.3 249.3 1340 122.3 184.8 141.1 209.7 249.7 1350 122.8 185.3 141.6 210.2 250.2 1360 123.3 185.9 142.0 210.6 250.6 1370 123.8 186.4 142.5 211.0 251.0 1380 124.3 186.9 142.9 211A 251.4 1390 124.8 187.5 143.4 211.9 251.9 1400 125.3 188.0 143.8 212.3 252.3 B-13

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-3. Browns Ferry Unit 2 P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 15 F/hr for Curve A For Figures 5-1 1, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 842 86.9 89.5 91.9 94.2 96.3 98.3 100.3 102.1 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 B-14

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Ri Non-Proprietary Version TABLE B-3. Browns Ferry Unit 2 P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 "FJhr for Curves B & C and 15 *F/hr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 (l~~~~~~~$1G}

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("F 240 65.0 83.0 83.0 68.0 103.9 83.0 250 68.0 83.0 83.0 68.0 105.6 83.0 260 68.0 83.0 83.0 68.0 107.2 83.0 270 68.0 83.0 83.0 68.0 108.8 83.0 280 68.0 83.0 83.0 68.0 110.3 84.7 290 68.0 83.0 83.0 68.0 111.8 91.2 300 68.0 83.0 83.0 68.0 113.2 97.0 310 68.0 83.0 83.0 68.0 114.5 102.2 312.5 68.0 83.0 83.0 68.0 114.9 103.4 312.5 68.0 113.0 113.0 68.0 143.0 143.0 320 68.0 113.0 113.0 68.0 143.0 143.0 330 68.0 113.0 113.0 68.0 143.0 143.0 340 68.0 113.0 113.0 68.0 143.0 143.0 350 68.0 113.0 113.0 68.0 143.0 143.0 360 68.0 113.0 113.0 68.0 143.0 143.0 370 68.0 113.0 113.0 68.0 143.0 143.0 380 68.0 113.0 113.0 68.0 143.0 143.0 390 68.0 113.0 113.0 68.0 143.0 143.0 400 68.0 113.0 113.0 68.0 143.0 143.0 410 68.0 113.0 113.0 68.0 143.0 143.0 420 68.0 113.0 113.0 68.0 143.0 143.0 430 68.0 113.0 113.0 68.0 143.0 143.0 440 68.0 113.0 113.0 68.0 143.0 143.4 450 68.0 113.0 113.0 68.0 143.0 145.5 460 68.0 113.0 113.0 68.0 143.0 147.5 470 68.0 113.0 113.0 68.0 143.0 149.5 B-15

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Ri Non-Proprietary Version TABLE B-3. Browns Ferry Unit 2 P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 15 °F/hr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 X."~~~~~

480 68.0 113.0 113.0 68.0 143.0 151.A 490 68.0 113.0 113.0 68.0 143.0 153.2 500 68.0 113.0 113.0 68.0 143.0 154.9 510 68.0 113.0 113.0 68.0 143.0 156.6 520 68.0 113.0 113.0 68.2 143.0 158.2 530 68.0 113.0 113.0 70.2 143.0 159.8 540 68.0 113.0 113.0 72.1 143.0 161.3 550 68.0 113.0 113.0 73.9 143.0 162.8 560 68.0 113.0 113.0 75.7 143.0 164.3 570 68.0 113.0 113.0 77.4 143.0 165.7 580 68.0 113.0 113.0 79.0 143.0 167.0 590 68.0 113.0 113.0 80.6 143.0 168.4 600 68.0 113.0 115.5 82.2 143.0 169.7 610 68.0 113.0 118.2 83.7 143.0 170.9 620 68.0 113.0 120.8 85.1 143.0 172.1 630 68.0 113.0 123.3 86.5 143.4 173.3 640 68.0 113.0 125.7 87.9 143.8 174.5 650 68.0 113.0 127.9 89.2 144.2 175.7 660 68.0 113.0 130.1 90.5 144.7 176.8 670 68.0 113.0 132.2 91.8 145.1 177.9 680 68.0 113.0 134.2 93.1 145.5 178.9 690 68.0 113.0 136.1 94.3 145.9 180.0 700 69.2 113.0 137.9 95.4 146.3 181.0 710 70.7 113.0 139.7 96.6 146.7 182.0 720 72.1 113.0 141.4 97.7 147.1 183.0 730 73.5 113.0 143.1 98.8 147.5 184.0 B-16

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-3. Browns Ferry Unit 2 P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 740 74.8 113.0 144.7 99.9 147.9 184.9 750 76.1 113.0 146.2 101.0 148.2 185.9 760 77.4 113.0 147.7 102.0 148.6 186.8 770 78.6 113.0 149.2 103.0 149.0 187.7 780 79.8 113.0 150.6 104.0 149.4 188.6 790 81.0 113.0 152.0 105.0 149.8 189.4 800 82.2 113.0 153.3 105.9 150.1 190.3 810 83.3 113.0 154.6 106.9 150.5 191.1 820 84.4 113.4 155.9 107.8 150.9 191.9 830 85.5 114.1 157.2 108.7 151.2 192.8 840 86.5 114.8 158.4 109.6 151.6 193.6 850 87.6 115.5 159.6 110.4 151.9 194.3 860 88.6 116.2 160.7 111.3 152.3 195.1 870 89.6 116.9 161.9 112.1 152.6 195.9 880 90.5 117.6 163.0 113.0 153.0 196.6 890 91.5 118.3 164.0 113.8 153.3 197.4 900 92.4 118.9 165.1 114.6 153.7 198.1 910 93.4 119.6 166.1 115.4 154.0 198.8 920 94.3 120.2 167.2 116.1 154.4 199.5 930 95.1 120.9 168.2 116.9 154.7 200.2 940 96.0 121.5 169.1 117.7 155.0 200.9 950 96.9 122.1 170.1 118.4 155.4 201.6 960 97.7 122.7 171.0 119.1 155.7 202.3 970 98.6 123.3 172.0 119.9 156.0 202.9 980 99.4 123.9 172.9 120.6 156.4 203.6 990 100.2 124.5 173.8 121.3 156.7 204.2 B-17

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version TABLE B-3. Browns Ferry Unit 2 P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 @F/hr for Curves B & C and 15 *F/hrfor Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 BC~~~ITOM i' UPE Q VP 0T PER-E 1000 101.0 125.1 174.6 122.0 157.0 204.8 1010 101.7 125.7 175.5 122.6 157.3 205.5 1020 102.5 126.2 176.3 123.3 157.6 206.1 1030 103.3 126.8 177.2 124.0 158.0 206.7 1040 104.0 127.4 178.0 124.6 158.3 207.3 1050 104.7 127.9 178.8 125.3 158.6 207.9 1060 105.4 128.5 179.6 125.9 158.9 208.5 1064 105.7 128.7 179.9 126.2 159.0 208.7 1070 106.2 129.0 180.4 126.5 159.2 209.1 1080 106.9 129.5 181.1 127.2 159.5 209.7 1090 107.6 130.1 181.9 127.8 159.8 210.2 1100 108.2 130.6 182.6 128.4 160.1 210.8 1105 108.6 130.8 183.0 128.7 160.3 211.1 1110 108.9 131.1 183.4 129.0 160A 211.4 1120 109.6 131.6 184.1 129.6 160.7 211.9 1130 110.2 132.1 184.8 130.2 161.0 212.4 1140 110.9 132.6 185.5 130.7 161.3 213.0 1150 111.5 133.1 186.2 131.3 161.6 213.5 1160 112.1 133.6 186.9 131.9 161.9 214.0 1170 112.8 134.1 187.5 132.4 162.2 214.6 1180 113.4 134.6 188.2 133.0 162.5 215.1 1190 114.0 135.1 188.9 133.5 162.7 215.6 1200 114.6 135.5 189.5 134.1 163.0 216.1 1210 115.2 136.0 190.2 134.6 163.3 216.6 1220 115.8 136.5 190.8 135.2 163.6 217.1 1230 116.3 136.9 191.4 135.7 163.9 217.6 B-18

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-3. Browns Ferry Unit 2 P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20

.I;KQ "EFP "01TOM 1240 116.9 137.4 192.0 136.2 164.2 218.1 1250 117.5 137.8 192.6 136.7 164.4 218.6 1260 118.0 138.3 193.2 137.2 164.7 219.0 1270 118.6 138.7 193.8 137.7 165.0 219.5 1280 119.1 139.2 194.4 138.2 165.2 220.0 1290 119.7 139.6 195.0 138.7 165.5 220.5 1300 120.2 140.0 195.6 139.2 165.8 220.9 1310 120.7 140.5 196.1 139.7 166.1 221.4 1320 121.3 140.9 196.7 140.2 166.3 221.8 1330 121.8 141.3 197.2 140.6 166.6 222.3 1340 122.3 141.7 197.8 141.1 166.8 222.7 1350 122.8 142.1 198.3 141.6 167.1 223.2 1360 123.3 142.6 198.9 142.0 167.4 223.6 1370 123.8 143.0 199.4 142.5 167.6 224.0 1380 124.3 143.4 199.9 142.9 167.9 224.4 1390 124.8 143.8 200.5 143.4 168.1 224.9 1400 125.3 144.2 201.0 143.8 168.4 225.3 B-19

GE Nuclear Energy GE-NE-0000-0013-3193-01a-RI Non-Proprietary Version TABLE B4. Browns Ferry Unit 2 Composite P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 °F/hrforCurvesB& C and 15 FihrforCurveA For Figures 5-14, 5-18, 5-19 & 5-20 10 20 30 40 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 68.0 53.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 83.0 84.2 86.9 89.5 91.9 94.2 96.3 98.3 100.3 83.0 83.0 83.0 83.0 83.0 84.0 91.2 97.2 102.3 106.8 110.9 114.7 118.2 121.4 124.2 126.9 129.5 131.9 134.2 136.3 138.3 140.3 B-20

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-4. Browns Ferry UnI 2 Composite P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 °F/hrfor Curves B & C and 15 *F/hrfor Curve A For Figures 5-14, 5-18, 5-19 & 5-20 230 68.0 83.0 68.0 102.1 142.1 240 68.0 83.0 68.0 103.9 143.9 250 68.0 83.0 68.0 105.6 145.6 260 68.0 83.0 68.0 107.2 147.2 270 68.0 83.0 68.0 108.8 148.8 280 68.0 83.0 68.0 110.3 150.3 290 68.0 83.0 68.0 111.8 151.8 300 68.0 83.0 68.0 113.2 153.2 310 68.0 83.0 68.0 114.5 154.5 312.5 68.0 83.0 68.0 114.9 154.9 312.5 68.0 113.0 68.0 143.0 183.0 320 68.0 113.0 68.0 143.0 183.0 330 68.0 113.0 68.0 143.0 183.0 340 68.0 113.0 68.0 143.0 183.0 350 68.0 113.0 68.0 143.0 183.0 360 68.0 113.0 68.0 143.0 183.0 370 68.0 113.0 68.0 143.0 183.0 380 68.0 113.0 68.0 143.0 183.0 390 68.0 113.0 68.0 143.0 183.0 400 68.0 113.0 68.0 143.0 183.0 410 68.0 113.0 68.0 143.0 183.0 420 68.0 113.0 68.0 143.0 183.0 430 68.0 113.0 68.0 143.0 183.0 440 68.0 113.0 68.0 143.4 183.4 450 68.0 113.0 68.0 145.5 185.5 B-21

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B4. Browns Ferry Unit 2 Composite P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 0F/hrfor Curves B & C and 15 eF/hrforCurveA For Figures 5-14, 5-18, 5-19 & 5-20 P~30~PY 460 68.0 113.0 68.0 1475 1875 470 68.0 113.0 68.0 149.5 189.5 480 68.0 113.0 68.0 151.4 191.4 490 68.0 113.0 68.0 153.2 193.2 500 68.0 113.0 68.0 154.9 194.9 510 68.0 113.0 68.0 156.6 196.6 520 68.0 113.0 68.2 158.2 198.2 530 68.0 113.0 70.2 159.8 199.8 540 68.0 113.0 72.1 161.3 201.3 550 68.0 113.0 73.9 162.8 202.8 560 68.0 113.0 75.7 164.3 204.3 570 68.0 113.0 77.4 165.7 205.7 580 68.0 113.0 79.0 167.0 207.0 590 68.0 113.0 80.6 168.4 208A 600 68.0 115.5 82.2 169.7 209.7 610 68.0 118.2 83.7 170.9 210.9 620 68.0 120.8 85.1 172.1 212.1 630 68.0 123.3 86.5 173.3 213.3 640 68.0 125.7 87.9 174.5 214.5 650 68.0 127.9 89.2 175.7 215.7 660 68.0 130.1 90.5 176.8 216.8 670 68.0 132.2 91.8 177.9 217.9 680 68.0 134.2 93.1 178.9 218.9 690 68.0 136.1 94.3 180.0 220.0 700 69.2 137.9 95.4 181.0 221.0 B-22

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B4. Browns Ferry Unit 2 Composite P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 "F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-14,5-18, 5-19 & 5-20

.v~~~~~~~~~a~~~~p"~~~~I M AT i=

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710 70.7 139.7 96.6 182.0 222.0 720 72.1 141.4 97.7 183.0 223.0 730 73.5 143.1 98.8 184.0 224.0 740 74.8 144.7 99.9 184.9 224.9 750 76.1 146.2 101.0 185.9 225.9 760 77.4 147.7 102.0 186.8 226.8 770 78.6 149.2 103.0 187.7 227.7 780 79.8 150.6 104.0 188.6 228.6 790 81.0 152.0 105.0 189.4 229.4 800 82.2 153.3 105.9 190.3 230.3 810 83.3 154.6 106.9 191.1 231.1 820 84.4 155.9 107.8 191.9 231.9 830 85.5 157.2 108.7 192.8 232.8 840 86.5 158.4 109.6 193.6 233.6 850 87.6 159.6 110.4 194.3 234.3 860 88.6 160.7 111.3 195.1 235.1 870 89.6 161.9 112.1 195.9 235.9 880 90.5 163.0 113.0 196.6 236.6 890 91.5 164.0 113.8 197A 237.4 900 92.4 165.1 114.6 198.1 238.1 910 93.4 166.1 115.4 198.8 238.8 920 94.3 167.2 116.1 199.5 239.5 930 95.1 1682 116.9 200.2 240.2 940 96.0 169.1 117.7 200.9 240.9 950 96.9 170.1 118.4 201.6 241.6 B-23

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE B-4. Browns Ferry Unit 2 Composite P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 *F/hr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20 960 97.7 171.0 119.1 202.3 242.3 970 98.6 172.0 119.9 202.9 242.9 980 99.4 172.9 120.6 203.6 243.6 990 100.2 173.8 121.3 204.2 244.2 1000 101.0 174.6 122.0 204.8 244.8 1010 101.7 175.5 122.6 205.5 245.5 1020 102.5 176.3 123.3 206.1 246.1 1030 103.3 177.2 124.0 206.7 246.7 1040 104.0 178.0 124.6 207.3 247.3 1050 104.7 178.8 125.3 207.9 247.9 1060 105.4 179.6 125.9 208.5 248.5 1064 105.7 179.9 126.2 208.7 248.7 1070 106.2 180.4 126.5 209.1 249.1 1080 106.9 181.1 127.2 209.7 249.7 1090 107.6 181.9 127.8 210.2 250.2 1100 108.2 182.6 128.4 210.8 250.8 1105 108.6 183.0 128.7 211.1 251.1 1110 108.9 183.4 129.0 211.4 251.4 1120 109.6 184.1 129.6 211.9 251.9 1130 110.2 184.8 130.2 212.4 252.4 1140 110.9 185.5 130.7 213.0 253.0 1150 111.5 186.2 131.3 213.5 253.5 1160 112.1 186.9 131.9 214.0 254.0 1170 112.8 187.5 132.4 214.6 254.6 1180 113.4 188.2 133.0 215.1 255.1 B-24

GE Nuclear Energy GE-NE-0000-0013-3193-Ole-Rl Non-Proprietary Version TABLE B4. Browns Ferry Unit 2 Composite P-T Curve Values for 30 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 Flhr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20 1190 114.0 188.9 133.5 215.6 255.6 1200 114.6 189.5 134.1 216.1 256.1 1210 115.2 19002 134.6 216.6 256.6 1220 115.8 190.8 135.2 217.1 257.1 1230 116.3 191.4 135.7 217.6 257.6 1240 116.9 192.0 136.2 218.1 258.1 1250 117.5 192.6 136.7 218.6 258.6 1260 118.0 193.2 137.2 219.0 259.0 1270 118.6 193.8 137.7 219.5 259.5 1280 119.1 194.4 138.2 220.0 260.0 1290 119.7 195.0 138.7 220.5 260.5 1300 120.2 195.6 139.2 220.9 260.9 1310 120.7 196.1 139.7 221.4 261.4 1320 121.3 198.7 140.2 221.8 261.8 1330 121.8 197.2 140.6 222.3 262.3 1340 122.3 197.8 141.1 222.7 262.7 1350 122.8 198.3 141.6 223.2 263.2 1360 123.3 198.9 142.0 223.6 263.6 1370 123.8 189.4 142.5 224.0 264.0 1380 124.3 199.9 142.9 224.4 264.4 1390 124.8 200.5 143.4 224.9 264.9 1400 125.3 201.0 143.8 225.3 265.3 B-25

GE Nuclear Energy GE-NE-OO00-0013-3193-Ola-Rl Non-Proprietary Version APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives Is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

One condition on monitoring the bottom head separately is that it must be demonstrated that the vessel beltline temperature can be accurately monitored during pressure testing.

An experiment has been conducted at a BWR-4 that showed that thermocouples on the vessel near the feedwater nozzles, or temperature measurements of water in the recirculation loops provide good estimates of the beltline temperature during pressure testing. Thermocouples on the RPV flange to shell junction outside surface should be used to monitor compliance with upper vessel curve. Thermocouples on the bottom head outside surface should be used to monitor compliance with bottom head curves. A description of these measurements is given in GE SIL 430, attached in Appendix D.

First, however, it should be determined whether there are significant temperature differences between the beltline region and the bottom head region.

C.2 DETERMINING WHICH CURVE TO FOLLOW The following subsections outline the criteria needed for determining which curve Is governing during different situations. The application of the P-T curves and some of the assumptions inherent in the curves to plant operation is dependent on the proper monitoring of vessel temperatures. A discussion of monitoring of vessel temperatures can be found in Section 4 of the pressure-temperature curve report prepared in 1989 [1].

C-2

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by *15 0F per hour. If the coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.

C.2.2 Curve B: Non-Nuclear HeatuplCooldown Curve B should be used whenever Curve A or Curve C do not apply. In other words, the operator must follow this curve during times when the coolant is heating or cooling faster than 150F per hour during a hydrotest and when the core is not critical.

C.2.3 Curve C: Core Critical Operation The operator must comply with this curve whenever the core is critical. An exception to this principle is for low-level physics tests; Curve B must be followed during these situations.

C.3 REACTOR OPERATION VERSUS OPERATING LMITS For most reactor operating conditions, coolant pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the right of the P-T curves). The operations where P-T curve compliance is typically monitored closely are planned

events, such as vessel bolt-up, leakage testing and startup/shutdown operations, where operator actions can directly influence vessel pressures and temperatures.

The most severe unplanned transients relative to the P-T curves are those that result from SCRAMs, which sometimes include recirculation pump trips.

Depending on operator responses following pump trip, there can be cases where stratification of colder water in the bottom head occurs while the vessel pressure is still relatively high.

Experience with such events has shown that operator action is necessary to avoid P-T curve exceedance, but there is adequate time for operators to respond.

C-3

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:

Head flange bolt-up Leakage test (Curve A compliance)

Startup (coolant temperature change of less than or equal to 100OF in one hour period heatup)

Shutdown (coolant temperature change of less than or equal to 1000F in one hour period cooldown)

Recirculation pump trip, bottom head stratification (Curve B compliance)

Ca4

GE Nuclear Energy GE-NE-0000-0013-3193-012-Ri Non-Proprietary Version APPENDIX C

REFERENCES:

1. T.A. Caine, "Pressure-Temperature Curves Per Regulatory Guide 1.99, Revision 2 for the Dresden and Quad Cities Nuclear Power Stations', SASR 89-54, Revision 1, August 1989.

C-5

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version APPENDIX D GE SIL 430 D-A

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions concerning primary and alternate reactor pressure vessel (RPV) temperature monitoring measurements for complying with RPV brittle fracture and thermal stress requirements.

As such, the purpose of this Service Information Letter is to provide a summary of RPV temperature monitoring measurements, their primary and alternate uses and their limitations (See the attached table). Of basic concern is temperature monitoring to comply with brittle fracture temperature limits and for vessel thermal stresses during RPV heatup and cooldown. General Electric recommends that BWR owners/operators review this table against their current practices and evaluate any inconsistencies.

TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)

Measurement Steam dome saturation temperature as determined from main steam instrument line pressure Recirc suction line coolant temperature.

Use Primary measurement above 2120F for Tech Spec 100°F/hr heatup and cooldown rate.

Primary measurement below 2120F for Tech Spec 100°F/hr heatup and cooldown rate.

Limitations Must convert saturated steam pressure to temperature.

Must have recirc flow.

Must comply with SIL 251 to avoid vessel stratification.

Alternate measurement above 212 0F.

When above 2120F need to allow for temperature varIations (up to 10-150F lower than steam dome saturation temperature) caused primarily by FW flow variations.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Alternate measurement for RPV drain line temperature (can use to comply with delta T limit between steam dome saturation temperature and bottom head drain line temperature).

Limitations RHR heat exchanger inlet coolant temperature RPV drain line coolant temperature Alternate measurement for Tech Spec 1000F/hr cooldown rate when in shutdown cooling mode.

Primary measurement to comply with Tech Spec delta T limit between steam dome saturated temp and drain line coolant temperature.

Must have previously correlated RHR inlet coolant temperature versus RPV coolant temperature.

Must have drain line flow. Otherwise, lower than actual temperature and higher delta Ts will be indicated Delta T limit is 1000F for BWR/6s and 1450F for earlier BWRs.

Primary measurement to comply with Tech Spec brittle fracture limits during cooldown.

Alternate information only measurement for bottom head inside/

outside metal surface temperatures.

Must have drain line flow. Use to verify compliance with Tech Spec minimum metal temperaturelreactor pressure curves (using drain line temperature to represent bottom head metal temperature).

Must compensate for outside metal temperature lag during heatupfcooldown.

Should have drain line flow.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Closure head flanges outside surface T/Cs Primary measurement for BWRI6s to comply with Tech Spec brittle fracture metal temperature limit for head bolt-up.

Use for metal (not coolant) temperature. Install temporary T/Cs for alternate measurement, if required.

One of two primary measure-ments for BWR16s for hydro test RPV flange-to-shell junction outside surface T/Cs Primary measurement for BWRs earlier than 6s to comply with Tech Spec brittle fracture metal temperature limit for head bolt-up.

Use for metal (not coolant) temperature. Response faster than closure head flange T/Cs.

One of two primary measurements for BWRs earlier than 6s for hydro test. Preferred in lieu of closure head flange T/Cs If available.

Use RPV closure head flange outside surface as alternate measurement.

RPV shell outside surface T/Cs Top head outside surface T/Cs Information only.

Information only.

Slow to respond to RPV coolant changes. Not available on BWR/6s.

Very slow to respond to RPV coolant changes. Not avail-able on BWRI6s.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Bottom head outside surface T/Cs Use I of 2 primary measurements to comply with Tech Spec brittle fracture metal temperature limit for hydro test.

Limitations Should verify that vessel stratification is not present for vessel hydro.

(see SIL No. 251).

Primary measurement to comply with Tech Spec brittle fracture metal temperature limits during heatup.

Use during heatup to verify compliance with Tech Spec metal temperature/reactor pressure curves.

Note: RPV vendor specified metal T limits for vessel heatup and cooldown should be checked during initial plant startup tests when initial RPV vessel heatup and cooldown tests are run.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version Product

Reference:

B21 Nuclear Boiler Prepared By: A.C. Tsang Approved for Issue:

Issued By:

B.H. Eldridge, Mgr.

D.L. Allred, Manager Service Information Customer Service Information and Analysis Notice:

SlLs pertain only to GE BWRs. GE prepares SILs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, if any, of information contained in SlLs to any plant or facility other than GE BWRs as designed and furnished by GE. Determination of applicability of information contained in any SIL to a specific GE BWR and implementation of recommended action are responsibilities of the owner of that GE BWR.SILs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of its owner. Such operation involves activities of which GE has no knowledge and over which GE has no control. Therefore, GE makes no warranty or representation expressed or implied with respect to the accuracy, completeness or usefulness of Information contained in SILs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SlLs.

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GE Nuclear Energy GE-NEE-0000-0013-3193-0la-R1 Non-Proprietary Version APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version 10CFR50, Appendix G defines the beitline region of the reactor vessel as follows:

'The region of 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 establish the value of peak fluence for identification of beltline materials (as discussed above), the 10CFR50 Appendix H fluence value used to determine the need for a surveillance program was used; the value specified is a peak fluence (E>1 MEV) of 1.0e17 n/cm2. Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm2, then It can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.

The following dimensions are obtained from the referenced drawings and are specified as the distance above vessel "U":

Shell # 2 - Top of Active Fuel (TAF) 366.3" [1]

Shell # 1 - Bottom of Active Fuel (BAF) 216.3" [1,2]

Centerline of Recirculation Outlet Nozzle N1 in Shell # 1 161.5" [2,3]

Top of Recirculation Outlet Nozzle N1 in Shell # 1 188.0" (4]

Centerline of Recirculation Inlet Nozzle N2 in Shell # 1 181.0"

[2,3]

Top of Recirculation Inlet Nozzle N2 in Shell # 1 193.5" [4]

Centerline of Instrumentation Nozzle N16 In Shell #2 366.0" [2,3]

Girth Weld between Shell Ring #2 and Shell Ring #3 385.8' [1,5]

From [2], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltine region (the top of the recirculation inlet nozzle is -23" below BAF and the top of the recirculation outlet nozzle is -28" below BAF).

As shown in 12,3], the N16 Instrumentation Nozzle Is contained within the core beltline region; however, this 2" nozzle is fabricated from Alloy 600 materials. As noted in Table A-2, components E-2

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version made from Alloy 600 and/or having a diameter of less than 2.5" do not require fracture toughness evaluations. No other nozzles are within the BAF-TAF region of the reactor vessel. The girth weld between Shell Rings #2 and #3 is -20" above TAF. Therefore, if it can be shown that the peak fluence at these locations is less than 1.0e17 n/cm2, it can be safely concluded that all nozzles and welds, other than those included in Tables 4-4 and 4-5, are outside the beltline region of the reactor vessel.

Based on the axial flux profile for EPU which bounds the pre-EPU axial flux profile, the RPV fluence drops to less than 1.0e17 n/cm2 at -8" below the BAF and at -1 1" above TAF. The beltline region considered In the development of the P-T curves Is adjusted to include the additional I' above the active fuel region and the additional 8" below the active fuel region. This adjusted beltline region extends from 208.3" to 377.3w above reactor vessel 'O' for 30 EFPY.

Based on the above, it is concluded that none of the Browns Ferry Unit 2 reactor vessel plates, nozzles, or welds, other than those included in Tables 4-4 and 4-5, are in the beltline region.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Ri Non-Proprietary Version APPENDIX E

REFERENCES:

1.

Task Data Input Request, "Pressure-Temperature Curves Browns Ferry Units 2&3", V. Schiavone, (TVA), February 25, 2003.

2.

Drawing 886D499, Revision 12, "Reactor Vessel",

General Electric Company, GENE, San Jose, California.

3.

Drawing #254185F, Revision 11, "General Outline", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-042).

4.

Drawing #122864E, Revision 4, uRecirculation Nozzles', Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-041).

5.

Drawing #24187F, Revision 11, 'Vessel Sub-Assembly', Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-059).

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

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GE Nuclear Energy GE-NE-000()-0013-3193-Ola-Rl Non-Proprietary Verslon Paragraph IV.B of IOCFR50 Appendix G [13 sets limits on the upper shelf energy of the beffline materials.

The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be 30 EFPY.

Calculations of 30 EFPY USE, using Regulatory Guide 1.99, Revision 2 [2] methods and BWROG Equivalent Margin Analyses [3, 4, 5] methods are summarized in Tables F-1 and F-2.

Unirradiated upper shelf data was not available for all of the material heats in the Browns Ferry Unit 2 beltline region. Therefore, Browns Ferry Unit 2 is evaluated to verify that the BWROG EMA Is applicable. The USE decrease prediction values from Regulatory Guide 1.99, Revision 2 are used for the beltline components as shown in Tables F-1 and F-2. These calculations are based upon the 30 EFPY peak 1/4T fluence as provided in Tables 44 and 4-5. The surveillance capsule data Is obtained from [6].

Based on the results presented In Tables F-1 and F-2, the USE EMA values for the Browns Ferry Unit 2 reactor vessel beltline materials remain within the limits of Regulatory Guide 1.99, Revision 2 and IOCFR50 Appendix G for 30 EFPY of operation.

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-R1 Non-Proprietary Version Table F-I Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 2 For 30 EFPY (including Extended Power Uprate)

BWR/3-6 PLATE Surveillance Plate (Heat A0981-1) USE:

%Cu

=

0.14 lit Capsule Fluence

=

1.52 x 1W'7 n/cm2 l1 Capsule Measured % Decrease

= 4 (Charpy Curves)

I~ CapsuleR.G. 1.99 Predicted % Decrease

= 9 (R.G. 1.99, Figure 2)

Umitina Betine Plate (Heat C2467-11 USE:

%Cu

=

0.16 30 EFPY 1/4T Fluence

=

9.2 x 1017 n/CM 2 R.G. 1.99 Predicted % Decrease

=

14.5 Adjusted % Decrease

= N/A 14.5% S 21 %, so vessel plates are bounded by equivalent margin analysis (R.G. 1.99, Figure 2)

(R.G. 1.99, Position 2.2)

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GE Nuclear Energy GE-NE-0000-0013-3193-Ola-Rl Non-Proprietary Version Table F-2 Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 2 For 30 EFPY (including Extended Power Uprate)

BWR/2-6 WELD Surveillance Weld (Heat DW5733) USE:

%Cu

=

0.20 I't Capsule Fluence

=

1.52 x 106 nlcmr it Capsule Measured % Decrease

= -3 (increase)

(Charpy Curves) li Capsule R.G. 1.99 Predicted % Decrease

= 13 (R.G. 1.99, Figure 2) iUmiting Beltline Weld (Electrosiag) USE:

%Cu

=

0.24 30 EFPY 114T Fluence

=

9.2 x 1017 n/crm2 R.G. 1.99 Predicted % Decrease

= 22 (R.G. 1.99, Figure 2)

Adjusted% Decrease

=

N/A (R.G. 1.99, Position 2.2) 22°6 34%, so vessel welds are bounded by equivalent margin analysis F-4

GE Nuclear Energy GE-NE-0000-0013-3193-Ola-RI Non-Proprietary Version APPENDIX F

REFERENCES:

1. uFracture Toughness Requirements", Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

2. "Radiation Embrittement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

3. J.T. Wiggins (NRC) to L.A. England (Gulf States Utilities Co.), "Acceptance for Referencing of Topical Report NEDO-32205, Revision 1, '10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWRI2 Through BWRI6 Vessels'", December 8,1993.

4. LA. England (BWR Owners' Group) to Daniel G. McDonald (USNRC), uBWR Owners' Group Topical Report on Upper Shelf Energy Equivalent Margin Analysis -Approved Version", BWROG-94037, March 21, 1994.

5. C.l. Grimes (NRC) to Carl Terry (Niagara Mohawk Power Company),

"Acceptance For Referencing Of EPRI Proprietary Report TR-1 13596, "BWR Vessel And Internals Project, BWR Reactor Pressure Vessel Inspection And Flaw Evaluation Guidelines (BWRVIP-74)" And Appendix A, "Demonstration Of Compliance With The Technical Information Requirements Of The Ucense Renewal Rule (10 CFR 54.21)9, October 18, 2001.

6. C. Oza, "Browns Ferry Steam Electric Station Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis", GE-NE, San Jose, CA, August 1995 (GENE-B1100639-01, Revision 1).

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