ML032540372

Pressure-Temperature Curves for TVA Browns Ferry Units 2 and 3

ML032540372
Person / Time
Site:	Browns Ferry
Issue date:	05/31/2003
From:	Branlund B, Frew B, Tilly L General Electric Co
To:	Office of Nuclear Reactor Regulation
References
TVA-BFN-TS-441 GE-NE-0000-0013-3193-01a, Rev 1, GE-NE-0000-0013-3193-02, Rev 0
Download: ML032540372 (269)

Text

GE Nuclear Energy Engineering and Technology GE-NE-0000-001 3-3193-01 a General Electric Company Revision 0 175 Curtner Avenue Class I San Jose, CA 95125 May 2003 Pressure-Temperature Curves For TVA Browns Ferry Unit 2 Prepared by: Lf ahiy L.J. Tilly, Senior Engineer Structural Mechanics and Materials Verified by: BaD rew B.D. Frew, Principal Engineer Structural Mechanics and Materials Approved by: 0B7 cBranfund B.J. Branlund, Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy GE-NE-0000-0013-3193-01a INFORMATION NOTICE This document is the non-proprietary version of the General Electric Company (GE) document GE-NE-0000-0013-3193-01. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [ ]. Large brackets, as shown on this paragraph, Identify figures and large equation objects that could not be appropriately identified with the brackets shown above.

DISCLAIMER OF RESPONSIBILITY This document was prepared by the General Electric Company (GE) and is furnished solely for the purpose or purposes stated In the transmittal letter. No other use, direct or indirect, of the document or the information It contains Is authorized. Neither GE nor any of the contributors to this document:

• Makes any representation or warranty (express or Implied) as to the completeness, accuracy or usefulness of the information contained In this document or that such use of any information may not Infringe privately owned rights; or

• Assumes any responsibility for liability or damage of any kind that may result from any use of such information.

Copyright, General Electric Company, 2003

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GE Nuclear Energy GE-NE-000001 3-3193-1 a 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 111; 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 Kc 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 fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14b], 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 12]:

• Closure flange region (Region A)

• Core beltline region (Region B)

• Upper vessel (Regions A & B)

• Lower vessel (Regions B & C)

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GE Nuclear Energy GE-NE-0000-0013-31 93-Ola For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 1000F/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 15"F/hr or less must be maintained at all times.

The P-T curves apply for both heatup and cooldown and for both the 14T 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, KIr, at 1/4T to be less than that at 314T 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, beftline (at 23 and 30 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

GE Nuclear Energy GE-NE-0000-0013-3193-01a TABLE OF CONTENTS

1.0 INTRODUCTION

I 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITIAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 14 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 20

5.0 CONCLUSION

S AND RECOMMENDATIONS 50

6.0 REFERENCES

73

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GE Nuclear Energy GE-NE-0000-001 3-3193-01 a TABLE OF APPENDICES APPENDIX A DESCRIPTION OF DISCONTINUITIES APPENDIX B PRESSURE-TEMPERATURE CURVE DATA TABULATION APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS APPENDIX D GE SIL 430 APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

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GE Nuclear Energy GE-NE-0000-0013-31 93-01 a 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. CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 31 FIGURE 4-3. FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 37 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE Al - 23 EFPY [I51F/HR OR LESS COOLANT HEATUP/COOLDOWN] 53 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] - 23 EFPY [151F/HR OR LESS COOLANT HEATUP/COOLDOWN] 54 0

FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 23 EFPY [15 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 55 0

FIGURE 5-4: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 23 EFPY [15 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 23 EFPY

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 57 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 23 EFPY

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 58 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 23 EFPY

[1000F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-8: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 23 EFPY

[100 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 0

FIGURE 5-9: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 23 EFPY [100 F/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 [1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-1 1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] - 30 EFPY

[1 50F/HR OR LESS COOLANT HEATUP/COOLDOWN] 63 FIGURE 5-12: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] - 30 EFPY [150 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64 FIGURE 5-13: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 30 EFPY [15'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 65 FIGURE 5-14: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 30 EFPY [15'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 66

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GE Nuclear Energy GE-NE-0000-0013-3193-01 a FIGURE 5-15: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 30 EFPY

[100 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 67 FIGURE 5-16: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 30 EFPY

[100 0 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

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN 69 FIGURE 5-18: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 30 EFPY

[100 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 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 B] NCLUDING BOTTOM HEAD AND CORE CRITICAL P-T CURVES [CURVE C] UP TO 30 EFPY [100 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 72

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GE Nuclear Energy GE-NE-0000-0013-3193-01 a TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR BROWNS FERRY UNIT 2 VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR BROWNS FERRY UNIT 2 NOZZLE AND WELD MATERIALS 12 TABLE 4-3: RTNDT 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 IOCFR50 APPENDIX G REQUIREMENTS 22 TABLE 4-7: APPLICABLE BWR/4 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 24 TABLE 4-8: APPLICABLE BWR/4 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

GE Nuclear Energy GE-NE-0000-0013-3193-01 a

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 [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC In a SER [14b], 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 Kic 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 10CFR50.55a 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-0000-001 3-3193-01 a methods for calculating ART. The value of ART is a function of RPV 14T 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 2 2 1.0 x 1018 n/cm (23 EFPY) and 1.33 x 1018 n/cm (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-0000-001 3-3193-01 a 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 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], Ineffect 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 M.

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-beitline 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-0000-0013-3193-01 a 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-0000-0013-3193-01 a 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 68F.

The fluence is conservatively calculated using an EPU [14a] flux for the entire plant life.

The flux is calculated in accordance with Regulatory Guide 1.190.

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 1II, 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 requirements 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 l1l, Subsection NB-2300 are as follows:

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

b. RTNDT is defined as the higher of the dropweight NDT or 600 F 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 (ST) of the bolting material, whichever is greater.

10CFR50 Appendix G [8] 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-0000-0013-31 93-01 a converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data In WRC Bulletin 217 [9] 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 [10], and approved by the NRC for generic use 11.

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 [1 3] (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 [12]). 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 2F 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:

T50L = 100F + [ (50 - 40) ft-lb 2F/ft-lb = 300 F 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, TsOT = 300F + 300F = 600F.

GE Nuclear Energy GE-NE-0000-0013-3193-01 a The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5or 601F).

Dropweight testing to establish NDT for plate material Is listed in the CMTR; the NDT for the case above is -200F. Thus, the Initial RTNDT for plate heat C2460-2 is 0F.

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 40"F and the lowest CVN data is 63 ft-lb at 40 0F. The corresponding value of (Thor 600F) Is:

(T5OT - 600F) = [ 40 30 - 600 F = 1F.

Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (Tsor 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 40 0F. The corresponding value of (T5aT - 600F) was:

(Ts5T - 60 0F) = { [40 + (50 - 34) ft-lb 2ft-lb + 300 F }-600 F = 420 F.

Therefore, the initial RTNDT 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 11I, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 60°F 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 10°F. Therefore, the LST for the bolting material is 700F. The highest GE Nuclear Energy GE-NE-0000-0013-31 93-01 a RTNDT in the closure flange region is 23.10 F, for the vertical electroslag weld material in the upper shell. Thus, the higher of the LST and the RTNDT +600F is 83.1 OF, the bolt-up limit Inthe 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 Ingenerating the P-T curves.

GE Nuclear Energy GE-NE-0000-0013-3193-01a

. TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELLCOURSE #5 SHELL COURSE #4 SHELL COURSE #3 TOP OF SHELL COURSE #2 ACTIVE. FUEL (TAF)366.3^

BOTIOM O: SHELL COURSE #1 ACTIVE FUEL (BAF) 216.3"

• BOTTO HEAD SU)PPORTSKIRT 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 belfline region.

Figure 4-1: Schematic of the Browns Ferry Unit 2 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-NE-0000-0013-3193-01a Table 4-1: RTNDT Values for Browns Ferry Unit 2 Vessel Materials Top Head &Flange Shell Flange (W 48) 48-127-2 ARZ 76 10 F 1105 115 1 -20 1o I o Top Head Flange (MK 209) 209-127-2 AKU75 10 F 108 111 106 -20 10 10 Top Head Dollar (MK201) 201-122-1 B5524-2 10 P 42 50 43 -4 10 10 Top Head Side Plates (MK202) 202-127-5 C2426-2 10 P 53 80 79 -20 10 10 202-127-6 C2426-2 10 P 80 69 75 -20 10 10 202-127-7 C2426-3 10 P 91 71 98 -20 10 10 202-127-9 C1717-3 10 P 74 69 95 -20 10 10 02-127-10 C1717-3 10 P 93 51 90 -20 10 10 02-127-11 C1722-3 10 P 81 80 101 -20 10 10 Shell Courses Upper Shell Plates (MK 60) 6-127-1 1 C2559-2 10 P 31 60 54 18 10 18 6-127-21 C2791-1 10 P 63 29 54 22 10 22 6-127-22 C2660-1 10 P 67 70 67 -20 10 10 Transiton Shell Plates (MK 16) 15-127-1 C2533-1 10 P 58 58 72 -20 10 10 15-127-3 C2533-3 10 P 57 67 47 -14 10 10 15-127-4 85842-3 10 P 63 81 46 -12 10 10 Upper Intermediate Shell Plates (MK 59) 6-127-18 C2528-1 10 P 66 79 71 -20 10 10 6-127-23 C2463-2 10 P 69 60 72 -20 10 10 6-127-24 C2605-2 10 P 56 71 57 -20 10 10 Lower Intermediate Shell Plates (MK 58) 6-127-6 A0981-1 10 P 72 68 65 10 -10 6-127-16 C2467-1 10 P 65 81 74 .20 -10 -10 6-127-20 C2849-1 10 P 61 59 74 10 -10 Lower Shell Plates (MK 57) 6-127-14 C2467-2 10 P 59 78 65 20 -20 6-127-15 C2463-1 10 P 85 74 54 20 -20 6-127-17 C2460-2 10 P 59 60 40 0 -20 0 Bottom Head Bottom Head Dollar (MK1) 1-139-1 C2669-2 40 P 34 40 34 42 40 42 Bottom Head Upper Torus (MK 2) 2-139-1 86747-1 40 P 78 80 75 10 40 40 2-139-2 B6747-1 40 P 60 54 77 10 40 40 2-139-3 B6776-2 40 P 90 88 92 10 40 40 2-139-4 86776-2 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)

-127-5 C2412-1 40 P 57 66 68 10 40 40

-127-8 C2412-1 40 P 67 45 76 20 40 40

-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 Nuclear Energy GE-NE-0000-0013-31 93-01 a Table 4-2: RTNDT Values for Browns Ferry Unit 2 Nozzle and Weld Materials N1 Recirc Outlet Nozzle (MK 8) 8-127-3 E31VW 431H-3 40 F 86 I 81 I 100 10 40 40 8-127I4 E30VW 431H-4 40 F 1 84 98 1 97 1 10 1 40 40 N2 Recirc Inlet Nozzle (MK 7) 7-127-11 E25VW 433H-11 40 F 97 106 86 10 40 40 7-127-12 E25VW 433H-12 40 F 63 105 94 10 40 40 7-127-13 E25VW 433H-13 40 F 95 79 101 10 40 40 7-127-14 E25VW 4331-1-14 40 F 93 103 107 10 40 40 7-127-15 E25VW 433H-15 40 F 105 113 98 10 40 40 7-127-16 E25VW 4331-1-16 40 F 100 97 86 10 40 40

-127-17 E25VW 4331--17 40 F 101 94 99 10 40 40

-127-18 E25VW 4331-18 40 F 112 96 85 10 40 40

-127-19 E25VW 43311-19 40 F 107 87 110 10 40 40 7-127-20 E25VW 4331-1-20 40 F 96 99 116 10 40 40 N3 Steam Outlet Nozzle (MK 14) 14-127-5 E26VW 435H-5 40 F 106 113 107 10 40 40 14-127-6 E26VW435H-6 40 F 112 99 116 10 40 40 14-127-7 E26VW 435H-7 40 F 87 99 106 10 40 40 14-127-8 E26VW435H-8 40 F 113 112 95 10 40 40 N4 Feedwater Nozzle (MK1O) 10-127-7 E25VW 436H-7 40 F 108 94 112 10 40 40 10-127-8 E25VW 436H-8 40 F 112 105 113 10 40 40 10-127-9 E25VW 436H-9 40 F 103 112 93 10 40 40 10-127-10 E25VW 436H-10 40 F 89 78 91 10 40 40 10-127-11 E25VW 436H-11 40 F 112 104 119 10 40 40 10-127-12 E25VW 4361-1-12 40 F 111 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 E26VW 437H-3 40 F 73 96 87 10 40 40 N6 Top Head Instrumentation Nozzle (MK 206) 206-145-1& -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 N8 Jet Pump Instrumentation (MK 19) 19-122-3 &-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 AP &Liquid Control Nozzle (MM7) _ __

17-139-1 ZT3043-1 40 F 155 154 156 10 40 40 N11, N12, N16 Instrumentation Nozzle (MK 12) Inronel 12-127-13 through 16 8601 12-139-11&-12 8601 N13,N14 High &Low Pressure Seal Leak (MK139_

IF 40 N15 Drain Nozzle (MK22) 22-139-1 7478 40 F 136 160 172 10 40 40 WELDS:

CylindricalShell Axial Welds Electrosag Welds ESW 23.1 Girh Welds Shell I to Shet 2 (MK57 to MK58) D55733

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

Weld initial RTNDT values obtained from 131.

NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a Table 4-3: RTNDT Values for Browns Ferry Unit 2 Appurtenance and Bolting Materials ISupport Skirt I24-127-5 (MK 24 )

through -8 A4846-5 40 11091961104 10 40 40 Shroud Support (MK 51, MK52, MK53) 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 56782-5 53-127-13 & -16 56782-6 53-127-15 56825-1 Steam Dryer Support Bracket (MK131) Stainless Steel 131 00431 Core Spray Bracket (MK132) Stainless Steel 132 3342230 Dryer Hold Down Bracket (MK133) 133 EV8446 40 94 110 113 10 40 40 Guide Rod Bracket (MKI34) Stainless Steel 134 00431 Feedwater Sparger Brackets -MKI35 Stainless Steel 135 00431 Stabilizer Bracket (MK 196) 196 C6458-1 10 60 59 56 -20 40 40 Surveillance Brackets MK199 & MK200) Stainless Steel 199 200 342633-2 _

Lffting Lugs (K210)_

210 A1210-3 10 98 108 72 -20 0 0 CRD penetrations (MK101 - MK128) Alloy 600 101 through 128 Refueling Containment Skirt (MK71) _._

71-127-5 through -8 B74784B 10 110 89 102 -20 10 10

....... nt:..~ Tm Epnk Closure (MK61) 6730502 10 34 52 68 eta 70 6780382 10 42 42 42 nfa 70 6720443 10 35 38 37 nta 70 NUTS:

Closure (MK62) 6730502 10 34 52 68 Wlea 70 BUSHINGS:

Closure (MK63) T3798 10 61 68 73 51 10 M2513 10 64 65 67 40 10 M3232 10 65 65 68 45 10 WASHERS:

Closure (MK64) 6790156 10 ra W/a n/a n/a 70 Closure (MK64 and MK65) 6730502 10 34 52 68 Iva 70 Closure (MK64) 6780278 10 40 43 43 nta 70 NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-0013-3193-01a 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting beitline 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 (RG1.99) 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 ARTNOT + Margin where, ARTNDT = [CF. f (0.28-010 lgt Margin = 2(a2 + aA2) 0.5 CF = chemistry factor from Tables 1or 2 of RG1.99 f = /4T fluence 1019 Margin = 2(Yl 2 + U7A2)0.5 a, = standard deviation on Initial RTNDT, which is taken to be 0°F (13'F for electroslag welds).

CrA = standard deviation on ARTNDT, 28SF for welds and 17*F for base material, except that a,& need not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT The margin term a, has constant values of 170F for plate and 280F for weld as defined In RG1.99. However, C,& 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-0000-0013-31 93-Ola of an is taken to be 0F 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 113].

The copper (Cu) and nickel (Ni) values were used with Tables I 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 (,) is specified.

4.2.1.2 Fluence An EPU (Extended Power Uprate) 114a] flux for the vessel ID wall was calculated using methods consistent with Regulatory Guide 1.190. The flux in Reference 14 Is determined for the EPU rated power of 3952 MWt.

The peak fast flux for the RPV inner surface from Reference 14 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 14] 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. Including the same bias adjustment as that applied to the RPV, the calculated fast flux at this capsule is 9.5e8 ncm2-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 n/cm2 -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 n/cm2, 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 n/cm2-s to determine the fluence for the entire 30 EFPY (representing the GE Nuclear Energy GE-NE-0000-0013-3193-01a 40 year Browns Ferry Unit 2 license period) is conservative. The RPV peak ID fluence is calculated as follows:

1.4e9 n/cm2-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 "O") [14a]; hence the peak IDfluence 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 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 30 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

1.33e18 n/cm2 exp(-0.24 (6.13 / 4)) = 9.2e17 n/cm2.

The 30 EFPY 1/4T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

1.1e18n/cm2 exp(-0.24 (6.1314))=7.4e17n/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) = 1.0e18 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 / lower shell location ratio of 0.81 for EPU conditions [14a]; hence the peak IDfluence used for these components is 8.2e17 n/cm 2.

The fuence 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 1/4T fluence for the lower-Intermediate shell plate and axial welds is:

1.Oe18 n/cm2

• exp(-0.24 (6.13 / 4)) = 7.0e17 n/cm 2.

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a The 23 EFPY 14T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld Is:

8.2e17 n/cm2

• exp(-0.24 (6.1314))=5.7e17 n/cm2.

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 RTNDr. 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 Nuclear Energy GE-NE-0000-0013-31 93-01 a Table 4-4: Browns Ferry Unit 2 Beltline ART Values (23 EFPY)

Lower4ntmedat.*Plate and Axil Welds Tlidoe - &l3 iNKI 30 ElPPk l. ec - 1.313118 e/cnt2 30 SPPYPek1M4Tl ouc*- 9.2E117 skae-2 23 EPbkIATIHw 75117l akm~l Lowr Plate and Aial We ind Low lo Lowwttemedat, WM Weld Thika- 6.13 li 30 EFPYPosi,.D.Ila~o- 11.llS IN vloe 30 EFPY P.k1M4 fl - 7AE117 xkv,?Z 23 EFPYPcklATflwc*- 5.7E117 u/cm'2 Imiaial 14 T 23 EFPY WI CA 23 EfPY 23 EPPY COMPONENT t IEATOR IAT/LOT ltu 1l CIF RTS& R = A RTII VAr& Shi ART IF *E/c2 P IP .P IF PLATES:

LvevrShel 6-127-14 C2467-2 0.1 OS2 112 -24 5.71 117 35 o 17 34 69 49 6-127-15 C2463-1 0.17 0.48 117 -20 5.7E 117 37 0 17 34 71 St 6-127-17 C2460-2 0.13 o51 88 0 .7E117 211 0 14 23 Ss 5s etw.mdlet Shel 6-1274 AMi1-1 0.14 056 De -10 758117 34 D 17 34 68 31 6-127.16 C2467-1 0.16 0.52 112 .10 7E1 17 39 0 l7 34 73 63 6-127-20 C2840-1 0.11 0.0 73 -10 7E1 17 26' 0 13 26 5l 41 Axiai ESW 0.24 0.37 141 23.1 716117 49 13 25 56 3o 12ti (tIdh D50733 0.09 0.66 117 .40 S.7E17 37 _ _I 37 73 33

• Ch~~drezokfawmdi o138j.

18

GE Nuclear Energy GE-NE-0000-0013-3193-01 a Table 4-5: Browns Ferry Unit 2 Beltline ART Values (30 EFPY)

Loritermedals Plate and Axial Welds Thcas 6.13 dh 30EFY PeekID.

O o- .3e18 lasr.2 30ERYPcklf4Tfllw- 9.2E17 We?,2 30 EPY Pcak lThc 92117 aukm'2 Lo. Plate ad AxialWelds nd Lower o Lertarmedate Glrlh Weld TFuic~ui~- 6.13 icl 30EFFYPta.D. 1mcc - P1.10t afae2 30EPPYPmkIATifmc- 7AE117 Wzm2 30 E YPookIATIw- 7.40117 nam'2 lehial 1 T 30 ePPY el 4a 30 EFPY 30 EPPY COMPONENT lEAT ORIEIATLOT %6 %Ni- CP RTa& pbc= A RTnAL Mlin shift ART Y _ _ _

IF 2 P IF P *P PLATKS:

LeerSIal 6-127-14 C2467-2 0 16 052 112 -20 7Jel 17 40 o 17 34 74 54 6-127-15 C2463-1 017 048 117 -20 74E117 42 17 34 76 56 6-127-17 C2460-2 0.13 0.51 8e 0 7A4E117 32 0 16 32 63 63 xtev4 m I SheHl 6-1274 AAI-1 0.14 0.5 8 -10 92E117 39 6 17 34 73 63 6-127-1e C2467-1 0.1e 052 112 .10 92E17 45 D 17 34 79 69

-127-20 C2849-1 0.11 o0o 73 -1o 92E117 29 D I 229 53 43 WELDS:

ESW 024 0.37 141 23.1 92E:17 56 13 2 62 II$ 141 cum D6733 0.0 as 117 40 7Ae 07 42 0 21 42 84 44 C*snuies oblained 1m(131.

GE Nuclear Energy GE-NE-0000-0013-3193-O1 a 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 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 XI 16]) forms the basis for the requirements of IOCFR50 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 RTNDT. 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 12] and the GE Nuclear Energy GE-NE-0000-0013-3193-01 a nozzle thermal cycle diagrams 3]. 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 150F/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 14T 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, Kj, at 14T 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-0000-001 3-3193-01 a Table 4-6: Summary of the 10CFR50 Appendix G Requirements I. Hydrostatic Pressure est & Lea I (Core is Not Critical) - Curve A

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

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 900F II. 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 RTNDT 60'F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 120°F

b. Core critical - Curve C

1. At < 20% of preservice hydrotest Larger of ASME Limits + 400F 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 + 400 F or the minimum permissible temperature for the nservice 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-beitline regions). The closure flange region limits are controlling at lower pressures primarily because of IOCFR50 Appendix G 8]

requirements. The non-beltline and beitline 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.

I GE Nuclear Energy GE-NE-0000-0013-319301 a 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltline 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 (<1.0e17 ncm2) 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 BWR/6 specifically for the purpose of fracture toughness analysis. The BWRI6 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 [61 to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWRI6 components: the feedwater nozzle (FW) 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-0000-0013-3193-01 a Table 4-7: Applicable BWR/4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Water Level Instrumentation Nozzle 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 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-beltline region were conservatively developed for a large BWR/6 (nominal inside diameter of 251 nches). The analysis is considered appropriate GE Nuclear Energy GE-NE-0000-0013-319301 a for Browns Ferry Unit 2 as the plant specific geometric values are bounded by the generic analysis for a large BWRI6, 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.

C~~~~~~~

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

In a [ I finite element analysis [ , the CRD penetration region was modeled to compute the local stresses for determination of the stress ntensity factor, K. The

[ l evaluation was modified to consider the new requirement for Mm as discussed In ASME Code Section Xl Appendix G [6] and shown below. The results of that computation were K, = 143.6 ksi-ln1n 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 84F. [

GE Nuclear Energy GE-NE-0000-0013-3193-01 a The limit for the coolant temperature change rate Is 15'Flhr or less.

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, tn = 2.83. The resulting value obtained was:

Mm = 1.85 for 4t1<2 Mm = 0.926 fi for 2< Fi 3.464 = 2.6206 Mm = 3.21 for >3.464 Kim is calculated from the equation In Paragraph G-2214.1 [6] and Klb Is calculated from the equation in Paragraph G-2214.2 [6]:

Kim = Mm pm = ksl-inl2 Klb = (2/3) Mm-ap[ ] ksi-ln' The total K, is therefore:

Ki = 1.5 (Kim+ Klb) + Mm (ask + (2/3) sb) = 143.6 ksiminm This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific K, is based on the K1, equation of Paragraph A-4200 in ASME Appendix A 17]:

GE Nuclear Energy GE-NE-0000-0013-3193-01 a (T - RTNDT) = In [(K -33.2) / 20.734] / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 84'F The generic curve was generated by scaling 143.6 ksi-in'2 by the nominal pressures and calculating the associated (T - RTNDT):

Pressure Test CRD Penetration K1 and (T - RTNDT)

Nominal Pressure as a Function Of Pressure i::::......

.~~~~~~~~~~~~~~~~~i E -Ir

7 e;

I -

-s .

i Ei:.:f; 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 42'F, as shown In Tables 4-1 and 4-2.

GE Nuclear Energy GE-NE-0000-0013-3193-01 a I

Second, the P-T curve is dependent on the calculated K value, and the K value is proportional to the stress and the crack depth as shown below:

K, a a (a) " 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) 2. The generic curve value of R/(t)', based on the generic BWR/6 bottom head dimensions, is:

Generic: R / (t) " 2 = 138 / (8) m = 49 inch (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 I (t)m = 125.7/ (8)"m = 44 inchm (4-3)

Since the generic value of R/(t) m 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 4.3.2.1.2 Core Not Critical Heatup/Cooldown - Non-Beitline 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. [

The calculated value of K, 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-in .

Therefore, the method to solve for (T - RTNDT) for a specific K is based on the KI, equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:

(T - RTNDT) = In [(K, - 33.2) / 20.734] 0.02 GE Nuclear Energy GE-NE-0000-0013-3193-01 a (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 102-F The generic curve was generated by scaling 192 ksi-inm by the nominal pressures and calculating the associated (T - RTNDT):

Core Not Critical CRD Penetration Kg and (T - RTNDT) as a Function of Pressure

.......... ..omna

. ressure r-.

.' ' ....  :,,,,: .. ,::., ^ ......... ~~~~~~~~~~~~~~~~~~~~~...............E .-

~~~~~~~~.. . .~~ .. -. :. .....-...,..-..... ....}.,E.

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 42°F, as shown in Tables 4-1 and 4-2. [

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-0000-0013-3193-01 a GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 4.3.2.1.3 Pressure Test - Non-Bettline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, K1, 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 K, = 200 ksi-in" 2 for an applied pressure of 1563 psig preservice hydrotest pressure. l l

The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1/4T through the comer thickness.

To evaluate the results, K, 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 K is shown below using the BWR/6, 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 (alrn) from Figure AS-1 of WRC-1 75 Is 1.4 where:

a= ( tn 2 t 2)1/2 =2.36Inches tn = thickness of nozzle = 7.125 nches tv = 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 r = nozzle radius (nozzle comer radius) = 3.75 inches Thus, ar = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, Is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, Is 1.5 a (a) `

• F(arn):

GE Nuclear Energy GE-NE-0000-0013-3193-01a Nominal K, = 1.5 34.97 ( 2.36) 2 1.4 = 200 ksi-ln 12 The method to solve for (T - RTHDT) for a specific K, is based on the KI, equation of Paragraph A-4200 in ASME Appendix A [17] 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 The generic pressure test P-T curve was generated by scaling 200 ksi-inm by the nominal pressures and calculating the associated T - RTNDT), L I

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1

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-0000-001 3-3193-01 a I

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 [1 9] and K,are shown below:

Vessel Radius, R, 125.7 inches Vessel Thickness, t, 6.125 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR I t = 1563 psig 125.7 inches / (6.125 inches) = 32,077 psi.

The dead weight and thermal RFE stress of 2.967 ksi Is conservatively added yielding a = 35.04 ksi. The factor F (a/rn) from Figure A5-1 of WRC-175 is determined where:

a= 1/4 ( t 2 + t 2)1/2 =2.32 Inches tn = thickness of nozzle = 6.96 inches tv = thickness of vessel = 6.125 inches rn= apparent radius of nozzle = r + 0.29 r=6.9 inches r = actual inner radius of nozzle = 6.0 inches rc= nozzle radius (nozzle comer radius) = 3.0 inches Thus, a/r, = 2.32 6.96 = 0.33. The value F(a/rj), taken from Figure AS-1 of WRC Bulletin 175 for an a/rn of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K 1, Is 1.5 a (a) m F(alr):

Nominal K, = 1.5 35.04 * ( 2.32)112- 1.4 = 198.7 ksi-in"2 GE Nuclear Energy GE-NE-0000-0013-3193-01 a

[

4.3.2.1.4 Core Not Critical Heatup/Cooldown - Non-Belftine Curve B (Using FeedwaterNozzle/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 [ j finite element analysis done specifically for the purpose of fracture toughness analysis [ . Analyses were performed for all feedwater nozzle transients that Involved rapid temperature changes. The most severe of these was normal operation with cold 400 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 (Kip) is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole Ina flat plate:

Kip = SF a (a)%* F(a/rJ) (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 Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/rn) for Equation 44. These values are shown In Figure A5-1 of WRC Bulletin 175 [15].

The stresses used In Equation 4-4 were taken from l ] design stress reports for the feedwater nozzle. The stresses considered are primary membrane, cpm, and primary bending, apb. Secondary membrane, arm, and secondary bending, acb, stresses are included in the total K, by using ASME Appendix G [6J methods for secondary portion, Kj,:

K1 = Mm (asm + (2/3) ab) (4-5)

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 [153.

However, the correction was not applied to primary membrane stresses because primary stresses satisfy the laws of equilibrium and are not self-limiting. Kp and KI, 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 K was calculated, the following relationship was used to determine (T - RTNDT).

The method to solve for (T - RTNDT) for a specific K is based on the K equation of Paragraph A-4200 in ASME Appendix A [171. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

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

Example Core Not Critical Heatup/Cooldown Calculation for Feedwater Nozzle/Upper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the [ l feedwater nozzle [ ] analysis, where feedwater Injection of 400 F into the vessel while at operating conditions (551 .4F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle corner stresses were obtained from finite element analysis ]. 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 (p) was adjusted for the actual ]vessel thickness of 6.1875 Inches (i.e., cap, = 20.49 ksi was revised to 20.49 ksi 7.5 nches/6.1875 inches = 24.84 ksi). These stresses, and other Inputs used in the generic calculations, are shown below:

apm = 24.84 ksi a.,m = 16.19 ksi cry. = 45.0 ksi t, = 6.1875 Inches apb = 0.22 ksi = 19.04

=ab ksi a = 2.36 Inches r, = 7.09 inches tn = 7.125 Inches GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 = [ads - apm + ((atoal - ays) / 30)] / (agoal - 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 apm. The resulting stresses are:

apm = 24.84 ksi csm = 9.44 ksi apb = 0.13 ksi ab = 11.10 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on the 4a thickness; hence, t2 = 3.072. The resulting value obtained was:

Mm = 1.85 for _t 2 Mm = 0.926 fi for 2< Fi 53.464 = 2.845 Mm = 3.21 for fi >3.464 The value F(arn), taken from Figure A5-1 of WRC Bulletin 175 for an ars of 0.33, is therefore, F(aIrn) =1.4 Kip is calculated from Equation 4-4:

Kip = 2.0 (24.84 + 0.13) (it . 2.36)"2 1.4 Kip = 190.4 ksi-in"2 Ki, is calculated from Equation 4-5:

GE Nuclear Energy GE-NE-0000-0013-3193-01 a K1, = 2.845 (9.44 + 213 11.10)

K 8 = 47.9 ksi-in1 The total K is, therefore, 238.3 ksi-in1'.

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

(T - RTNDT) = In [(238.3-33.2) 20.734 / 0.02 (T - RTNDT) = 115eF The [ l curve was generated by scaling the stresses used to determine the Ki; 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 400F water Injected into the hot reactor vessel nozzle. In the base case that yielded a K, value of 238 ksi-in"7, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4°F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tsa.u.i. - 40) / (551.4 - 40). From K the associated (T - RTNDT) can be calculated:

Core Not Critical Feedwater Nozzle K and (T - RTNDT) as a Function of Pressure anomInalPressure Saturation Temp. R .F t- TNT

....1PSig) E (F)

_______E : - . -...f. -(ksI-ln'~

E-

...... .. . i -('F):.

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 K,.

GE Nuclear Energy GE-NE-0000-0013-3193-01 a The highest non-beltine RTNDT for the feedwater nozzle at Browns Ferry Unit 2 Is 40°F as shown in Table 4-2. However, the RTNDT was Increased to 44F 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.

4.3.2.2 CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beltline region are determined according to the ASME Code 161. As the beitline fluence increases with the Increase in operating life, the P-T curves shift to a higher temperature.

The stress intensity factors (K,), 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 1000F/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.

4.3.2.21 Beltline Region - Pressure Test The methods of ASME Code Section Xi, Appendix G [6 are used to calculate the pressure test beitline limits. The vessel shell, with an inside radius (R) to minimum GE Nuclear Energy GE-NE-0000-0013-3193-01a thickness (tmin) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

Am = PR / tmin (4-8)

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

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6J for comparison with Kjc, 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 Kic equation of Paragraph A-4200 in ASME Appendix A [17]

for the pressure test condition:

Kim SF = Kc = 20.734 exp[0.02 (T - RTNDT )] + 33.2 (4-9)

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

GE's current practice for the pressure test curve is to add a stress Intensity factor, Kgt, for a coolant heatup/cooldown 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 100F/hr. The Kit calculation for a coolant heatup/cooldown rate of 100F/hr Is described in Section 4.3.2.2.3 below.

4.3.2.2.2 Calculations for the Beitline 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-0000-0013-3193-01 a 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 (4-10)

= 1064 + (875.13-216.3) 0.0361 = 1088 psig Pressure stress:

cr = PR/t (4-11)

= 1.088 125.7/6.13 = 22.3 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [61 was based on a thickness of 6.13 inches (the minimum thickness without cladding);

hence, tn = 2.48. The resulting value obtained was:

Mm = 1.85 for %ft2 M, = 0.926 Jt for 2< i <3.464 = 2.29 Mm = 3.21 for ft >3.464 The stress intensity factor for the pressure stress Is Kim = Mm a. The stress Intensity factor for the thermal stress, Kt, is calculated as described in Section 4.3.2.2.4 except that the value of G" is 15°F/hr instead of 100'F/hr.

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

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

GE Nuclear Energy GE-NE-0000-0013-3193-01 a (T - RTNDT) = ln[(1.5 Kim + Kit- 33.2) 20.734] / 0.02 (412)

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

= 38.90 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]:

KR = 2.0* Kim +Kh (4.13) where Kim is primary membrane K due to pressure and Kft is radial thermal gradient K due to heatup/cooldown.

The pressure stress intensity factor Kim 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 M,from Figure G-2214-1 of ASME Appendix G [6] by the through-wall temperature gradient AT, 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-000-0013-3193-01 a a 2T(x,t) / ax 2 = I / p (ff(x,t) I t) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that T(x,t) at = dT(t) / 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 / 23 - GCx I p + To (4-15)

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

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

The M 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-OO0O-0013-319301 a 4.3.2.2.4 Calculations for the Beitline Region Core Not Critical Heatup/Cooldown 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 Kim 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 KR term for the thermal stress. The additional inputs used to calculate Kt are:

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

Thermal diffusivity at 5500 F (most conservative value) P = 0.354 ff1 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 /2j2 (4-16)

= 100 (0.526)2/ (2 0.354) = 39"F The analyzed case for thermal stress is a 14T 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, K,, = M,

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

= ln[(2*51.1 + 11.39-33.2)/20.734]/0.02

= 67.8 F GE Nuclear Energy GE-NE-0000-0013-3193-01 a T can be calculated by adding the adjusted RTNDT:

T = 67.8 141 = 208.8 F 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 Inthe ASME Code, based on the closure flange region RTNDT. Similar to the evaluations performed for the bottom head and upper vessel, a BWR/6 finite element analysis [18] was used to model the flange region. The local stresses were computed for determination of the stress intensity factor, K1. Using a 1/4T flaw size and the Kic 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 + 900 F (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 54°F); 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 60°F adder is Included by GE for two reasons: 1)the pre-1971 requirements of the ASME Code Section III, Subsection NA, Appendix G included the 600 F adder, and 2) inclusion of the additional 600F requirement above the RTNDT provides the additional assurance that a 14T 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-00000013-3193-01 a 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 + 90OF) and Curve B temperature no less than (RTNDT + 1200F).

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 68OF for the reason discussed below.

The shutdown margin, provided in the Browns Ferry Unit 2 Technical Specification, is calculated for a water temperature of 68OF. 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 68°F limit, further extensive calculations would be required to justify a lower temperature. The 83°F limit for the upper vessel and beltline region and the 68OF 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 Git-N E-0000-001 3-3193-01 a 4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFR50 Appendix G [8J, Table 1. Table I of [8] requires that core critical P-T limits be 40'F 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 I of I OCFR50 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 23.1F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160°F or the temperature required for the hydrostatic pressure test (Curve A at 1064 psig). The requirement of closure region RTNDT + 160'F causes a temperature shift in Curve C at 312 psig.

.l

GE Nuclear Energy GE-NE-0000-0013-3193-01a

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 1000F/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]. 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 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, KIr, at 1/4T to be less than that at 314T for a given metal temperature.

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 Limits (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.

While 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 flux from Reference 14, 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-NE-0000-0013-31 93-Ola Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves 23 EFPY Curves A Bottom Head Limits (CRD Nozzle) -23 EFPY Figure 5-1 Table B-1 A l Upper Vessel Limits (FW Nozzle) - 23 EFPY Figure 5-2 Table B-1 A Beltline Limits - 23 EFPY Fiqure 5-3 Table B-1 A Bottom Head and Composite Curve A - 23 EFPY Figure 5-4 Table B-2 B Bottom Head Limits (CRD Nozzle) - 23 EFPY Figure 5-5 Table B-1 B Upper Vessel Limits (FW Nozzle) - 23 EFPY Figure 5-6 Table B-1 B Beftline Limits - 23 EFPY Figure 5-7 Table B-1 B Bottom Head and Composite Curve B - 23 EFPY* Figure 5-8 Table B-2 C Composite 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 B-4 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 B-3 B Beltline Limits - 30 EFPY Fiqure 5-17 Table B-3 B Bottom Head and Composite Curve B - 30 EFPY* Figure 5-18 Table B4 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 & B curve Is the more limiting of three limits: 10CFR5O Bolt-up Limits, Upper Vessel Limits (FW Nozzle), and Beitline Limits. 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: 10CFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Limits (FW Nozzle), and Beltline Limits.

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 _- _ _

1300 49°F FOR BOTTOM H 1200 HEATUPICOOLDOY RATE OF COOLAN I 5'F/HR 1100 - _____

- 1000 IL 900 W 800 a) j _ _

O 700

• 90 PSI - _ - ACCEPTABLE ARE OF OPERATION TC THE RIGHT OF THI L600 CURVE l400 E

300 E

ii

_ LII 200- -BOTTOM HEAD L 100 __ __ __ __

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

[1 5 0F/hr or less coolant heatuplcooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 400

1. INITIAL RTndt VALUE IS 44F FOR UPPER VESSELl 1.1200 _ HEATUPICOOLDOWN 200 RATE OF COOLANT

< 15F/IHR 1.1000 900 -

I 800 810PSIG - _ __

ACCEPTABLE AREA OF 700 _ _ OPERATION TO THE E02 w RIGHT OF THIS CURVE A.

600- - _ _

500 - _ _ __ _

uJ 02 400 - _ _ _ _

300 - ___ _ __ _

FLNGE REGION 83F -UPPER VESSEL

_____ J____ _____

_____ _____ ~ LIM ITS (Including 100 - - Flange and FW

_ _ __ _ ~~~~~~~~Nozzle Limits) 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-2: Upper Vessel P-T Curve for Pressure Test [Curve AJ - 23 EFPY 15°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 INITIAL RTndt VALUE IS 1300 23.1F FOR BELTUNE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

1100 EFPY SHIFT ( F) 23 105

-a 0 1000 HEATUPICOOLDOWN IL 900 RATE OF COOLANT 0

< 15 FHR i

'U Co 800 o 700 ACCEPTABLE AREA OF E 600 OPERATION TO THE RIGHT OF THIS CURVE t 500

'l i 400 UJ a3 300 200 100 0

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

Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 23 EFPY

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

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a 1400 1300 I INITIAL RTndt VALUES ARE 23.1F FOR BELTUNE, 44-F FOR UPPER VESSEL, 1200 I AND 49 F FOR BOTTOM HEAD BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (F) 23 105 a 1000 a

HEATUP/COOLDOWN IL 900 RATE OF COOLANT 0

I.- < 15*FIHR

-j

,, 800 co o 700 t 600 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 500 IZ U 400 300

- UPPER VESSEL 200 AND BELTLINE UMITS BOTTOM HEAD .

100 CURVE 0

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

Figure 5-4: Composite Pressure Test P-T Curves [Curve A] up to 23 EFPY

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

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 1300 1200 1100

-1000 f 900 0

UJ to 800 rn o 700 it 600 n 500 t 400 IL 300 200 100 0

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

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

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

.A

GE Nuclear Energy GE-NE-0000-0013-3193-01a 1400 INITIAL RTndt VALUE IS 1300 44 F FOR UPPER VESSELl 1200 HEATUP1COOLDOWN RATE OF COOLANT I 100rF/HR 1100 C.

Caoo

~;1000 0

a- 900 0

U

0) 800 (0 ACCEPTABLE AREA OF o 700 OPERATION TO THE RIGHT OF THIS CURVE 3 600 3

3 500

'u 0 400 a.

300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Umits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

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

[10001Fhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 1400 INITIAL RTndt VALUE IS 1300 23.1VF FOR BELTUNE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (¶F) 1100 23 105

.9 i 1000 HEATUP/COOLDOWN z RATE OF COOLANT

< 100F/HR R 900 o4 800 o 700 t ACCEPTABLE AREA OF OPERATION TO THE E 600 R[GHT OF THIS CURVE R

x 500 K

00 i

300 200 100 0

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

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

[100 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a 1400 INITIAL RTndt VALUES ARE 23.1F FOR BELTLINE, 1300 44-F FOR UPPER VESSEL, AND 49'F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (F) 23 105 a 1000 a

HEATUP/COOLDOWN

0. 900 RATE OF COOLANT 0 c 100FFIHR UI In 800 o 700 ACCEPTABLE AREA OF D 600 OPERATION TO THE z RIGHT OF THIS CURVE 3 500 VW 400 U

300

- UPPER VESSEL 200 AND BELTUNE LIMITS 100 BOTTOM HEAD .

CURVE 0

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

Figure 5-8: Composite Core Not Critical P-T Curves [Curve B1up to 23 EFPY

[1 000F/hr or less coolant heatup/cooldownj GE Nuclear Energy GE-NE-O000-O013-3193-01 a 1400 INITIAL RTndt VALUES ARE 1300 23.1-F FOR BELTLINE, 44*F FOR UPPER VESSEL, 1200 AND 49 F FOR BOTTOM HEAD 1100 vi>

co

-, 1000 BELTUNE CURVE ADJUSTED AS SHOWN:

a EFPY SHIFT (F)

IL 900 23 105 0

R.

u o 800 HEATUPICOOLDOWN RATE OF COOLANT

< 100°F/HR f700 600 ACCEPTABLE AREA OF E 500 OPERATION TO THE UJ RIGHT OF THIS CURVE 0:

at 400 300 200

-BLTNNE AND NON-BELTLINE 100 LIMITS 0

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

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

[100*F/hr or less coolant heatupfcooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a B C 1400 INITIAL RTndt VALUES ARE 23.1-F FOR BELTLINE, 1300 44 F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (F) 23 105 S 1000 a

HEATUP/COOLDOWN IL 900 RATE OF COOLANT 0 c 100F/HR 0

0z 800 51 SIG I 513 1 1 1 111- -

o 700 ACCEPTABLE AREA OF 8 600 OPERATION TO THE RIGHT OF THIS CURVE I~ I ~ . IEA 3 500

) 400 300 200 -COMPOSITE CURVE B

-v-I- -- - __ -EG_ HEAD .

100 0

[ ]' 12 -_I_83 -

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

(.F)

Figure 5-10: Composite Core Not Critical [Curve B] including Bottom Head and 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-OO13-3193-01 a 1400 - _

i ~~~~~~INITIAL RTnidt VALUE 1300 /49F FOR BOTTOM HE 1200 - _ _ HEATUPICOOLDOW N RATE OF COOLANI

< 156F/HR 1100 0.

-. 1000 900 u

0 800 co o 700 690 PSI - ACCEPTABLE AREi k OF OPERATION TO THE RIGHT OF THIS 600 CURVE__

10:

3 00 C300

• BlD 200 - BOTTOM HEAD U 100 0 - _ _

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

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

[1150 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-31 93-Ola 1400 INITIAL RTndt VALUE IS 44F FOR UPPER VESSEL 1300 HEATUP/COOLDOWN 1200 RATE OF COOLANT

< 1-F/HR 1100 D 1000 0

0. 900 0

us

'U U) 800 U) ACCEPTABLE AREA OF OPERATION TO THE o 700 RIGHT OF THIS CURVE

'U

500 L

w 4003I00 30 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle jms) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE OF)

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

[1 5 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 INITIAL RTndtVALUE IS 1300 23.1*F FOR BELTUNE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

1100 EFPY SHIFT F) 30 118 a-a1000 HEATUP/COOLDOWN a- 900 RATE OF COOLANT 0 < 15F/HR I-8 800 o 700 ACCEPTABLE AREA OF OPERATION TO THE X 600 RIGHT OF THIS CURVE t

i 500 Uj CC 400 300 200 100 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 A] up to 30 EFPY (15°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-319301 a 1400 INITIAL RTndt VALUES ARE 1300 I 23.1 F FOR BELTUNE, 44-F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD 1200 BELTLINE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (F) 30 118

. 1000 a HEATUPICOOLDOWN RATE OF COOLANT A. 900 c 15*FIHR 0

'U

,, 800 o 700 ACCEPTABLE AREA OF A 600 OPERATION TO THE RIGHT OF THIS CURVE 3Z00 400 3

- UPPER VESSEL 200 AND BELTUNE LIMITS HEAD 100 CURVE 0

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

Figure 5-14: Composite Pressure Test P-T Curves [Curve A] up to 30 EFPY 1[50Fhr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 lINITIAL RTnidt VALUE ISl 1300 149'F FOR BOTTOM HEAD HEATUP/COOLDOWN 1200 I RATE OF COOLANT l]< 100-F/HR 1100 L 1000 IL 900 0

I-

-i NJ

() 800 ACCEPTABLE AREA OF OPERATION TO THE o 700 RIGHT OF THIS CURVE a 600 t

i 500

) 400 300 200 100 0

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

Figure 5-15: Bottom Head P-T Curve for Core Not Critical [Curve B] - 30 EFPY

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

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 lINITIAL RTndl: VALUE ISl 1300 144 F FOR UPPER VESSEL 1200 lHEATUPCOOLDOWN RATE OF COOLANT

< 1000F/HR 1100 C.

0.

1000 c

30 i-900 it 02 800 ACCEPTABLE AREA OF In OPERATION TO THE 700 RIGHT OF THIS CURVE Is 600 500 m

m ox 400 U.

cL 300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 0 I I I I1 I I I 1 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

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

[1 000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 1400 1300 INITIAL RTndt VALUE IS 23.1VF FOR BELTUNE 1200 BELTLINE CURVE 1100 ADJUSTED AS SHOWN:

ea EFPY SHIFT ('F) 30 118 a 1000 HEATUP/COOLDOWN a-m 900 RATE OF COOLANT

< 100'FHR C° 800 o 700 0 600 ACCEPTABLE AREA OF OPERATION TO THE 9

E 500 RIGHT OF THIS CURVE 8 400 300 200 100 0

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

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

[1 00 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 INITIAL RTndt VALUES ARE 23.1-F FOR BELTLINE, 1300 44'F FOR UPPER VESSEL, AND 49'F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT -F)

D 30 118 CI

-1000 a

HEATUP/COOLDOWN IL 900 RATE OF COOLANT

< 100°F/HR w

°0 800 o 700 0 600 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE

> 500 w

an 400 w

300

- UPPER VESSEL 200 AND BELTLINE LIMITS 100 ---- HEAD -

CURVE 0

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

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

[I000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 1400 INITIAL RTndt VALUES ARE 23.1F FOR BELTUNE, 1300 44F FOR UPPER VESSEL, AND 1200 49'F FOR BOTTOM HEAD 1100 BELTLINE CURVE ADJUSTED AS SHOWN:

to 1000 EFPY SHIFT (F)

C 30 118 cL 900 0

HEATUPICOOLDOWN RATE OF COOLANT en 800 us < 100 FIHR o 700 U

- 600 E

ACCEPTABLE AREA OF 3 500 OPERATION TO THE RIGHT OF THIS CURVE 2

Si400 wj a-300 200 Temperature 83T] , e BELTLINE AND I I NON-BELTLINE 100 _ - - _ -- _ _ _ _ _ _LIMITS 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

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

GE Nuclear Energy GE-NE-0000-0013-319301 a B C 1400 INITIAL RTndt VALUES ARE 23.1F FpR BELTUNE, 1300 44 F FOR UPPER VESSEL, 1200 - .' - 1 -I - -

AND 49°F FOR BOTTOM HEAD BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

.2 30 118

1000 a

HEATUPICOOLDOWN IL 900 RATE OF COOLANT 0 < 100°F/HR I

In 800 0+/-

O 700 z 600 ACCEPTABLE AREA OF]

OPERATION TO THE l RIGHT OF THIS CURVE]

i 500 w

. PSIG*--

- -S-V 400

- COMPOSITE CURVE B 300 HEAD .

CURVE B 200

-COMPOSITE 68 F i / FLANGE CURVE C REGION 100 83F 0

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

Figure 5-20: Composite Core Not Critical [Curve B] Including Bottom Head and Core Critical P-T Curves [Curve C up to 30 EFPY [IO00F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-01a

6.0 REFERENCES

1. RG Carey (GE) to HL Williams (VA), 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 13589990, 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, Califomia, September 1994 (GE Proprietary).

GE Nuclear Energy GE-NE-0000-0013-3193-01 a

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. Vemon, 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 1 - Pressure-Temperature (P-T) Curve Update", Docket Nos. 50-260 and 50-296, (TVA-BFN-TS-393, Supplement 1, 10 CFR 50.90 (R08 981215 742)),

December 15, 1998.

14. a) S. Wang, Project Task Report, Tennessee Valley Authority Browns Ferry Unit 2 and Unit 3 Extended Power Uprate, Task T0313: RPV Flux Evaluation", GE-NE, San Jose, CA, March 2002 (GE-NE-A22-00125-19-01, Revision 0)(GE Proprietary Information).

b) Letter, S.A. Richard, USNRC to J.F. Klapproth, 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 XI of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

GE Nuclear Energy GE-NE-0000-0013-3193-Ola

18. [

19. Bottom Head and Feedwater Nozzle Dimensions:

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

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

20.

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-B1 100639-01, Revision 1).

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-NE-000O-001 3-3193-01 a

[

A-2

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a Table A Geometric Discontnuities 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 RTNDT plus 600F. Also Inconel discontinuities require no fracture toughness evaluations.

Nozzle or Appurtenance Material Reference Remarks MKI2 - 2" Instrumentation (attached to Alloy 600 1,6,9,10,23 Nozzles made from Alloy 600 and less Shells 58,59,60) Shell 58 MKI2 nozzle than 2.5' require no fracture toughness Iswithin 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.

MKI01 - 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.

MK132 - 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 evaluaffon.

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 Specimen SA276 304 1,21,22 Appurtenances made from Stainless Brackets (Shells 58 and 59) Steel require no fracture toughness 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. I

• The high/low pressure leak detector, and the seal leak detector are the same nozzle; these nozzles are the dosure flange leak detection nozzles.

A-3

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 1, Vessel Sub-Assembly", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-059).

• Drawing #122855E, Revision 14, List 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 Forging", 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, Lower Head Assembly", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-013).

A-4

GE Nuclear Energy GE-NE-0000-0013-3193-01 a

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 & Wilcox Company, Mt. 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. Vemon, 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-0000-0013-31 93-01 a APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-NE-0000-0013-31 93-01a 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 gROM.UPE R a E;P 80T> UPE 23 EF 0 68.0 83.0 83.0 68.0 83.0 83.0 10 68.0 83.0 83.0 68.0 83.0 83.0 20 68.0 83.0 83.0 68.0 83.0 83.0 30 68.0 83.0 83.0 68.0 83.0 83.0 40 68.0 83.0 83.0 68.0 83.0 83.0 50 68.0 83.0 83.0 68.0 83.0 83.0 60 68.0 83.0 83.0 68.0 83.0 83.0 70 68.0 83.0 83.0 68.0 83.0 83.0 80 68.0 83.0 83.0 68.0 83.0 83.0 90 68.0 83.0 83.0 68.0 83.0 83.0 100 68.0 83.0 83.0 68.0 83.0 83.0 110 68.0 83.0 83.0 68.0 83.0 83.0 120 68.0 83.0 83.0 68.0 83.0 83.0 130 68.0 83.0 83.0 68.0 83.0 83.0 140 68.0 83.0 83.0 68.0 83.0 83.0 150 68.0 83.0 83.0 68.0 84.2 83.0 160 68.0 83.0 83.0 68.0 86.9 83.0 170 68.0 83.0 83.0 68.0 89.5 83.0 180 68.0 83.0 83.0 68.0 91.9 83.0 190 68.0 83.0 83.0 68.0 94.2 83.0 200 68.0 83.0 83.0 68.0 96.3 83.0 210 68.0 83.0 83.0 68.0 98.3 83.0 220 68.0 83.0 83.0 68.0 100.3 83.0 230 68.0 83.0 83.0 68.0 102.1 83.0 B-2

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 & 5-10 240 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 90.4 312.5 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 68.0 143.0 143.0 450 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-0000-0013-31 93-01a 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 & 5-10 490 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 73.9 143.0 149.8 560 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 86.5 143.4 160.3 640 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 128A 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-0000-0013-31 93-01 a 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 & 5-10 740 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 86.5 114.8 145.4 109.6 151.6 180.6 850 87.6 115.5 146.6 110.4 151.9 181.3 860 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 93.4 119.6 153.1 115.4 154.0 185.8 920 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-NE-0000-0013-3193-01 a 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 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 106.9 129.5 168.1 127.2 159.5 196.7 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-0000-0013-31 93-01 a 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 1250 117.5 137.8 179.6 136.7 164.4 205.6 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 143A 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-0000-0013-31 93-01 a 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 U

10 68.0 83.0 68.0 83.0 83.0 20 68.0 83.0 68.0 83.0 83.0 30 68.0 83.0 68.0 83.0 83.0 40 68.0 83.0 68.0 83.0 83.0 50 68.0 83.0 68.0 83.0 83.0 60 68.0 83.0 68.0 83.0 84.0 70 68.0 83.0 68.0 83.0 91.2 80 68.0 83.0 68.0 83.0 97.2 90 68.0 83.0 68.0 83.0 102.3 100 68.0 83.0 68.0 83.0 106.8 110 68.0 83.0 68.0 83.0 110.9 120 68.0 83.0 68.0 83.0 114.7 130 68.0 83.0 68.0 83.0 118.2 140 68.0 83.0 68.0 83.0 121.4 150 68.0 83.0 68.0 84.2 124.2 160 68.0 83.0 68.0 86.9 126.9 170 68.0 83.0 68.0 89.5 129.5 180 68.0 83.0 68.0 91.9 131.9 190 68.0 83.0 68.0 94.2 134.2 200 68.0 83.0 68.0 96.3 136.3 210 68.0 83.0 68.0 98.3 138.3 220 68.0 83.0 68.0 100.3 140.3 B-8

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a 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 & 5-10 240 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-0000-0013-3193-01a 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 & 5-10 460 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-I 0

GE Nuclear Energy GE-NE-0000-0013-3193-1Oa 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 & 5-10 710 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 81.0 139.0 105.0 176.4 216.4 800 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 89.6 148.9 112.1 182.9 222.9 880 90.5 150.0 113.0 183.6 223.6 890 91.5 151.0 113.8 184.4 224.4 900 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-1I

GENuclear Energy GE-NE-0000-0013-3193-01a 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 1 O.3 970 98.6 159.0 119.9 189.9 229.9 980 99.4 159.9 120.6 190.6 230.6 990 100.2 160.8 121.3 191.2 231.2 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 104.7 165.8 125.3 194.9 234.9 1060 105.4 166.6 125.9 195.5 235.5 1064 105.7 166.9 126.2 195.7 235.7 1070 106.2 167.4 126.5 196.1 236.1 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

GENuclear Energy GE-NE-0000-0013-3193-01 a 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 & 5-10 114.U 103.0 ZU.1 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 124.3 186.9 142.9 211.4 251.4 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-0000-0013-31 93-01 a 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 NE~~t> VE$$EL 8etitL1.HEf V$Et ELL S~tJRECURVEACUWE A '<WWEA CURVAD CUVE E ~UV I 0 68.0 83.0 83.0 68.0 83.0 83.0 10 68.0 83.0 83.0 68.0 83.0 83.0 20 68.0 83.0 83.0 68.0 83.0 83.0 30 68.0 83.0 83.0 68.0 83.0 83.0 40 68.0 83.0 83.0 68.0 83.0 83.0 50 68.0 83.0 83.0 68.0 83.0 83.0 60 68.0 83.0 83.0 68.0 83.0 83.0 70 68.0 83.0 83.0 68.0 83.0 83.0 80 68.0 83.0 83.0 68.0 83.0 83.0 90 68.0 83.0 83.0 68.0 83.0 83.0 100 68.0 83.0 83.0 68.0 83.0 83.0 110 68.0 83.0 83.0 68.0 83.0 83.0 120 68.0 83.0 83.0 68.0 83.0 83.0 130 68.0 83.0 83.0 68.0 83.0 83.0 140 68.0 83.0 83.0 68.0 83.0 83.0 150 68.0 83.0 83.0 68.0 84.2 83.0 160 68.0 83.0 83.0 68.0 86.9 83.0 170 68.0 83.0 83.0 68.0 89.5 83.0 180 68.0 83.0 83.0 68.0 91.9 83.0 190 68.0 83.0 83.0 68.0 94.2 83.0 200 68.0 83.0 83.0 68.0 96.3 83.0 210 68.0 83.0 83.0 68.0 98.3 83.0 220 68.0 83.0 83.0 68.0 100.3 83.0 230 68.0 83.0 83.0 68.0 102.1 83.0 B-14

GE Nuclear Energy GE-NE-000OO013-3193-01 a 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 ffi 'fg  ::Z  : "'I"" p1.R:.a..,Ffi ]0 T0 B.

(P$G).. ... i. F.( l 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 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-1 5

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 & 5-20 480 68.0 113.0 113.0 68.0 143.0 1514 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-0000-0013-3193-01 a 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 P. L 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 118A 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-0000-0013-31 93-01 a 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/lr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 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 128A 160.1 210.8 1105 108.6 130.8 183.0 128.7 160.3 211.1 1110 108.9 131.1 183.4 129.0 160.4 211.4 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-1 8

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 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 HEAD; :ROiLTLINE Wl- . BE.

Elilll llU lllA CURVE A (lJRE.. ..  ; a r k 1240 116.9 1374 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-0000-0013-3193-01 a 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 U

10 68.0 83.0 68.0 83.0 83.0 20 68.0 83.0 68.0 83.0 83.0 30 68.0 83.0 68.0 83.0 83.0 40 68.0 83.0 68.0 83.0 83.0 50 68.0 83.0 68.0 83.0 83.0 60 68.0 83.0 68.0 83.0 84.0 70 68.0 83.0 68.0 83.0 91.2 80 68.0 83.0 68.0 83.0 97.2 90 68.0 83.0 68.0 83.0 102.3 100 68.0 83.0 68.0 83.0 106.8 110 68.0 83.0 68.0 83.0 110.9 120 68.0 83.0 68.0 83.0 114.7 130 68.0 83.0 68.0 83.0 118.2 140 68.0 83.0 68.0 83.0 121.4 150 68.0 83.0 68.0 84.2 124.2 160 68.0 83.0 68.0 86.9 126.9 170 68.0 83.0 68.0 89.5 129.5 180 68.0 83.0 68.0 91.9 131.9 190 68.0 83.0 68.0 94.2 134.2 200 68.0 83.0 68.0 96.3 136.3 210 68.0 83.0 68.0 98.3 138.3 220 68.0 83.0 68,0 100.3 140.3 B-20

GE Nuclear Energy GE-NE-0000-0013-3193-Ola 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 & 5-20 240 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-0000-0013-3193-01 a 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 & 5-20 460 68.0 113.0 68.0 147.5 187.5 470 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 80.6 168.4 208.4 600 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-0000-0013-31 93-01 a 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 W~~~~~~UQM UPP ~ ~ ~ ~ ~ P -- .....

HEAD.

~JM~IN~T 1~~~ *EAE ~TNiA ICNSA 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 91.5 164.0 113.8 197.4 237.4 900 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 95.1 168.2 116.9 200.2 240.2 940 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-0000-0013-3193-01 a 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

~~rI~ssURE

~ c'JR\.*.A.~~..... t~JIWE ~ .. ..... ..

• .......*........... T 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 183A 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-0000-001 3-3193-01 a TABLE B4. Browns Ferry Unit 2 Composite 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 5-14, 5-18, 5-19 & 5-20

~~~U~t~

~ cURVE A:. I A M;BE M 1190 114.0 188.9 133.5 215.6 255.6 1200 114.6 189.5 134.1 216.1 256.1 1210 115.2 190.2 134.6 216.6 256.6 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 121.3 196.7 140.2 221.8 261.8 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 123.8 199.4 142.5 224.0 264.0 1380 124.3 199.9 142.9 224.4 264A 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-0000-0013-31 93-Ola APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 [11.

C-2

GE Nuclear Energy GE-NE-0000-0013-31 93-01 a 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 0 F 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 Heatup/Cooldown 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 LIMITS 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-0000-0013-3193O1 a 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 1000F in one hour period heatup)

• Shutdown (coolant temperature change of less than or equal to 1001F In one hour period cooldown)

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

Cut

GE Nuclear Energy GE-NE-0000-0013-3193-1a 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-0000-0013-31 93-01 a APPENDIX D GE SIL 430 D-1

GE Nuclear Energy GE-NE-O00-O001 3-3193-01 a 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 isto 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 Use Limitations Steam dome saturation Primary measurement Must convert saturated temperature as determined above 2120 F for Tech steam pressure to from main steam instrument Spec 100°F/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 2121F for Tech Must comply with SIL 251 Spec I000F/hr heatup to avoid vessel stratification.

and cooldown rate.

Alternate measurement When above 2120 F need to above 212F. allow for temperature variations (up to 10-15 0F lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GE Nuclear Energy GE-NE-0000-0013-319301 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations 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).

RHR heat exchanger Alternate measurement Must have previously inlet coolant for Tech Spec 00°F/hr correlated RHR inlet temperature cooldown rate when in coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow. Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta T's will be Indicated coolant temperature. Delta T limit is 100F for BWR/6s and 145 0F for earlier BWRs.

Primary measurement to Must have drain line comply with Tech Spec flow. Use to verify brittle fracture compliance with Tech limits during cooldown. Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).

Alternate Information Must compensate for outside only measurement for metal temperature lag bottom head inside/ during heatup/cooldown.

outside metal surface Should have drain line flow.

temperatures.

D-3

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Closure head flanges Primary measurement for Use for metal (not coolant) outside surface T/Cs BWRI6s to comply with temperature. Install Tech Spec brittle fracture temporary T/Cs for metal temperature limit alternate measurement, If for head bolt-up. required.

One of two primary measure-ments for BWR/6s for hydro test.

RPV flange-to-shell Primary measurement for Use for metal (not coolant) junction outside BWRs earlier than 6s to temperature. Response surface T/Cs comply with Tech Spec faster than closure head brittle fracture metal flange T/Cs.

temperature limit for head bolt-up.

One of two primary Use RPV closure head flange measurements for BWRs outside surface as alternate earlier than 6s for measurement.

hydro test. Preferred in lieu of closure head flange T/Cs if available.

RPV shell outside Information only. Slow to respond to RPV surface T/Cs coolant changes. Not available on BWRI6s.

Top head outside Information only. Very slow to respond to RPV surface T/Cs coolant changes. Not avail-able on BWR/6s.

D-4

GE Nuclear Energy GE-NE-0000013-3193-01a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Bottom head outside I of 2 primary measurements Should verify that vessel surface T/Cs to comply with stratification is not Tech Spec brittle fracture present for vessel hydro.

metal temperature (see SIL No. 251).

limit for hydro test.

Primary measurement to Use during heatup to verify comply with Tech Spec compliance with Tech Spec brittle fracture metal metal temperature/reactor temperature limits pressure curves.

during heatup.

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.

D-5

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 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:

SiLs 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 SiLs 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 SLs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SLs.

D-6

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy GE-NE-000-0013-3193-01a 10CFR5O, Appendix G defines the beltline 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 IOCFR50 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 "Om:

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

Girth Weld between Shell Ring #2 and Shell Ring #3 385.8" 11.5]

From [2], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline 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 [2,3], 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-0000-0013-3193-01 a 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 1.0e7 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 [61 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

-11" above TAF. The beltline region considered in the development of the P-T curves is adjusted to include the additional 11" above the active fuel region and the additional 8" below the active fuel region. This adjusted beltline region extends from 208.3 to 377.3" above reactor vessel M"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.

E-3

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 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, Recirculation 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).

6. S. Wang, Project Task Report, Tennessee Valley Authority Browns Ferry Unit 2 and Unit 3 Extended Power Uprate, Task T0313: RPV Flux Evaluation", GE-NE, San Jose, CA, March 2002 (GE-NE-A22-00125-19-01, Revision 0)(GE Proprietary Information).

E-4

GE Nuclear Energy GE-NE-0000-0013-3193-01 a APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

F-1

GE Nuclear Energy GE-NE-0000-0013-3193-01 a Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy of the beltline 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 4-4 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 I OCFR50 Appendix G for 30 EFPY of operation.

F-2

GE Nuclear Energy GE-NE-0000-0013-3193-Ola 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 18t Capsule Fluence 1.52 x 1017 n/cm2 1' Capsule Measured %Decrease = 4 (Charpy Curves) 1" Capsule R.G. 1.99 Predicted % Decrease = 9 (R.G. 1.99, Figure 2)

Limitina Beitline Plate (Heat C2467-1) USE:

%Cu = 0.16 30 EFPY 114T Fluence = 9.2 x 1017 n/cm2 R.G. 1.99 Predicted % Decrease = 14.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 14.5% s 21%, so vessel plates are bounded by equivalent margin analysis F-3

GE Nuclear Energy GE-NE-0000-001 3-3193-01 a 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 D55733) USE:

%Cu 2 0.20 1t Capsule Fluence = 1.52 x 1017 n/cm 2 Is Capsule Measured % Decrease = 3 (increase) (Charpy Curves)

I Capsule R.G. 1.99 Predicted % Decrease = 13 (R.G. 1.99, Figure 2)

Limiting Beltline Weld (Electroslap) USE:

%Cu = 0.24 30 EFPY 114T Fluence = 9.2 x 1017 n/cm2 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% s 34%, so vessel welds are bounded by equivalent margin analysis F-4

GE Nuclear Energy GE-NE-0000-0013-3193-01 a 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 Embrittlement 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 BWR/6 Vessels'", December 8, 1993.

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

5. C.l. Grimes (NRC) to Cart 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 License Renewal Rule (10 CFR 54.21)", 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-BI 100639-01, Revision 1).

F-5

W79 030520 005 GE Nuclear Energy General EkeMc Company 175 CurtnerAvenue, San Jose C4 95125 May 20, 2003 Action Requested by: N/A GE-ERJXARG-009 Response to: N/A DRF 0000-0013-3321 Project Deliverable: Yes cc: D. E. Porter (GE)

L. J. Tilly (GE)

F. E. Hartwig B. J. Branlund (GE)

Project Manager, Tennessee Valley Authority Browns Ferry Nuclear Plant P. O. Box 2000 Decatur, AL 35602 From: E. G. Thacker, II mi/c 747 175 Curtner Avenue San Jose, CA 95125

Subject:

T0301 P-T Curve Report for Browns Ferry Unit 3 - Final Non-proprietary

Reference:

1. TVA Contract No. 00001704, Release No. 00307, March 26, 2003

2. Letter GE-ERJXARG-007, "T0301 P-T Curve Reportfor Browns Ferry Unit 3

- Finalw/RIMS No.," May 13, 2003 GE has completed the P-T report for Browns Ferry Unit 3 authorized by Reference 1. The final report, a proprietary document, has been previously transmitted (Reference 2). Attached to this letter is the non-proprietary version of the Browns Ferry Unit 3 P-T curve report. The contents of the attachment have been design verified and evidence of verification is contained in DRF 0000-0013-3321.

A signed copy of this letter is included in DRF 0000-0013-3321. If you have any questions please call me or Betty Branlund at (408) 925-1472.

E. G. Thacker, 1!

Project Manager, Technical Projects 408 925-6154 att: GE-NE-0000-001 3-31 93-02a (non-proprietary)

k)

GE Nuclear Energy Engineering and Technology GE-NE-0000-001 3-31 93-02a General Electric Company Revision 0 175 Curtner Avenue Class I San Jose, CA 95125 May 2003 Pressure-Temperature Curves

- For TVA Browns Ferry Unit 3 Prepared by: Lf qaI y L.J. Tilly, Senior Engineer Structural Mechanics and Materials Verified by: BD TFrew B.D. Frew, Principal Engineer Structural Mechanics and Materials Approved by: WB1 Branfund B.J. Branlund, Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy GE-NE-000-001 3-31 93-02a INFORMATION NOTICE This document is the non-proprietary version of the General Electric Company (GE) document GE-NE-000-0013-3193-02. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [ ]. Large brackets, as shown on this paragraph, identify figures and large equation objects that could not be appropriately identified with the brackets shown above. J DISCLAIMER OF RESPONSIBILITY This document was prepared by the General Electric Company (GE) and is furnished solely for the purpose or purposes stated In the transmittal letter. No other use, direct or indirect, of the document or the Information it contains is authorized. Neither GE nor any of the contributors to this document:

• Makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of any information may not Infringe privately owned rights; or

• Assumes any responsibility for liability or damage of any kind that may result from any use of such Information.

Copyright, General Electric Company, 2003

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GE Nuclear Energy GE-NE-0000-0013-3193-02a 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 111; the P-T curves In this report represent 20 and 28 effective full power years (EFPY), where 28 EFPY represents the end of the 40 year license, and 20 EFPY is provided as a midpoint between the current EFPY and 28 EFPY. The P-T curve methodology includes the following: 1) the use of Kic from Figure A4200-1 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 [4J, in effect at the time of this evaluation. This report Incorporates a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14b], 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 12]:

• Closure flange region (Region A)

• Core beltline region (Region B)

• Upper vessel (Regions A & B)

• Lower vessel (Regions B & C)

- iv .

GE Nuclear Energy GE-N E-0000-0013-31 93-02a 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 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 14T and 314T locations because the maximum tensile stress for either heatup or cooldown is applied at the 14T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at I4T 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 20 and 28 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 20 and 28 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a TABLE OF CONTENTS

1.0 INTRODUCTION

1 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITLAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 14 43 PRESSURE-TEMPERATURE CURVE METHODOLOGY 20

5.0 CONCLUSION

S AND RECOMMENDATIONS 50

6.0 REFERENCES

73

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GE Nuclear Energy GE-NE000-001 3-31 93-02a TABLE OF APPENDICES APPENDIX A DESCRIPTION OF DISCONTINUITIES APPENDIX B PRESSURE-TEMPERATURE CURVE DATA TABULATION APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS APPENDIX D GE SIL 430 APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

- vii-

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE BROWNS FERRY UNIT 3 RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2. CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 31 FIGURE 4-3. FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 37 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] - 20 EFPY [150 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 53 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] - 20 EFPY 15'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 54 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 20 EFPY [15 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 55 FIGURE 5-4: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 20 EFPY [15 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 20 EFPY

[100 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 57 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 20 EFPY

[1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 58 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 20 EFPY

[1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-8: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 20 EFPY

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 FIGURE 5-9: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 20 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 20 EFPY [100F/HR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-11: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] - 28 EFPY

[15 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 63 FIGURE 5-12: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] - 28 EFPY [151F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64 FIGURE 5-13: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 28 EFPY [15F/HR OR LESS COOLANT HEATUP/COOLDOWN] 65 FIGURE 5-14: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 28 EPPY [15/HR OR LESS COOLANT HEATUP/COOLDOWN] 66

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GE Nuclear Energy GE-NE-0000-001 3-31 93-02a FIGURE 5-15: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 28 EFPY

[1000F/HR OR LESS COOLANT HEATUP/COOLDOWN] 67 FIGURE 5-16: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 28 EFPY

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 68 FIGURE 5-17: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 28 EFPY

[100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 69 FIGURE 5-18: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 28 EFPY

[1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 70 0

FIGURE 5-19: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 28 EFPY [100 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 71 FIGURE 5-20: COMPOSITE CORE NOT CRITICAL [CURVE B] INCLUDING BOTTOM HEAD AND CORE CRITICAL P-T CURVES [CURVE C] UP TO 28 EFPY [1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 72

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GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR BROWNS FERRY UNIT 3 VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR BROWNS FERRY UNIT 3 NOZZLE AND WELD MATERIALS 12 TABLE 4-3: RTNDT VALUES FOR BROWNS FERRY UNIT 3 APPURTENANCE AND BOLTING MATERIALS 13 TABLE 4-4: BROWNS FERRY UNIT 3 BELTLINE ART VALUES (20 EFPY) 18 TABLE 4-5: BROWNS FERRY UNIT 3 BELTLINE ART VALUES (28 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 BWRI4 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

GE Nuclear Energy GE-NE-0000-0013-3193-02a

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-beitline limits and irradiation embrittlement effects in the beltline.

Complete P-T curves were developed for 20 and 28 effective full power years (EFPY),

where 28 EFPY represents the end of the 40 year license, and 20 EFPY Is provided as a midpoint between the current EFPY and 28 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 [14a] calculated In accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER 14b], and is in compliance with Regulatory Guide 1.190. This fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MWt.

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 Kic 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 10CFR50.55a [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 InSection 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-0000-0013-31 93-02a 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 20 and 28 EFPY are included In Section 4.2. The peak ID fluence values of 2 2 8.9 1017 n/cm (20 EFPY) and 1.24 x 1018 n/cm (28 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-0000-0013-31 93-02a 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 Kic 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 [4J, 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 18] 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 [71.

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 3 vessel components. The non-beftline 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 beitline 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-0000-0013-3193-02a 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-000-0013-31 93-02a 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 3 Technical Specification 5], is calculated for a water temperature of 68°F.

The fluence is conservatively calculated using an EPU [14a] flux for the entire plant life.

The flux is calculated in accordance with Regulatory Guide 1.190.

GE Nuclear Energy GE-NE-0000-0013-3193-02a 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 60 0F above the qualification test temperature for the vessel materials.

The current requirements 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 III, Subsection NB-2300 are as follows:

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

b. RTNDT is defined as the higher of the dropweight NDT or 600F 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 (ST) of the bolting material, whichever is greater.

IOCFR50 Appendix G [8] 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-0000-001 3-31 93-02a converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data In WRC Bulletin 217 [9] 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 10), and approved by the NRC for generic use [11].

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

To establish the initial RTNDT temperatures for the Browns Ferry Unit 3 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 [12]). For Browns Ferry Unit 3 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 2F per ft-b energy difference from 50 ftlb.

For example, for the Browns Ferry Unit 3 beltline plate heat C3222-2 in the lower shell course; the lowest Charpy energy and test temperature from the CMTRs is 35 ft-lb at I 0OF. The estimated 50 ft-lb longitudinal test temperature is:

TSOL = I 0OF + [ (50 - 35) ft-lb 2Ftft-lb I = 400F 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 = 400F + 300F = 700F.

GE Nuclear Energy GE-NE-0000-0013-31 93-02a The Initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5Or 60'F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is -200F. Thus, the initial RTNDT for plate heat C3222-2 is I 0OF.

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 3, Heat AV1941, the NDT is 400F and the lowest CVN data is 37 ft-lb at 40°F. The corresponding value of (Taro 600F) is:

(TSOT - 600F) = { [40 + (50 - 37) ft-lb 20F/ft-lb I + 30 - 600 F = 360F.

Therefore, the initial RTNDT Is the greater of nil-ductility transition temperature (NDT) or (Twr- 600F), which is 401F.

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 3 (Heat C3067-2), the NDT is 400F and the lowest CVN data was 30 ft-lb at 400F. The corresponding value of (TOT - 60 0F) was:

(TSOT - 600F) = { [40 + (50 - 30) ft-lb 2Ffft-lb + 30°F - 60°F = 50 0F.

Therefore, the initial RTNDT was 500F.

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 60°F is the LST for the bolting materials. All Charpy data for the Browns Ferry Unit 3 closure studs did not meet the 45 ft-lb, 25 MLE requirements at 100F. Therefore, the LST for the bolting material is 70°F. The highest GE Nuclear Energy GE Energy GE-NE-0000-0013-3193-02a Nuclear 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 83.1'F, the bolt-up limit in the closure flange region.

The Initial RTNDT values for the Browns Ferry Unit 3 reactor vessel (refer to Figure 4-1 for the Browns Ferry Unit 3 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-000-001 3-31 93-02a p , -\ TOP HCAO

,,- 70P HEAD fANGE SSHELL FLANGE SHELL COURSE #5

Sz~e~z ~z'"zkV e' zz ZX Z . .. SHELICOURSE #4 7.SHELL COURSE #3 TOP O1S ACTIVE FUEL (TAF) 3.S ACTIE FUEL ACTIVE iBAF) 28.3' Z*

XAL%*ZD GIRTH WELDl (ESW~ $ .. .

SHELL CLOVRSE #2 SHELL COURSEw)

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 beitline region.

Figure 4-1: Schematic of the Browns Ferry Unit 3 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-N E-0000-0013-31 93-02a Table 4-1: RTNDT Values for Browns Ferry Unit 3 Vessel Materials Top Head &Flange Shell Flange (MK48) 48145-1 ARB 1031 10 52 63 90 -20 10 10 Top Head Flange (MK209) 209-145-1 AXG 90 10 70 120 110 -20 10 10 Top Head Dollar (MK201) 201-145-1 C2588-2B 40 76 68 76 10 40 40 Top Head Side Plates (MK202) 202-145-2 C3131-3 10 g0 55 49 18 10 10 202-122-4 C1498-3 10 68 55 63 -20 10 10 202-122-3 C1498-3 10 141 94 102 -20 10 10 202-146-2 C3262-4 10 100 74 83 -20 10 10 202-146-1 C3262-4 10 102 110 110 -20 10 10 202-127-4 C2426-1 10 70 54 54 -20 10 10 Shell Courses Upper Shell Plates (MK60) 6-145-8 C3132-2 10 67 79 61 -20 -20 -20 6-145-10 C3408-2 10 65 61 63 -20 -10 -10 6-145-13 C5582-2 10 50 70 42 -4 0 0 Transition Shell Plates (MK16) 15-145-1 C2775-1 10 64 53 54 -20 10 10 15-145-2 C2793-2 10 74 49 43 -6 10 10 15-145-3 87392-1 10 80 49 97 -18 10 10 Upper Intermediate Shell Plates (MK59) 6-145-3 C3170-1 10 59 67 66 -20 10 10 6-145-5 C3383-2 10 49 75 88 -18 10 10 6-145-11 C3170-2 10 91 74 64 .20 10 10 Lower-intermediate Shell Plates (MK58) 6-145-1 C3201-2 10 70 59 72 -20 -30 -20 6-145-2 C3188-2 10 83 70 92 -20 -30 -20 6-145-6 B72671 10 58 69 71 -20 -20 -20 Lower Shell Plates (MK57) 6-145-4 C3222-2 10 35 46 59 10 -20 10 6-145-7 C3213-1 10 60 69 77 -20 -30 -20 6-145-12 C3217-2 10 54 60 42 -4 -50 -4 Bottom Head Bottom Head Dollar (MK1) 1-145-1 C3067-2 40 31 30 35 50 40 50 Bottom Head Side Plates (MK2) 2-145-2 C2521-2 40 110 135 122 10 40 40 2-145-3 C3069-1 40 45 35 30 S0 40 50 2-146-3 87255-2 40 86 80 81 10 40 40 2-146-5 B7291-2 40 58 73 78 10 40 40 2-139-10 C2702-1 40 63 75 61 10 40 40 2-139-12 A1888-2 40 61 52 53 10 40 40 4-145-1 C3067-3 40 37 44 44 36 40 40 4-145-2 C3067-3 40 38 40 44 34 40. 40 4-145-3 C3067-1 40 26 36 32 58 40 58 4-145-4 C3067-1 40 36 40 35 40 40 40 NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a Table 4-2: RTNDT Values for Browns Ferry Unit 3 Nozzle and Weld Materials N1Redrc Outlet Nozzle (MK8) 8-145-1 AV2012 40 1 30 36 47 s0 40 50 8-145-2 AV2008 40 32 36 32 461 40 46 N2 Recir Inlet Nozzle (W7) 7-145-1 AV1941 7K-226B 40 37 70 79 36 40 40 7-145-2 AV1941 7K-6225B 40 99 94 90 10 40 40 7-145-3 AV1941 7K-226A 40 96 86 96 10 40 40 7-145-4 AV19317K-6222A 40 82 82 73 10 40 40 7-145-5 AV1931 7K-6222B 40 60 78 70 10 40 40 7-145-6 AV1931 7K-6223B 40 66 84 72 10 40 40 7-145-7 AV1931 7K-6223A 40 70 64 80 10 40 40 7-145-8 AV1931 7K-6224B 40 32 48 46 46 40 46 7-145-9 AV1941 7K-225A 40 109 95 82 10 40 40 7-145-10 AV1931 7K-8224A 40 70 105 85 10 40 40 N3 Steam Outlet Nozzle (MK14) 14-145-1 AV1985 8A4060 40 122 116 120 10 40 40 14-145-2 AV1986 8A-6061 40 46 40 39 32 40 40 14-1454 AV2037 8A-6187 40 63 86 83 10 40 40 14-145-4 AV2041 8A-6188 40 46 34 41 42 40 42 N4 Feedwater Nozzle (MKIO) 10-145-1 AV1909 87K-6127A 40 62 38 84 34 40 40 10-145-2 AV19097K-6127B 40 62 38 84 34 40 40 10-145-3 AV19457K-245A 40 64 38 76 34 40 40 10-145-4 AV1951 848189B 40 63 61 69 10 40 40 10-145-5 AV194577K-8245B 40 84 38 76 34 40 40 10-145-6 AV1951 8A-6189A 40 63 e1 69 10 40 40 NSCore Spray Nozzle (MK1) 11-145-1 AV1951 7K-248B 40 104 97 104 10 40 40 11-145-2 AV1951 7K-248A 40 106 114 109 10 40 40 N6 Top Head Instrumentation Nozzle (MK206) 206-127-3 &-4 ZT3043-4 40 170 158 142 10 40 40 N7 Top Head Vent Nozzle (MK204) 204-145-1 ZT3043-3 40 113 122 146 10 40 40 N8 Jet Pump Instrumentation (WI 9) 19-122-1 & 2 214484 40 26 34 35 58 40 58 N9 CRD HYD System Return Nozzle (W13) 13-139-2 EV91U 7N4020B 40 94 92 96 10 -10 10 N1O Core DP &Uquid Control Nozzle (MK17) 17-146-1 ZT3043-1 40 155 154 156 10 40 40 Ni1. N12. N16 Instrumentation Nozzle (MKI2 Inconel 12-145-3 trough 6 8601 12-1394& -7 071708 N13. N14 High &Low PressUe Seal Laak Detector (MK139)

B139-001 45E639415 1 10 40 31 32 18 40- 40 NIS Drain Nozzle (WK22) 22-145-1 7579 40 39 34 55 42 40- 42 WELDS:

Cyndickl Shel Adaf Welds Electroslag Welds ESW _ _ _ 23.1 Girth Welds Shell to Shah 2 (MK57 to MK58) D55733

• No NDT value available on CMTR; obtained from Purchase Specification 21A1111.

I Weld initial RTNOT values obtained from 1131.

NOTE: These are minimum Charpy values, GE Nuclear Energy GE-NE-0000-0013-31 93-02a Table 4-3: RTNDT Values for Browns Ferry Unit 3 Appurtenance and Bolting Materials 24-145-1 through -4 M846-1 40 1 122 i 109 1 127 1 10 1 30 1 30 Shroud Support (MK51, MKSZ MK53) Inconel 51-145-1 trough 4 8601-1A 52-145-1 & -2 8601-1A 53-145-1 through 8 8601-1 Steam Dryer Support Bracket (MK131) Stainless Steel 131 139506 Core Spray Bracket (MK132) Stainless Steel 132 3342230 _

Dryer Hold Down Bracket (MK133) 133 BT3078 40 119 119 112 10 40 40 133 BT2615 40 133 122 124 10 40 40 Guide Rod Bracket (MK134) Stainless Steel 134 139506  :

Feedwater Sparger Bracket (MK135) Stainless Steel 135 139506 _

Stabilizer Bracket (MK196) 196 ZT4777 40 93 70 109 10 40 40 Surveillance Bracket (MK199 &MK200) Stainless Steel 199,200 342633 Top Head Utlting Lugs (MK210) 210 A2918-3S 10 57 50 52 -20 10 10 CRD Stub Tubes (MK101 - MK128) Alloy 600 101-145 N/X0384-1 &-2 128-145 NX0433-1 &-2 Refueling Containment Skirt (MK71) 71-145-1 through4 B9593-5 10 127 108 115 -20 10 10 STUDS:

Closure llUK61) 3P2838 10 38 36 36 nta 70 NUTS:

Closure (MK62) 23514 10 49 53 63 29 10 WASHERS:

Closure (MK84 &MK65) 34328 10 32 36 48 20 70 NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 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 InTables 4-4 and 4-5 for 20 and 28 EFPY, respectively.

4.2.1 Regulatory Guide 1.99, Revision 2 (RG1.99) 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, aRTNDT = [CF] . f028-0101O0f)

Margin = 2(cr2 + OA) 0.5 CF = chemistry factor from Tables 1or 2 of RG1.99 f = 1/4T fluence 1019 Margin = 2(ai 2 + 2)05 al = standard deviation on initial RTNDT, which is taken to be 00F (13'F for electroslag welds).

aiA = standard deviation on ARTNDT, 28°F for welds and 17eF for base material, except that oE need not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT The margin term a, has constant values of 17°F for plate and 28°F for weld as defined in RG1.99. However, 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-0000-0013-3193-02a of a, is taken to be 0F for the vessel plate and most weld materials, except that A is 130F for the beitline electroslag weld materals [13].

4.2.1.1 Chemistry The vessel beltline chemistries were obtained from 1131.

The copper (Cu) and nickel (Ni) values were used with Tables I 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) [14a] flux for the vessel ID wall was calculated using methods consistent with Regulatory Guide 1.190. The flux in Reference 14 Is determined for the EPU rated power of 3952 MWt.

The peak fast flux for the RPV Inner surface from Reference 14 is 1.4e9 n/cm 2-s for EPU conditions. The calculated fast flux at the representative (Browns Ferry Unit 2 Cycle 7) capsule center is 8.85e8 n/cm2-s [14j with a corresponding lead factor of 0.98; Browns Ferry Unit 2 is used as a representative capsule because Browns Ferry Unit 3 has not yet removed a capsule. This calculation was performed prior to Regulatory Guide 1.190 (RG1.190), using methodology similar to RG1.190. Including the same bias adjustment as that applied to the RPV, the calculated 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 nlcm2-s [221 (with a lead factor of 0.98),

resulting in a calculation-to-measurement ratio of 1.6. The currently licensed Browns Ferry Unit 3 P-T curves are based upon a 32 EFPY fluence of 1.1e18 n/cm2, which was derived from the first cycle dosimetry flux of 1.04e9 ni/cm 2 -s.

28 EFPY Fluence Browns Ferry Unit 3 will begin EPU operation at approximately 13 EFPY, thereby operating for 15 EFPY at EPU conditions for 28 EFPY. As can be seen above, use of the EPU flux of 1.4e9 n/cm2-s to determine the fluence for the entire 28 EFPY (representing the 40 year GE Nuclear Energy GE-NE-0000-0013-31 93-02a Browns Ferry Unit 3 license period) Is conservative. The RPV peak IDfluence is calculated as follows:

1.4e9 n/cm2-s 8.83e8 s = 1.24e18 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 "0")[14a];hence the peak ID fluence used for these components is 1.0e8 n/cm 2. It was determined that the EPU axial flux distribution factor bounds the pre-EPU factor calculated during the 1995 Browns Ferry Unit 2 capsule evaluation 22].

The fluence at 14T is calculated per Equation 3 of Regulatory Guide 1.99, Revision 2 ]

using the Browns Ferry Unit 3 plant specific fluence and vessel thickness of 6.13". The 28 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds Is:

1.24e18 n/cm2 exp(-0.24 (6.13/4)) = 8.6e17 n/cm2.

The 28 EFPY 14T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

1.0e18 n/cm2 exp(0.24 * (6.1314)) = 6.9e17 n/cm 2.

20 EFPY Fluence The RPV peak ID fluence for 20 EFPY Is scaled from the 28 EFPY calculation above:

1.24e18 n/cm2 * (20 / 28) = 8.9e17 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 / lower shell location ratio of 0.81 for EPU conditions 114a]; hence the peak IDfluence used for these components Is 7.2e17 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 3 plant specific fluence and vessel thickness of 6.13". The 20 EFPY 114T fluence for the lower-intermediate shell plate and axial welds is:

GE Nuclear Energy GE-NE-0000-0013-31 93-02a 8.9e17 n/cm2 exp(-0.24 (6.13/4))=6.1e17 n/cm 2.

The 20 EFPY 1/4T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

7.2e17 n/cm2

• exp(-0.24 (6.13/4))=5.0e17n/cm2.

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 RTNDT. 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 20 and 28 EFPY, respectively.

GE Nuclear Energy GE-NE-0000-0013-31 93-02a Table 4-4: Browns Ferry Unit 3 Beltline ART Values (20 EFPY)

Lowulrimdlae Plate and Axdal Welds Thdmess. 6.13 hd"es 28 EFPYPeakI.D.11jance = 1.2E1+18 Wlfw,2 28 EFFYPeak l4T fence 6.61*17 dnfon2 2DEFPYPeak A Tfueice 6.II117 nkbo*2 L .er Plate and Aidall Welds idWLe lb Lowerbineadella 91M Weld Thickness 5 13 hds 281 EFFY Peak .D.fuence

• 1(.018 nkm-2 28EFPY Pe IM ltT uenc GAE117 foaym-2 2DEPY Peak1I4 T luece

• 6.0E+II7 nfcm'2 Iial 1r4T 20 EFPY 0i Co 20 EFPY 20 EFY COMPONENT HEATOR HEATILOT %Cu- %W CF Tdt Pewm A RFrK Margn Shf ART

______________ 1~~~~~~~~~~~'

,*gn'2 IF F 'F 'F PLATES:

6145-4 C3222-2 0.15 0.52 108 10 5DE*117 31 0 15 31 62 72

.1457 C3213-1 0.13 0.S 0 -20 8OE.17 28 0 13 2852 52 8.145-12 C3217-2 0.14 0.es 1015 4 5U0E117 30 0 15 30 so 5s Ler ldi awe S l 5145 1 C32D1-2 0.13 o.o *1 -20 E.1E117 30 0 16 30 89 39 s 145-2 C3188.2 0.10 0.45 65 .20 e 1E+17 21 0 11 21 42 22

.145-6 87287-1 0.13 0.51 s -20 811.17 28 0 14 2 7 37 WELDS Axial ESW 0.24 0.37 141 23.1 e.11+17 46 13 23 5. s8 122 GMi D55733 0.09 0.88 117 -40 s OE17 34 0 17 34 6s 28

• amiHsea dllainedinan('131.

GE Nuclear Energy GE-NE-0000-0013-3193-02a Table 4-5: Browns Ferry Unit 3 Beltline ART Values (28 EFPY)

Lower-etmedate Pkt and Axal Welds Thimcnss 0.13 Indes 28 EFPYPeakt ID. mce 12E18 frne2 28 EFPYPeak 114T luc

• 8E17 dWrn2 28EFPYP a 14T Sie- 8AE-17 rkm2 Lower Plate and Axu Weld, and Lower to Lowerjneiedctl OkM Weld Sa Thicknesse 8.13 dIches 28EFPYPeekl.DIketie* 1.OE 18 ntfcm2 28 EFPYPeak 114T ftiems = 69E+17 rxs2 28EFPYPeak 14Tblioe- 6.OE17 nss2 WileS 14T 28EFPPY C 0o 2 EY 28 EFPY COWONENT HEATORHEATh.OT %Cu- XM- CF RTdt Fumnce A RTdt Margn Shilt ART

'F sIma"^2 .F 'F F PLATE&

Lower Shell 6-145-4 Q222-2 0.15 0.52 105 10 6E+17 37 0 17 34 71 81 6145-7 C3213-1 0.13 0.58 0 -20 6.9.E17 31 0 16 31 e3 43 6-145-12 C217-2 0.14 0.66 101.6 -4 6.0E.17 36 0 17 34 6o 65 Lwrhbitrmdate Shell 6-145-1 Q2D1-2 0.13 0.60 01 -20 16E17 35 0 17 34 e0 48 6-145-2 C318-2 0.10 0.48 65 -20 &SE+17 25 0 13 25 S0 30 6-145- 7267-1 0.13 0 51 88 -20 &SE+17 34 0 17 34 8 48 WELDS:

Ada ESW 024 0.37 141 231 .SE-17 54 13 27 60 115 138 Ginr D5733 0.9 0.60 117 -40 .9E+17 41 0 20 41 81 41

• Ch sieelsrotAldriotms1131.

- 19 -

GE Nuclear Energy GE-NE-0000-0013-3193-02a 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 Appendix G [8J 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 IOCFR50 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 12J:

• 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 RTNDT. 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 [21 and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves l

GE Nuclear Energy GE-NE-00004001 3-31 93-02a are described in the sections below. 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 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 14T 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, Kir, at 1/4T to be less than that at 314T 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-0000-001 3-31 93-02a Table 4-6: Summary of the 10CFR50 Appendix G Requirements Hydrostatic Pressure I est & Le (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 + 600 F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 900F II. 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 RTNDT + 60'F*

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

b. Core critical - Curve C

1. At < 20% of preservice hydrotest Larger of ASME Limits + 400F or of a.I pressure, with the water level within the normal range for power operation

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

'hydrostatic pressure test

• 60 0F 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 IOCFR50 Appendix G [8]

requirements. The non-beltline and beltline region operating limits are evaluated according to procedures In 10CFR50 Appendix G 18], 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-0000-001 3-3193-02a 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltine 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.el 7 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 BWR/6 specifically for the purpose of fracture toughness analysis. The BWRI6 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 100F/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 BWRI6 components: the feedwater nozzle (FW) 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.

-4

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a Table 4-7: Applicable BWRI4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Water Level Instrumentation Nozzle Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Drain Nozzle Table 4-8: Applicable BWRI4 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B CRD and Bottom Top Head Nozz Recirculation Outlet Drain Nozzle Shell**

Support Skirt' Shroud r r Attachments-

• Support 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 BWRI6 (nominal inside diameter of 251 Inches). The analysis is considered appropriate for Browns Ferry Unit 3 as the plant specific geometric values are bounded by the generic GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 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 3 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 PressureTest - Non-Beltline, Curve A (Using Bottom Heac)

In a [ ] finite element analysis [ , the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K. The I I generic evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section Xl Appendix G 161 and shown below. The results of that computation were K = 143.6 ksi-in"2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top ofthe vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 84'F. [

I The limit for the coolant temperature change rate Is 15°F1hr or less.

GE Nuclear Energy GE-NE-0000-001 3-3193-02a I

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, t 2 = 2.83. The resulting value obtained was:

Mm = 1.85 for i<2 Mm = 0.926 Vi for 2< /i <3.464 = 2.6206 Mm = 3.21 for Jt >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Klb Is calculated from the equation in Paragraph G-2214.2 6]:

Ki = Mm - pm =1 ksi-inl'2 KWb = (2/3) Mm = I ] ksi-in"2 The total Kg is therefore:

K = 1.5 (Kim+ Kb) + Mm (asm + (2/3) crb) = 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 Ki is based on the K,0 equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) = In [(K, - 33.2)/20.734] / 0.02 GE Nuclear Energy GE-NE-0000-001 3-3193-02a (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 84'F X  :

The generic curve was generated by scaling 143.6 ksi-in'2 by the nominal pressures and calculating the associated (T - RTNDT):

Pressure Test CRD Penetration K and (T - RTNDT) as a Function Of Pressure f 0 : ~~~~~~~~~~~~~~~~....... ............. . .. ... . .. . .. . . . . . . .. . ..... ... ...

-Nominal Pressure 1 .T- T-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 580F, as shown in Tables 4-1 and 4-2. [

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a Second, the P-T curve is dependent on the calculated K value, and the K value is proportional to the stress and the crack depth as shown below:

Kim ca (a) 2 (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, Is t4. Thus, K, is proportional to R/(t)1m. The generic curve value of R/(t)m, based on the generic BWRI6 bottom head dimensions, is:

Generic: R / (t) m = 138/ (8) m = 49 inch 2 (4-2)

The Browns Ferry Unit 3 specific bottom head dimensions are R = 125.7 inches and t =8 inches minimum [19], resulting in:

Browns Ferry Unit 3 specific: R (t) m = 125.7 / (8)m = 44 inch's (4-3)

Since the generic value of R/(t) 2is larger, the generic P-T curve Is conservative when applied to the Browns Ferry Unit 3 bottom head.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 4.3.2.1.2 Core Not Critical Heatup/Cooldown - Non-Beitline 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. [

The calculated value of K, 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-in'm.

Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the K, equation of Paragraph A-4200 InASME Appendix A [171 for the core not critical curve:

(T - RTNDT) = In [(K, - 33.2)120.734] / 0.02 GE Nuclear Energy GE-NE-0000-001 3-3193-02a (T - RTNDT) = In [(191.5-33.2)/ 20.734] /0.02 (T - RTNDT) = 1020F The generic curve was generated by scaling 192 ksi-in'n by the nominal pressures and calculating the associated (T - RTNDT):

Core Not Critical CRD Penetration K and (T - RTNDT) as a Function of Pressure

-Nominal Presr , ;RN~

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 58°F, as shown in Tables 4-1 and 4-2. [

I 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-0000-013-3193-02a GE Nuclear Energy GE-NE-0000-0013-3193-02a 4.3.2.1.3 Pressure Test - Non-Beitline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress Intensity factor, K1, for the feedwater nozzle was computed using the methods from WRC 175 [15] together with the nozzle dimension for a generic 251-inch BWRI6 feedwater nozzle. The result of that computation was K,= 200 ksi-in"2 for an applied pressure of 1563 psig preservice hydrotest pressure. [

] The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1/4T through the corner thickness.

To evaluate the results, K, 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 K 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/rn) from Figure A5-1 of WRC-175 is 1.4 where:

a 1 (it 2 + t 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 r = nozzle radius (nozzle comer radius) = 3.75 inches Thus, a/r, = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (a) 1t

• F(a/rn):

GE Nuclear Energy GE-NE-0000-0013-3193-02a Nominal K = 1.5 34.97 ( 2.36) "2 1.4 = 200 ksi-in" 2 The method to solve for (T - RTNDT) for a specific K is based on the K equation of Paragraph A-4200 In ASME Appendix A [17] for the pressure test condition:

(T - RTNDT) = In [(K, - 33.2) / 20.734] /0.02 (T - RTNDT) = In [(200 - 33.2) / 20.7341/0.02 (T - RTNDT) = 104.20F The generic pressure test P-T curve was generated by scaling 200 ksi-n 2 by the nominal pressures and calculating the associated (T - RTNDT), [

1

-33 -

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a The highest RTNDT for the feedwater nozzle materials Is 400F as shown in Table 4-2.

However, the RTNDT was increased to 490 F to consider the stresses in the recirculation outlet nozzles together with the nitial RTNDT as described below. The generic pressure test P-T curve is applied to the Browns Ferry Unit 3 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 49 0F.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a I

Second, the P-T curve Is dependent on the K value calculated. The Browns Ferry Unit 3 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K are shown below:

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

• 125.7 inches / (6.125 Inches) = 32,077 psi. The dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding as =

35.04 ksi. The factor F (a/rn) from Figure A5-1 of WRC-175 is determined where:

i/4 ( t 2 + t 2)1r2 =2.32 inches a=

tn = thickness of nozzle = 6.96 inches tv = thickness of vessel = 6.125 inches rn= apparent radius of nozzle = r1+ 0.29 rc=6.9 inches r, = actual inner radius of nozzle = 6.0 inches rc = nozzle radius (nozzle comer radius) = 3.0 inches Thus, ar = 2.32 / 6.96 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an arn of 0.33, Is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (a) F(a/rn):

Nominal K, = 1.5

• 35.04 ( 2.32)"2 - 1.4 = 198.7 ksi-in" 2 GE Nuclear Energy GE-NE-0000-013-3193-02a 4.3.2.1.4 Core Not Critical Heatup/Cooldown - Non-Beitline 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 [ l finite element analysis done specifically for the purpose of fracture toughness analysis . 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 (Kip) Is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:

K1p= SF a (na)Y2 F(a/rr) (4-4) where SF is the safety factor applied per WRC Bulletin 176 recommended ranges, and F(a/rn) is the shape correction factor.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/rQ) 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 [ I design stress reports for the feedwater nozzle. The stresses considered are primary membrane, pm, and primary bending, a. Secondary membrane, sm, and secondary bending, asb, stresses are included in the total K1 by using ASME Appendix G [6] methods for secondary portion, K1s:

K = Mm (a + (213) CSb) (4-5)

GE Nuclear Energy GE-NE-0000-0013-3193-02a 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. K 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 K, was calculated, the following relationship was used to determine (T - RTNDT)-

The method to solve for (T - RTNDT) for a specific K, Is based on the Kic 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) / 20.734] / 0.02 (4-6)

Example Core Not Critical HeatuplCooldown Calculation for Feedwater NozzlelUpper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the [ l feedwater nozzle [ 3analysis, where feedwater injection of 40°F into the vessel while at operating conditions (551.4°F 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 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 (p) was adjusted for the actual [ ] 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:

CFpm = 24.84 ksi Fs = 16.19 ksi ay., = 45.0 ksi tv = 6.1875 inches Upb = 0.22 ksi asb = 19.04 ksi a = 2.36 inches rn = 7.09 inches t = 7.125 inches GE Nuclear Energy GE-NE-0000-001 3-31 93-02a In this case the total stress, 60.29 ksi, exceeds the yield stress, ay, 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 = lays - capm + ((ctAaI - Gys) / 30) / (am - apm) (47)

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

apm = 24.84 ksi s. = 9.44 ksi Cpb = 0.13ksi asb 11.IOksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 16]

was based on the 4a thickness; hence, t'2 = 3.072. The resulting value obtained was:

Mm = 1.85 for _it:2 Mm = 0.926 rt for 2< /i c3.464 = 2.845 Mm = 3.21 for Fi >3.464 The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an arn of 0.33, Is therefore, F(ar) =1.4 Kip is calculated from Equation 4-4:

Kip = 2.0 (24.84 + 0.13) (

• 2.36) 1 1.4 Kip = 190.4 ksi-in"2 Ki. is calculated from Equation 4-5:

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a KI, = 2.845 (9.44 23 11.10)

KI, = 47.9 ksi-in t 2 The total K, is, therefore, 238.3 ksi-inm.

The total K, 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 0F The [ l 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 400F water Injected into the hot reactor vessel nozzle. In the base case that yielded a K,value of 238 ksi-in1 2,the pressure is 1050 psig and the hot reactor vessel temperature is 551.4°F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (T., 8ft,.n - 40) /

(551.4 - 40). From K,the associated (T - RTNDT) can be calculated:

Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp. R . .. - -

-,fpSI9)~~~~~~.

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 K,.

GE Nuclear Energy GE-NE0000-0013-3193-02a The highest non-beltline RTNDT for the feedwater nozzle at Browns Ferry Unit 3 is 40 F as shown In Table 4-2. However, the RTNDT was increased to 490F to consider the stresses in the recirculation outlet nozzles as previously discussed. The generic curve is applied to the Browns Ferry Unit 3 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 490F as discussed in Section 4.3.2.1.3.

4.3.2.2 CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beltline 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 (K1), calculated for the beitline region according to ASME Code Appendix G procedures 161, 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 1000 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 Beltline 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-N E-0000-0013-3193-02a thickness (tOn) ratio of 15, Is treated as a thin-walled cylinder. The maximum stress Is the hoop stress, given as:

am = PR I tmin (4-8)

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

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with Kjc, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

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

for the pressure test condition:

Kim SF = K = 20.734 exp[0.02 (T - RTNDT)] + 33.2 (4-9)

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

respectively).

GE's current practice for the pressure test curve Is to add a stress Intensity factor, Ki, for a coolant heatup/cooldown rate, specified as 15'F/hr for Browns Ferry Unit 3, 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 100OF/hr. The Kt calculation for a coolant heatup/cooldown rate of 100 0F/hr is described in Section 4.3.2.2.3 below.

GE Nuclear Energy GE-NE-0000-0013-3193-02a 4.3.2.2.2 Calculations for the Bettline Region - Pressure Test This sample calculation is for a pressure test pressure of 1064 psig at 28 EFPY. The following inputs were used in the beltline limit calculation:

Adjusted RTNDT = Initial RTNDT + Shift A = 23 + 115 = 138'F (Based on ART values in Table 4-5)

Vessel Height H = 875.13 inches Bottom of Active Fuel Height B = 216.3 nches 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 (4-10)

= 1064 + (875.13 - 216.3) 0.0361 = 1088 psig Pressure stress:

a = PR/t (4-11)

= 1.088 125.7/6.13 = 22.3 ksi 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, t2 = 2.48. The resulting value obtained was:

Mm= 1.85 for ftIc2 Mm = 0.926 rt for 2< Fi <3.464 = 2.29 Mm 3.21 for fi >3.464 GE Nuclear Energy GE GE-NE-0000-0013-3193-02a Nuclear Energy The stress intensity factor for the pressure stress is Km = Mm a. The stress intensity factor for the thermal stress, K11, is calculated as described in Section 4.3.2.2.4 except that the value of "G"is 15OF/hr instead of 100F/hr.

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

Using the Kic equation of Paragraph A-4200 In ASME Appendix A [17], Km = 51.1, and Kit= 1.71 for a 15'F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

(T - RTNDT) = In[(1.5 Km + Kn- 33.2) 20.734] /0.02 (4-12)

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

= 38.90F T can be calculated by adding the adjusted RTNDT:

T = 38.9 + 138 = 176.9°F for P = 1064 psig at 28 EFPY 4.3.2.2.3 Beitline Region - Core Not Critical Heatup/Cooidown 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]:

Kic = 2.0 Kim +Kt (4-13) where Kim is primary membrane K due to pressure and Kt Is radial thermal gradient K due to heatup/cooldown.

The pressure stress intensity factor Kim 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.

GE Nuclear Energy GE-NE-000-001 3-3193-02a 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 M, from Figure G-2214-1 of ASME Appendix G [61 by the through-wall temperature gradient ATE, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G 16]. The relationship used to compute the through-wall ATw is based on one-dimensional heat conduction through an insulated flat plate:

a 2T(xt) / a x2 = 1/ (f(x,t) at) (4-14) where T(x,t) Is temperature of the plate at depth x and time t, and j3 is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that oT(x,t) I ot = dT(t) / dt = G where G is the coolant heatup/cooldown rate, normally 100Flhr. 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 / 2f - 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 16]. Therefore, AT, calculated from Equation 4-15 is used with the appropriate M,of Figure G-2214-1 of ASME Appendix G [61 to compute Kit for heatup and cooldown.

GE Nuclear Energy - GE-NE-0000-0013-3193-02a The M relationships were derived in the Welding Research Council (WRC)

Bulletin 175 [15] for infinitely long cracks of 14T. For the flat plate geometry and radial thermal gradient, orientation of the crack Is not important.

4.3.2.2.4 Calculations for the Beitline Region Core Not Critical Heatup/Cooldown This Browns Ferry Unit 3 sample calculation is for a pressure of 1064 psig for 28 EFPY.

The core not critical heatup/cooldown curve at 1064 psig uses the same Km 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 Kft term for the thermal stress. The additional inputs used to calculate Kg, are:

Coolant heatup/cooldown rate, normally 100 0F/hr G = 100 F/hr Minimum vessel thickness, Including dad thickness C = 0.526 ft (6.125m + 0.188" = 6.313")

Thermal diffusivity at 550"F (most conservative value) P = 0.354 fO2/ 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 = GC 2 /213 (4-16)

= 100 (0.526f / (2 0.354) = 390 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, K,, = M, 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):

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a (T - RTNDT) = In[((2 Kim + Kt)-33.2)/20.734] /0.02 (4-17)

= In[(2 51.1 + 11.39-33.2)120.734]/0.02

= 67.8 F T can be calculated by adding the adjusted RTNDT:

T = 67.8 + 138 = 205.8 F for P = 1064 psig at 28 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, K1. Using a 1/4T flaw size and the Kic formulation to determine T - RTNDT, for pressures above 312 psig the P-T limits for all flange regions are bounded by the I OCFR50 Appendix G requirement of RTNDT + 90 0F (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 3 at low pressures.

The approach used for Browns Ferry Unit 3 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 600 F adder is Included by GE for two reasons: 1)the pre-1971 GE Nuclear Energy GE-NE-0000-001 3-3193-02a requirements of the ASME Code Section III, Subsection NA, Appendix G Included the 60"F adder, and 2) inclusion of the additional 60'F requirement above the RTNDT provides the additional assurance that a 14T 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 represented by the electroslag weld materials In the upper shell at 23.10F, and the LST of the closure studs is 70*F; therefore, the bolt-up temperature value used is the more conservative value of 830F. 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 18] 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 + 90'F) and Curve B temperature no less than (RTNDT + 120 0F).

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 68F for the reason discussed below.

The shutdown margin, provided in the Browns Ferry Unit 3 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 68°F limit, further extensive calculations would be required to Justify a lower temperature. The 83°F limit for the upper vessel and beltline region and the 68@F 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 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFR50 Appendix G [8], Table 1. Table 1 of [8] requires that core critical P-T limits be 40"F 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 I 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 + 600F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 83°F, based on an RTNDT of 23.1VF. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160OF or the temperature required for the hydrostatic pressure test (Curve A at 1064 psig). The requirement of closure region RTNDT + 160OF causes a temperature shift in Curve C at 312 psig.

GE Nuclear Energy GE-NE-0000-001 3-3193-02a

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 12]:

• 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 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 12] and the nozzle thermal cycle diagrams [31. 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 apply for both heatuplcooldown 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, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a The following P-T curves were generated for Browns Ferry Unit 3:

• Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 20 and 28 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 Limits (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 20 and 28 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 20 and 28 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beftline, upper vessel, bottom head, and closure assembly P-T limits.

While the Bottom Head (CRD Nozzle) and Upper Vessel (FW Nozzle) curves are valid for the entire plant license period (28 EFPY), for clarity and convenience of Browns Ferry Unit 3 personnel, two (2) sets of these curves are provided, each with a designation of EFPY (either 20 or 28) 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 flux from Reference 14, the P-T curves are beitline limited above 600 psig for Curve A and above 540 psig for Curve B at 28 EFPY. At 20 EFPY, the P-T curves become beltline limited above 660 psig for Curve A and above 600 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-NE-0000-001 3-31 93-02a Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves 28 EFPY Curves A Bottom Head Limits (CRD Nozzle) - 28 EFPY Figure 5-11 Table B-3 A Upper Vessel Limits (FW Nozzle) - 28 EFPY Figure 5-12 Table B-3 A Belfine Limits - 28 EFPY Figure 5-13 Table 8-3 A Bottom Head and Composite Curve A - 28 EFPY* Figure 5-14 Table 8-4 B Bottom Head Limits (CRD Nozzle) - 28 EFPY Figure 5-15 Table B-3 B Upper Vessel Limits (FW Nozzle) -28 EFPY Figure 5-16 Table B-3 B Beltline Limits - 28 EFPY Figure 5-17 Table B-3 B Bottom Head and Composite Curve B - 28 EFPY* Figure 5-18 Table B-4 C Composite Curve C - 28 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 - 28 EFPY

• The Composite Curve A & B curve Is the more limiting of three limits: 10CFR50 Bolt-up Limits, Upper Vessel Limits (FW Nozzle), and Beltline Limits. 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: 10CFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Lmits (FW Nozzle), and Beltline Limits.

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 - p - - Y- P .  !- -

1300 4 4 4.

I E-.----4 4- 4 INITIAL RTndt VALUE IS 158'F FOR BOTTOM HEADl 1200 HEATUPItOOLDOWNI RATE OF COOLANT

< 15'F/HR 1100 a

A 1000 I. 900 0

.3 _SII

,,, 800 ACCEPTABLE AREA OF I0 o 700 OPERATION TO THE RIGHT OF THIS CURVE 0: 600 B.TTOM E 500 HEAD __

Uj E e B:

300 200 100 0

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

Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve A] - 20 EFPY

[1 50F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-3193-02a 1400 INITIAL RTndt VALUE IS 149*F FOR UPPER VESSELl 1300 HEATUP/COOLDOWN 1200 RATE OF COOLANT c 15'F/HR 1100 a 1000 0 900 0

ACCEPTABLE AREA OF 0

CL 800 900 OPERATION TO THE

0) RIGHT OF THIS CURVE I-0 700

'U 1% o300 00 30 202500 UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 0l 4.

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

Figure 5-2: Upper Vessel P-T Curve for Pressure Test Curve A] - 20 EFPY 115 0F/hr or less coolant heatup/cooldown)

GE Nuclear Energy GE-NE-0000401 3-31 93-02a 1400 1300 1200 BELTLINE CURVE AWUSTED AS SHOWN:

1100 EFPY SHIFT (°F) 20 99

a. 1000 a

HEATUPICOOLDOWN CL 900 RATE OF COOLANT 0 ' 15 F/HR

($ 800 R 700 ACCEPTABLE AREA OF M 600 OPERATION TO THE RIGHT OF THIS CURVE 3 500 a 400 300 200 100 0

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

Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 20 EFPY

[1 50 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 -

1407 . INITIAL RTndt VALUES ARE

/.' 23.1VF FOR BELTUNE, 1300 49°F FOR UPPER VESSEL, AND 8F FOR BOTTOM HEAD 1200 BELTLINE CURVES ADJUSTED AS SHOWN:

1100 -4_- EFPY SHIFT (F) 2

- 1000 _ .t / HEATUP/COOLDOWN RATE OF COOLANT 9900 < 15'FHR 0

,, 800

~700_ _ , _

• f l8 PSIG n ACCEPTABLE

.w AREA OF

.800. ___ OPERATION TO THE RIGHT OF THIS CURVE 500 BOTTOM . _

Uj HEAD .

ly. ~~~68.F 400 - __

300 i _ _ _

-UPPER VESSEL 200 - FLANGE AND BELTLNE REFO LIMITS BOTTOM HEAD 100 . CURVE 0 - - . - - _

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

(*F)

Figure 5-4: Composite Pressure Test P-T Curves [Curve A] up to 20 EFPY

[1 50 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-31 93-02a 1400 1300 HEATUPICOOLDOWN 1200 RATE OF bOOLANT

< 100'F/HR 1100 L

.- 1000 i

C 900 0

U X 800 ACCEPTABLE AREA OF OPERATION TO THE o 700 RIGHT OF THIS CURVE z 600 M 500

'U X 400 U

0.

300 _ _ _

200 HEiAD 00 _ HEA _ - BOTTOM HEAD LIM 100 - -_

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

Figure 5-5: Bottom Head P-T Curve for Core Not Critical [Curve B] - 20 EFPY

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

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTndt VALUE IS 49-F FOR UPPER VESSEL 1300 HEATUP/COOLDOWN 1200 RATE OF COOLANT c 100'FIHR 1100 c

C- 1000

a. 900 0

ul 10 800 to ACCEPTABLE AREA OF o 700 OPERATION TO THE RIGHT OF THIS CURVE m-E 600

500 uJ400 300 200

-UPPER VESSEL 30 LIMITS (Including 100 Flange and FW Nozzle Umits)

I I I I I -'I 1 I I 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (eF)

Figure 5-6: Upper Vessel P-T Curve for Core Not Critical [Curve B] - 20 EFPY

[100 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 -F 1300 -

r 1400- _ - - IINITIAL RTndt VALUE 23.1F FOR BELTLII 1200 -- _ _ _ _ _

BELTLINE CURVE ADJUSTED AS SHOVVN:

1100 - - - - EFPY SHIFT('I 20 99 C1000 _- -_ - I __

I HEATUP/COOLDOW VN

. 900 __/RATE OF COOLAN T 900 -

<~~~~~~~~~~

100oFHR 0m 0 500- - -

o700- - -

I-I K 600-Z , . / ACCEPTABLE AREA 0 F t 8540 PSG OPERATION TO THE 2 500 - _ RIGHT OF THIS CURVE 0,400 w

300-200 - - - --- -

I-OLR0 -BELTLINE LIMIT IBOLTIPI 100 - - _ 1 T 0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

-- (OF)

Figure 5-7: Beltline P-T Curve for Core Not Critical [Curve B] up to 20 EFPY

[1 00° F/hr or less coolant heatuplcooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-02a 1400 INITIAL RTndt VALUES ARE 23.1F FOR BELTLINE, 1300 49F FOR UPPER VESSEL, AND 58°F FOR BOTTOM HEAD 1200 BELTUNE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (-F) 20 99 a 1000 a HEATUPICOOWOWN RATE OF COOLANT IL 900 c 100*F/HR 0

UJ 0 800 o

I-700 ACCEPTABLE AREA OF 600 3 OPERATION TO THE RIGHT OF THIS CURVE L 500 to400

'U 0.

300

- UPPER VESSEL 200 AND BELTLINE LIMITS 100 BOTTOM HEAD .

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 20 EFPY

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

GE Nuclear Energy GE-NE-000001 3-31 93-02a 1400 INITIAL RTndt VALUES ARE 1300 23.1F FOR BELTLINE, 49'F FOR UPPER VESSEL, 1200 AND 58F FOR BOTTOM HEAD 1100

f. BELTLINE CURVE 1000 a

ADJUSTED AS SHOWN:

EFPY SHIFT ( F) 0 900 20 99 01 X 800 HEATUPICOOLDOWN

$ RATE OF COOLANT

< 100 F/HR o 700 t 600 _ . .__ I - __ _

/

l 500 I _ _IL - I - . - I l . . -

ACCEPTABLE AREA OF mu OPERATION TO THE RIGHT OF THIS CURVE 0 400 I l ,

I

_ _+

300

__1IIIII Minimum Criticality

/ I 200 Temperature 83F J BELTUNE AND 100 IX I __ _I_ NON-BELTLINE UIMITS 0 I 1~~~~~~~~~~~

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

Figure 5-9: Composite Core Critical P-T Curves [Curve CJ up to 20 EFPY

[1 000 F/hr or less coolant heatup/cooldownJ GE Nuclear Energy GE-NE-0000-0013-3193-02a B C 1400 I

INITIAL RTndt VALUES ARE 23.1F FOR BELTLINE, 1300 49F FOR UPPER VESSEL, 1200 I AND 58'F FOR BOTTOM HEAD BELTIINE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (F)

CD 20 99 a 1000 i;a HEATUPICOOLDOWN RATE OF COOLANT D 900 < 100rF/HR

-I Uj X 800 o 700 l-w 600 t

n 500

) 400 3U 300

-COMPOSITE CURVE B 200 BOTTOM HEAD .

CURVE B 100 - COMPOSITE CURVE C 0

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

Figure 5-10: Composite Core Not Critical [Curve B] including Bottom Head and Core Critical P-T Curves [Curve C] up to 20 EFPY [000 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000013-31 93-02a 1400 1300 1200 HEATUPICOOLDOWN RATE OF COOLANT

' 15°FHR 1100 la 1000 A-0 900 IA (0 800 I-U)

R 700 ACCEPTABLE AREA OF1 0 OPERATION TO THE 600 RIGHT OF THIS CURVEI Ir 500 I=

us cj: 400 300 200 100 0

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

Figure 5-11: Bottom Head P-T Curve for Pressure Test [Curve A] - 28 EFPY

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

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTndt VALUE IS 49F FOR UPPER VESSEL 1300 HEATUPICOOLDOWN 1200 RATE OF COOLANT

< 15FIHR 1100 CL

- 1000 2

0 900 0

U co 800 10 ACCEPTABLE AREA OF OPERATION TO THE o 700 RIGHT OF THIS CURVE I-

- 500 w 400 mu 300 200

-UPPER VESSEL UMITS (Including 100 Flange and FW Nozzle Lmits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(.F)

Figure 512: Upper Vessel P-T Curve for Pressure Test [Curve A] - 28 EFPY

[150 F1hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0013-3193-02a t 1400 INITIAL RTndt VALUE IS 1300 23.1-F FOR BELTLINE 1200 BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F) 1100 28 115 Is a.

-C 1000 i HEATUPICOOLDOWN RATE OF COOLANT 0- 900 _c15'FIHR 0

U-co 800 o 700 i 600 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 3 500 gA 400 300 200 100 0

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

Figure 5-13: Beltline P-T Curve for Pressure Test [Curve A] up to 28 EFPY

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

GE Nuclear Energy GE-NE-0000-0013-3193-02a 1400 1300 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (F) 28 115 1000 a

HEATUP/COOLDOWN IL 900 RATE OF COOLANT 0 c 15 FIHR en 800 o 700 W 600 z

, 500 400 300

- UPPER VESSEL 200 AND BELTLINE LIMITS BOTTOM HEAD .

100 CURVE 0

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

(*F)

Figure 5-14: Composite Pressure Test P-T Curves [Curve A] up to 28 EFPY

[1 5 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTndt VALUE IS l58F FOR BOTTOM HEAD 1300 1200 HEATUP/COOLDOWN RATE OF COOLANT

< 100°FIHR 1100 Is i; 1000 a

IL 900 0

u) 800 to o 700 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE 0 600 z

i 500 0 400 U

300 200 _ U IonoM _. __

200 - EAOI _ _BOTTOM I HEAD LIMI V I 68*F

~~ ~ ~ I 0 25 50 76 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-15: Bottom Head P-T Curve for Core Not Critical [Curve B - 28 EFPY

[1 00F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTndt VALUE IS 1300 49 F FOR UPPER VESSEL 1200 HEATUPICOOLDOWN RATE OF COOLANT 1100 S 100 F/HR n

-9 1000 a

z aL 900 0

LU (e 800 o 700 ACCEPTABLE AREA OF 0 0 OPERATION TO THE 0 600 RIGHT OF THIS CURVE 5o g 400 U

300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limts) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-16: Upper Vessel P-T Curve for Core Not Critical [Curve BJ - 28 EFPY

[1 000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTrdt VALUE IS 1300 23. 1F FOR BELTLINE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F) 1100 28 115 C.

o 1000 HEATUPICOOLDOWN z RATE OF COOLANT

a. 900 < 100FIHR 0

UJ tO 800 (a

o 700 t 600 ACCEPTABLE AREA OF OPERATION TO THE 2 500 RIGHT OF THIS CURVE w 400 0

300 200 100 0 4P 0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-17: Beltline P-T Curve for Core Not Critical [Curve BJ up to 28 EFPY 11 00°F1hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTndt VALUES ARE 23.1-F FOR BELTUNE, 1300 49-F FOR UPPER VESSEL, AND 58 F FOR BOTTOM HEAD 1200 BELTUNE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (-F)

.e 28 115

-1000 a

HEATUPICOOLDOWN 0- 900 RATE OF COOLANT 0 c 100°FIHR JI on 800 w

0 700 I-

600 _- - . 4- - - - - ACCEPTABLE AREA OF OPERATION TO THE t

RIGHT OF THIS CURVE 4 500 470 PSIGl; 1=' l ~~~50 PSIG 0: BO TM 0.

HEAD I 300

___ ~~~~~-UPPER VESSEL 200 ___ - -- - - - - ~~AND BELTUNE FLANGE LIMITS REGION... BOTTOM HEAD 100

- - - ~~~~~~~~~~CURVE 0

0 25 60 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-18: Composite Core Not Critical P-T Curves [Curve B] up to 28 EFPY

[1 000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a 1400 INITIAL RTndt VALUES ARE 1300 23.1-F FOR BELTLINE, 49'F FOR UPPER VESSEL, 1200 AND 58F FOR BOTTOM HEAD 1100 V

- 1000 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT CF)

I 28 115 IL 900 Eo 800 HEATUPICOOLDOWN RATE OF COOLANT

< 100°FIHR o 700 I-Z 600 2

M 500 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS CURVE I 400 300 200

-BELTLINE AND NON-BELTLINE 100 LIMITS 0

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

Figure 5-19: Composite Core Critical P-T Curves [Curve C) up to 28 EFPY

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

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a B C 1400 INITIAL RTndt VALUES ARE 23.1-F FOR BELTUNE, 49'F FOR UPPER VESSEL, 1300 AND 68'F FOR BOTTOM HEAD 1200 BELTUNE CURVES ADJUSTED AS SHOWN:

1100 EFPY SHIFT (-F) 28 115

- 1000 a HEATUPICOOLDOWN RATE OF COOLANT IL 900 I 100'FHR 0

U) 800 o 700 ACCEPTABLE AREA OF OPERATION TO THE j 60 0 RIGHT OF THIS CURVE 3 500 470 PSIG 450 PSIG

'* 400 300 - COMPOSITE CURVE B 200 BOTTOM HEAD .

CURVE B BOTTOM ./'/ lF^G HE_

100 68F - COMPOSITE CURVE C 0

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

Figure 5-20: Composite Core Not Critical [Curve B] including Bottom Head and Core Critical P-T Curves [Curve C] up to 28 EFPY [100°F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a

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 I (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 3.

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-0000-0013-3193-02a

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 3 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 (R08 981215 742)),

December 15,1998.

14. a) S. Wang, Project Task Report, Tennessee Valley Authority Browns Ferry Unit 2 and Unit 3 Extended Power Uprate, Task T0313: RPV Flux Evaluation, GE-NE, San Jose, CA, March 2002 (GE-NE-A22-00125-19-01, Revision 0)(GE Proprietary Information).

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

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

Welding Research Council Bulletin 175, August 1972.

16.

1

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-0000-0013-3193-02a 18.

19. Bottom Head and Feedwater Nozzle Dimensions:

a. Babcock & Wilcox Company Drawing 131838E, Revision 2, Lower Head Forming Details' (GE VPF 1974-017).

b. Babcock & Wilcox Company Drawing 131845E, Revision 2, "12" Feedwater Nozzle' (GE VPF 1974-037).

20. [

21. Materials - Properties", Part D to Section lt'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-B 1100639-01, Revision 1).

-75 -

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a

- APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a I

A-2

GE Nuclear Energy GE-NE-0000-0013-3193-02a 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 RTNDT plus 60TF. 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.

MK01- 128-Control Rod Drive Stub Alloy600 1,12,15,16 Nozzles madefrom Afloy 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.

MK132 - 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 X_________________ _ _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 evaluation.

MK 139*- N13 High and N14 Low Carbon Steel 1,24 Nota pressure boundary component Pressure Seal Leak Detection therefore requires no fracture toughness Penetration (Shell Flange) evaluation.

MK199, 200 - Surveillance Specimen SA276 304 1,21, 22 Appurtenances made from Stainless Brackets (Shells 58 and 59) Steel require no fracture toughness 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 high/low 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-0000-0013-3193-02a APPENDIX A

REFERENCES:

1. Vessel Drawings and Materials:

• Drawing #25469F, Revision 8, General Outline", Babcock & Wilcox Company, Mt. Vemon, Indiana (GE VPF #1974 050).

• Drawing #25470F, Revision 10, Outline Sections', Babcock & Wilcox Company, Mt. Vemon, Indiana (GE VPF #1974 051).

• Drawing #25471 F, Revision 9, 'Vessel Sub-Assembly, Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-052).

• Drawing #131834E, Revision 9, "List of Materials", Babcock & Wilcox Company, Mt. Vemon, Indiana (GE VPF #1974054).

• 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 #131838E, Revision 2, Lower Head Forming Details", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974017).

4. Drawing #131839E, Revision 4, Shell Segment Assembly Course #1 and #4",

Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-031).

5. Drawing #131843E, Revision 4, Recirculation Nozzles", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-035).

6. Drawing #131840E, Revision 5, Shell Segment Assembly Course #3", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-032).

7. Drawing #131845E, Revision 2, "12" Feedwater Nozzle", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-037).

8. Drawing #131846E, Revision 3, 10" Core Spray Nozzle", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-065).

9. Drawing #131847E, Revision 4, "2" Instrument and 4" CRD HYD System Return Nozzles", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-038).

10. Drawing #131841E, Revision 4, Shell Segment Assembly Course #5", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-033).

11. Drawing #131844E, Revision 2, 26" Steam Outlet Nozzle", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-036).

12. Drawing #131835E, Revision 9, "Lower Head Assembly", Babcock & Wilcox Company, Mt. Vemon, Indiana (GE VPF #1974-028).

A-4

GE Nuclear Energy GE-NE-0000-001 3-3193-02a

13. Drawing #131837E, Revision 8, aLower Head Upper Segment Assembly", Babcock

& Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-030).

14. Drawing #131848E, Revision 3, 4" Jet Pump Nozzle", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-039).

15. Drawing #131836E, Revision 6, Lower Head Bottom Segment Assembly", Babcock

& Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-029).

16. Drawing #149940E, Revision 2, Control Rod Nozzles", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-085).

17. Drawing #131855E, Revision 6, Closure Head Assembly", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-043).

18. Drawing #131856E, Revision 3, CIosure Head Nozzles", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974 044).

19. Drawing #131851E, Revision 4, "Support Skirt Assembly and Details", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-041).

20. Drawing #131849E, Revision 5, "Shroud Support", Babcock & Wilcox Company, Mt.

Vernon, Indiana (GE VPF #1974-040).

21. Drawing #131860E, Revision 8, "Vessel Subassembly Details", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-047).

22. Drawing #131850E, Revision 6, Vessel Attachment Details", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-023).

23. Drawing #142116E, Revision 3, Shell Segment Assembly Course #2", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974062).

24. Drawing #131842E, Revision 4, Shell Flange Details", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-034).

25. Drawing #131854E, Revision 1, Refueling Containment Skirt", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-042).

26. Drawing #131852E, Revision 1, Vessel Insulation Support", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974 024).

27. Drawing #131853E, Revision 1, Vessel Thermocouple Pads, Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-025).

A-5

GE Nuclear Energy GE-NE-000-001 3-31 93-02a APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B-1. Browns Ferry Unit 3 P-T Curve Values for 20 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 UPPER ~~~~~~Q EFPY"A

~~QUQM W.~~~~~~~~~~~~~~~~~~~ ~~~~

~~~......... ...... ..

0 68.0 83.0 83.0 68.0 83.0 83.0 10 68.0 83.0 83.0 68.0 83.0 83.0 20 68.0 83.0 83.0 68.0 83.0 83.0 30 68.0 83.0 83.0 68.0 83.0 83.0 40 68.0 83.0 83.0 68.0 83.0 83.0 50 68.0 83.0 83.0 68.0 83.0 83.0 60 68.0 83.0 83.0 68.0 83.0 83.0 70 68.0 83.0 83.0 68.0 83.0 83.0 80 68.0 83.0 83.0 68.0 83.0 83.0 90 68.0 83.0 83.0 68.0 83.0 83.0 100 68.0 83.0 83.0 68.0 83.0 83.0 110 68.0 83.0 83.0 68.0 83.0 83.0 120 68.0 83.0 83.0 68.0 83.0 83.0 130 68.0 83.0 83.0 68.0 83.2 83.0 140 68.0 83.0 83.0 68.0 86.4 83.0 150 68.0 83.0 83.0 68.0 89.2 83.0 160 68.0 83.0 83.0 68.0 91.9 83.0 170 68.0 83.0 83.0 68.0 94.5 83.0 180 68.0 83.0 83.0 68.0 96.9 83.0 190 68.0 83.0 83.0 68.0 99.2 83.0 200 68.0 83.0 83.0 68.0 101.3 83.0 210 68.0 83.0 83.0 68.0 103.3 83.0 220 68.0 83.0 83.0 68.0 105.3 83.0 230 68.0 83.0 83.0 68.0 107.1 83.0 B-2

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-1. Browns Ferry Unit 3 P-T Curve Values for 20 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

• Q....4 UP..PER 20UY ~ UM UPR 2~~

240 68.0 83.0 83.0 68.0 108.9 83.0 250 68.0 . 83.0 83.0 68.0 110.6 83.0 260 68.0 83.0 83.0 68.0 112.2 83.0 270 68.0 83.0 83.0 68.0 113.8 83.0 280 68.0 83.0 83.0 68.0 115.3 83.0 290 68.0 83.0 83.0 68.0 116.8 83.0 300 68.0 83.0 83.0 68.0 118.2 83.0 310 68.0 83.0 83.0 68.0 119.5 83.2 312.5 68.0 83.0 83.0 68.0 119.9 84.4 312.5 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 6,8.0 113.0 113.0 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 68.0 143.0 143.0 480 68.0 113.0 113.0 68.5 143.0 143.0 S-3

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B-1. Browns Ferry Unit 3 P-T Curve Values for 20 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 & 5-10 490 68.0 113.0 113.0 70.8 143.0 143.0 500 68.0 113.0 113.0 73.0 143.0 143.0 510 68.0 113.0 113.0 75.2 143.0 143.0 520 68.0 113.0 113.0 77.2 143.0 143.0 530 68.0 113.0 113.0 79.2 143.0 143.0 540 68.0 113.0 113.0 81.1 143.0 143.0 550 68.0 113.0 113.0 82.9 143.6 143.8 560 68.0 113.0 113.0 84.7 144.4 145.3 570 68.0 113.0 113.0 86.4 145.1 146.7 580 68.0 113.0 113.0 88.0 145.9 148.0 590 68.0 113.0 113.0 89.6 146.6 149.4 600 68.0 113.0 113.0 91.2 147.1 150.7 610 68.0 113.0 113.0 92.7 147.6 151.9 620 68.0 113.0 113.0 94.1 148.0 153.1 630 68.0 113.0 113.0 95.5 148.4 154.3 640 68.5 113.0 113.0 96.9 148.8 155.5 650 70.2 113.0 113.0 98.2 149.2 156.7 660 71.9 113.0 113.0 99.5 149.7 157.8 670 73.6 113.0 113.2 100.8 150.1 158.9 680 75.2 113.0 115.2 102.1 150.5 159.9 690 76.7 113.0 117.1 103.3 150.9 161.0 700 78.2 113.0 118.9 104.4 151.3 162.0 710 79.7 113.0 120.7 105.6 151.7 163.0 720 81.1 113.0 122.4 106.7 152.1 164.0 730 82.5 113.0 124.1 107.8 152.5 165.0 740 83.8 113.0 125.7 108.9 152.9 165.9 750 85.1 113.0 127.2 110.0 153.2 166.9 B-4

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-1. Browns Ferry Unit 3 P-T Curve Values for 20 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

.N

..W 760 86.4 113.8 128.7 111.0 153.6 167.8 770 87.6 114.6 130.2 112.0 154.0 168.7 780 88.8 115.3 131.6 113.0 154.4 169.6 790 90.0 116.1 133.0 114.0 154.8 170.4 800 91.2 116.9 134.3 114.9 155.1 171.3 810 92.3 117.6 135.6 115.9 155.5 172.1 820 93.4 118.4 136.9 116.8 155.9 172.9 830 94.5 119.1 138.2 117.7 156.2 173.8 840 95.5 119.8 139.4 118.6 156.6 174.6 850 96.6 120.5 140.6 119.4 156.9 175.3 860 97.6 121.2 141.7 120.3 157.3 176.1 870 98.6 121.9 142.9 121.1 157.6 176.9 880 99.5 122.6 144.0 122.0 158.0 177.6 890 100.5 123.3 145.0 122.8 158.3 178.4 900 101.4 123.9 146.1 123.6 158.7 179.1 910 102.4 124.6 147.1 124.4 159.0 179.8 920 103.3 125.2 148.2 125.1 159.4 180.5 930 104.1 125.9 149.2 125.9 159.7 181.2 940 105.0 126.5 150.1 126.7 160.0 181.9 950 105.9 127.1 151.1 127.4 160.4 182.6 960 106.7 127.7 152.0 128.1 160.7 183.3 970 107.6 128.3 153.0 128.9 161.0 183.9 980 108.4 128.9 153.9 129.6 161.4 184.6 990 109.2 129.5 154.8 130.3 161.7 185.2 1000 110.0 130.1 155.6 131.0 162.0 185.8 1010 110.7 130.7 156.5 131.6 162.3 186.5 1020 111.5 131.2 157.3 132.3 162.6 187.1 B-5

GE Nuclear Energy GE-NE-OO00-0013-31 93-02a TABLE B-1. Browns Ferry Unit 3 P-T Curve Values for 20 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

=p~~~~~~~~~~~~~~~~S 1030 112.3 131.8 158.2 133.0 163.0 187.7 1040 113.0 132.4 159.0 133.6 163.3 188.3 1050 113.7 132.9 159.8 134.3 163.6 188.9 1060 114.4 133.5 160.6 134.9 163.9 189.5 1064 114.7 133.7 160.9 135.2 164.0 189.7 1070 115.2 134.0 161.4 135.5 164.2 190.1 1080 115.9 134.5 162.1 136.2 164.5 190.7 1090 116.6 135.1 162.9 136.8 164.8 191.2 1100 117.2 135.6 163.6 137.4 165.1 191.8 1105 117.6 135.8 164.0 137.7 165.3 192.1 1110 117.9 136.1 164.4 138.0 165.4 192.4 1120 118.6 136.6 165.1 138.6 165.7 192.9 1130 119.2 137.1 165.8 139.2 166.0 193.4 1140 119.9 137.6 166.5 139.7 166.3 194.0 1150 120.5 138.1 167.2 140.3 166.6 194.5 1160 121.1 138.6 167.9 140.9 166.9 195.0 1170 121.8 139.1 168.5 141.4 167.2 195.6 1180 122.4 139.6 169.2 142.0 167.5 196.1 1190 123.0 140.1 169.9 142.5 167.7 196.6 1200 123.6 140.5 170.5 143.1 168.0 197.1 1210 124.2 141.0 171.2 143.6 168.3 197.6 1220 124.8 141.5 171.8 144.2 168.6 198.1 1230 125.3 141.9 172.4 144.7 168.9 198.6 1240 125.9 142.4 173.0 145.2 169.2 199.1 1250 126.5 142.8 173.6 145.7 169.4 199.6 1260 127.0 143.3 174.2 146.2 169.7 200.0 1270 127.6 143.7 174.8 146.7 170.0 200.5 B-6

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-1. Browns Ferry Unit 3 P-T Curve Values for 20 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 1280 128.1 144.2 175.4 147.2 170.2 201.0 1290 128.7 144.6 176.0 147.7 170.5 201.5 1300 129.2 145.0 176.6 148.2 170.8 201.9 1310 129.7 145.5 177.1 148.7 171.1 202.4 1320 130.3 145.9 177.7 149.2 171.3 202.8 1330 130.8 146.3 178.2 149.6 171.6 203.3 1340 131.3 146.7 178.8 150.1 171.8 203.7 1350 131.8 147.1 179.3 150.6 172.1 204.2 1360 132.3 147.6 179.9 151.0 172.4 204.6 1370 132.8 148.0 180.4 151.5 172.6 205.0 1380 133.3 148.4 180.9 151.9 172.9 205.4 1390 133.8 148.8 181.5 152.4 173.1 205.9 1400 134.3 149.2 182.0 152.8 173.4 206.3 B-7

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-2. Browns Ferry Unit 3 Composite P-T Curve Values for 20 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 R~~tVQ~~4 UPPER .. ..RPV & I '0QTTOM. UPPER RP.....V...P...

0 68.0 83.0 68.0 83.0 83.0 10 68.0 83.0 68.0 83.0 83.0 20 68.0 83.0 68.0 83.0 83.0 30 68.0 83.0 68.0 83.0 83.0 40 68.0 83.0 68.0 83.0 83.0 50 68.0 83.0 68.0 83.0 83.0 60 68.0 83.0 68.0 83.0 89.0 70 68.0 83.0 68.0 83.0 96.2 80 68.0 83.0 68.0 83.0 102.2 90 68.0 83.0 68.0 83.0 107.3 100 68.0 83.0 68.0 83.0 111.8 110 68.0 83.0 68.0 83.0 115.9 120 68.0 83.0 68.0 83.0 119.7 130 68.0 83.0 68.0 83.2 123.2 140 68.0 83.0 68.0 86.4 126.4 150 68.0 83.0 68.0 89.2 129.2 160 68.0 83.0 68.0 91.9 131.9 170 68.0 83.0 68.0 94.5 134.5 180 68.0 83.0 68.0 96.9 136.9 190 68.0 83.0 68.0 99.2 139.2 200 68.0 83.0 68.0 101.3 141.3 210 68.0 83.0 68.0 103.3 143.3 220 68.0 83.0 68.0 105.3 145.3 B-8

GE Nuclear Energy ;E-NE-0o00-0o13-31 93-02a TABLE B-2. Browns Ferry Unit 3 Composite P-T Curve Values for 20 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 04.U 1EU.1 240 68.0 83.0 68.0 108.9 148.9 250 68.0 83.0 68.0 110.6 150.6 260 68.0 83.0 68.0 112.2 152.2 270 68.0 83.0 68.0 113.8 153.8 280 68.0 83.0 68.0 115.3 155.3 290 68.0 83.0 68.0 116.8 156.8 300 68.0 83.0 68.0 118.2 158.2 I

310 68.0 83.0 68.0 119.5 159.5 312.5 68.0 83.0 68.0 119.9 159.9 312.5 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-0000-0013-3193-02a TABLE B-2. Browns Ferry Unit 3 Composite P-T Curve Values for 20 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 DOM.U 470 68.0 113.0 68.0 143.0 183.0 480 68.0 113.0 68.5 143.0 183.0 490 68.0 113.0 70.8 143.0 183.0 500 68.0 113.0 73.0 143.0 183.0 510 68.0 113.0 75.2 143.0 183.0 520 68.0 113.0 77.2 143.0 183.0 530 68.0 113.0 79.2 143.0 183.0 540 68.0 113.0 81.1 143.0 183.0 550 68.0 113.0 82.9 143.8 183.8 560 68.0 113.0 84.7 145.3 185.3 570 68.0 113.0 86A 146.7 186.7 580 68.0 113.0 88.0 148.0 188.0 590 68.0 113.0 89.6 149.4 189.4 600 68.0 113.0 91.2 150.7 190.7 610 68.0 113.0 92.7 151.9 191.9 620 68.0 113.0 94.1 153.1 193.1 630 68.0 113.0 95.5 154.3 194.3 640 68.5 113.0 96.9 155.5 195.5 650 70.2 113.0 98.2 156.7 196.7 660 71.9 113.0 99.5 157.8 197.8 670 73.6 113.2 100.8 158.9 198.9 680 75.2 115.2 102.1 159.9 199.9 690 76.7 117.1 103.3 161.0 201.0 700 78.2 118.9 104.4 162.0 202.0 B-10

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-2. Browns Ferry Unit 3 Composite P-T Curve Values for 20 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

......... UPER-T..

710 79.7 120.7 105.6 163.0 203.0 720 81.1 122.4 106.7 164.0 204.0 730 82.5 124.1 107.8 165.0 205.0 740 83.8 125.7 108.9 165.9 205.9 750 85.1 127.2 110.0 166.9 206.9 760 86.4 128.7 111.0 167.8 207.8 770 87.6 130.2 112.0 168.7 208.7 780 88.8 131.6 113.0 169.6 209.6 790 90.0 133.0 114.0 170.4 210.4 800 91.2 134.3 114.9 171.3 211.3 810 92.3 135.6 115.9 172.1 212.1 820 93.4 136.9 116.8 172.9 212.9 830 94.5 138.2 117.7 173.8 213.8 840 95.5 139.4 118.6 174.6 214.6 850 96.6 140.6 119.4 175.3 215.3 860 97.6 141.7 120.3 176.1 216.1 870 98.6 142.9 121.1 176.9 216.9 880 99.5 144.0 122.0 177.6 217.6 890 100.5 145.0 122.8 178.4 218.4 900 101.4 146.1 123.6 179.1 219.1 910 102.4 147.1 124.4 179.8 219.8 920 103.3 148.2 125.1 180.5 220.5 930 104.1 149.2 125.9 181.2 221.2 940 105.0 150.1 126.7 181.9 221.9 950 105.9 151.1 127.4 182.6 222.6 B-II

GE Nuclear Energy - GE-NE-0000-0013-3193-02a TABLE B-2. Browns Ferry Unit 3 Composite P-T Curve Values for 20 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 Ctfl WE. A

~~E~SUR~ *C ~ .. ............... W 960 106.7 152.0 128.1 183.3 223.3 970 107.6 153.0 128.9 183.9 223.9 980 108.4 153.9 129.6 184.6 224.6 990 109.2 154.8 130.3 185.2 225.2 1000 110.0 155.6 131.0 185.8 225.8 1010 110.7 156.5 131.6 186.5 226.5 1020 111.5 157.3 132.3 187.1 227.1 1030 112.3 158.2 133.0 187.7 227.7 1040 113.0 159.0 133.6 188.3 228.3 1050 113.7 159.8 134.3 188.9 228.9 1060 114.4 160.6 134.9 189.5 229.5 1064 114.7 160.9 135.2 189.7 229.7 1070 115.2 161.4 135.5 190.1 230.1 1080 115.9 162.1 136.2 190.7 230.7 1090 116.6 162.9 136.8 191.2 231.2 1100 117.2 163.6 137.4 191.8 231.8 1105 117.6 164.0 137.7 192.1 232.1 1110 117.9 164.4 138.0 192.4 232.4 1120 118.6 165.1 138.6 192.9 232.9 1130 119.2 165.8 139.2 193.4 233.4 1140 119.9 166.5 139.7 194.0 234.0 1150 120.5 167.2 140.3 194.5 234.5 1160 121.1 167.9 140.9 195.0 235.0 1170 121.8 168.5 141.4 195.6 235.6 1180 122.4 169.2 142.0 196.1 236.1 B-12

GE Nuclear Energy GE-WE-O000-0013-31 93-02a TABLE B-2. Browns Ferry Unit 3 Composite P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 Flhr for Curves B & C and 15 F/hr for Curve A For Figures 5-4, 5-8, 5-9 & 5-10

.. .... . ... ........AT. 1 JN AT L hJ E A 1190 123.0 169.9 142.5 196.6 236.6 1200 123.6 170.5 143.1 197.1 237.1 1210 124.2 171.2 143.6 197.6 237.6 1220 124.8 171.8 144.2 198.1 238.1 1230 125.3 172A 144.7 198.6 238.6 1240 125.9 173.0 145.2 199.1 239.1 1250 126.5 173.6 145.7 199.6 239.6 1260 127.0 174.2 146.2 200.0 240.0 1270 127.6 174.8 146.7 200.5 240.5 1280 128.1 175.4 147.2 201.0 241.0 1290 128.7 176.0 147.7 201.5 241.5 1300 129.2 176.6 148.2 201.9 241.9 1310 129.7 177.1 148.7 202.4 242.4 1320 130.3 177.7 149.2 202.8 242.8 1330 130.8 178.2 149.6 203.3 243.3 1340 131.3 178.8 150.1 203.7 243.7 1350 131.8 179.3 150.6 204.2 244.2 1360 132.3 179.9 151.0 204.6 244.6 1370 132.8 180.4 151.5 205.0 245.0 1380 133.3 180.9 151.9 205.4 245.4 1390 133.8 181.5 152.4 205.9 245.9 1400 134.3 182.0 152.8 206.3 246.3 B-13

GE Nuclear Energy Gt*E-0000-0013-31 93-02a TABLE B-3. Browns Ferry Unit 3 P-T Curve Values for 28 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 68.0 10 68.0 83.0 83.0 68.0 83.0 83.0 20 68.0 83.0 83.0 68.0 83.0 83.0 30 68.0 83.0 83.0 68.0 83.0 83.0 40 68.0 83.0 83.0 68.0 83.0 83.0 50 68.0 83.0 83.0 68.0 83.0 83.0 60 68.0 83.0 83.0 68.0 83.0 83.0 70 68.0 83.0 83.0 68.0 83.0 83.0 80 68.0 83.0 83.0 68.0 83.0 83.0 90 68.0 83.0 83.0 68.0 83.0 83.0 100 68.0 83.0 83.0 68.0 83.0 83.0 110 68.0 83.0 83.0 68.0 83.0 83.0 120 68.0 83.0 83.0 68.0 83.0 83.0 130 68.0 83.0 83.0 68.0 83.2 83.0 140 68.0 83.0 83.0 68.0 86.4 83.0 150 68.0 83.0 83.0 68.0 89.2 83.0 160 68.0 83.0 83.0 68.0 91.9 83.0 170 68.0 83.0 83.0 68.0 94.5 83.0 180 68.0 83.0 83.0 68.0 96.9 83.0 190 68.0 83.0 83.0 68.0 99.2 83.0 200 68.0 83.0 83.0 68.0 101.3 83.0 210 68.0 83.0 83.0 68.0 103.3 83.0 220 68.0 83.0 83.0 68.0 105.3 83.0 230 68.0 83.0 83.0 68.0 107.1 83.0 B-14

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B-3. Browns Ferry Unit 3 P-T Curve Values for 28 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 & 5-20 240 68.0 83.0 83.0 68.0 108.9 83.0 250 68.0 83.0 83.0 68.0 110.6 83.0 260 68.0 83.0 83.0 68.0 112.2 83.0 270 68.0 83.0 83.0 68.0 113.8 83.0 280 68.0 83.0 83.0 68.0 115.3 83.0 290 68.0 83.0 83.0 68.0 116.8 88.2 300 68.0 83.0 83.0 68.0 118.2 94.0 310 68.0 83.0 83.0 68.0 119.5 99.2 312.5 68.0 83.0 83.0 68.0 119.9 100.4 312.5 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 68.0 143.0 143.0 450 68.0 113.0 113.0 68.0 143.0 143.0 460 68.0 113.0 113.0 68.0 143.0 144.5 470 68.0 113.0 113.0 68.0 143.0 146.5 480 68.0 113.0 113.0 68.5 143.0 148.4 B-15

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a TABLE B-3. Browns Ferry Unit 3 P-T Curve Values for 28 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 & 5-20 490 68.0 113.0 113.0 70.8 143.0 150.2 500 68.0 113.0 113.0 73.0 143.0 151.9 510 68.0 113.0 113.0 75.2 143.0 153.6 520 68.0 113.0 113.0 77.2 143.0 155.2 530 68.0 113.0 113.0 79.2 143.0 156.8 540 68.0 113.0 113.0 81.1 143.0 158.3 550 68.0 113.0 113.0 82.9 143.6 159.8 560 68.0 113.0 113.0 84.7 144.4 161.3 570 68.0 113.0 113.0 86.4 145.1 162.7 580 68.0 113.0 113.0 88.0 145.9 164.0 590 68.0 113.0 113.0 89.6 146.6 165.4 600 68.0 113.0 113.0 91.2 147.1 166.7 610 68.0 113.0 115.2 92.7 147.6 167.9 620 68.0 113.0 117.8 94.1 148.0 169.1 630 68.0 113.0 120.3 95.5 148.4 170.3 640 68.5 113.0 122.7 96.9 148.8 171.5 650 70.2 113.0 124.9 98.2 149.2 172.7 660 71.9 113.0 127.1 99.5 149.7 173.8 670 73.6 113.0 129.2 100.8 150.1 174.9 680 75.2 113.0 131.2 102.1 150.5 175.9 690 76.7 113.0 133.1 103.3 150.9 177.0 700 78.2 113.0 134.9 104.4 151.3 178.0 710 79.7 113.0 136.7 105.6 151.7 179.0 720 81.1 113.0 138.4 106.7 152.1 180.0 730 82.5 113.0 140.1 107.8 152.5 181.0 740 83.8 113.0 141.7 108.9 152.9 181.9 750 85.1 113.0 143.2 110.0 153.2 182.9 B-1 6

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-3. Browns Ferry Unit 3 P-T Curve Values for 28 EFPY Required Coolant Temperatures at 100 OF/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 & 5-20 760 86.4 113.8 144.7 111.0 153.6 183.8 770 87.6 114.6 146.2 112.0 154.0 184.7 780 88.8 115.3 147.6 113.0 154.4 185.6 790 90.0 116.1 149.0 114.0 154.8 186.4 800 91.2 116.9 150.3 114.9 155.1 187.3 810 92.3 117.6 151.6 115.9 155.5 188.1 820 93.4 118.4 152.9 116.8 155.9 188.9 830 94.5 119.1 154.2 117.7 156.2 189.8 840 95.5 119.8 155.4 118.6 156.6 190.6 850 96.6 120.5 156.6 119A 156.9 191.3 860 97.6 121.2 157.7 120.3 157.3 192.1 870 98.6 121.9 158.9 121.1 157.6 192.9 880 99.5 122.6 160.0 122.0 158.0 193.6 890 100.5 123.3 161.0 122.8 158.3 194.4 900 101.4 123.9 162.1 123.6 158.7 195.1 910 102A 124.6 163.1 124.4 159.0 195.8 920 103.3 125.2 164.2 125.1 159.4 196.5 930 104.1 125.9 165.2 125.9 159.7 197.2 940 105.0 126.5 166.1 126.7 160.0 197.9 950 105.9 127.1 167.1 127.4 160.4 198.6 960 106.7 127.7 168.0 128.1 160.7 199.3 970 107.6 128.3 169.0 128.9 161.0 199.9 980 108.4 128.9 169.9 129.6 161.4 200.6 990 109.2 129.5 170.8 130.3 161.7 201.2 1000 110.0 130.1 171.6 131.0 162.0 201.8 1010 110.7 130.7 172.5 131.6 162.3 202.5 1020 111.5 131.2 173.3 132.3 162.6 203.1 B-17

GE Nuclear Energy GE-NE0000-001 3-31 93-02a TABLE B-3. Browns Ferry Unit 3 P-T Curve Values for 28 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 M T . PPR 2 FY.. OTTOM UPR 28EP VESSEL I-lEAD ~ELTN HEADftREVSE ETI 1030 112.3 131,8 174.2 133.0 163.0 203.7 1040 113.0 132.4 175.0 133.6 163.3 204.3 1050 113.7 132.9 175.8 134.3 163.6 204.9 1060 114.4 133.5 176.6 134.9 163.9 205.5 1064 114.7 133.7 176.9 135.2 164.0 205.7 1070 115.2 134.0 177.4 135.5 164.2 206.1 1080 115.9 134.5 178.1 136.2 164.5 206.7 1090 116.6 135.1 178.9 136.8 164.8 207.2 1100 117.2 135.6 179.6 137.4 165.1 207.8 1105 117.6 135.8 180.0 137.7 165.3 208.1 1110 117.9 136.1 180.4 138.0 165.4 208.4 1120 118.6 136.6 181.1 138.6 165.7 208.9 1130 119.2 137.1 181.8 139.2 166.0 209.4 1140 119.9 137.6 182.5 139.7 166.3 210.0 1150 120.5 138.1 183.2 140.3 166.6 210.5 1160 121.1 138.6 183.9 140.9 166.9 211.0 1170 121.8 139.1 184.5 141.4 167.2 211.6 1180 122.4 139.6 185.2 142.0 167.5 212.1 1190 123.0 140.1 185.9 142.5 167.7 212.6 1200 123.6 140.5 186.5 143.1 168.0 213.1 1210 124.2 141.0 187.2 143.6 168.3 213.6 1220 124.8 141.5 187.8 144.2 168.6 214.1 1230 125.3 141.9 188.4 144.7 168.9 214.6 1240 125.9 142.4 189.0 145.2 169.2 215.1 1250 126.5 142.8 189.6 145.7 169.4 215.6 1260 127.0 143.3 190.2 146.2 169.7 216.0 1270 127.6 143.7 190.8 146.7 170.0 216.5 B-18

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-3. Browns Ferry Unit 3 P-T Curve Values for 28 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 E$$URE CU1W........... . .. ..... .. ... .................

1280 128.1 144.2 191.4 147.2 170.2 217.0 1290 128.7 144.6 192.0 147.7 170.5 217.5 1300 129.2 145.0 192.6 148.2 170.8 217.9 1310 129.7 145.5 193.1 148.7 171.1 218.4 1320 130.3 145.9 193.7 149.2 171.3 218.8 1330 130.8 146.3 194.2 149.6 171.6 219.3 1340 131.3 146.7 194.8 150.1 171.8 219.7 1350 131.8 147.1 195.3 150.6 172.1 220.2 1360 132.3 147.6 195.9 151.0 172.4 220.6 1370 132.8 148.0 196.4 151.5 172.6 221.0 1380 133.3 148.4 196.9 151.9 172.9 221.4 1390 133.8 148.8 197.5 152A 173.1 221.9 1400 134.3 149.2 198.0 152.8 173.4 222.3 B-19

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B4. Browns Ferry Unit 3 Composite P-T Curve Values for 28 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 0 68.0 83.0 68.0 83.0 83.0 10 68.0 83.0 68.0 83.0 83.0 20 68.0 83.0 68.0 83.0 83.0 30 68.0 83.0 68.0 83.0 83.0 40 68.0 83.0 68.0 83.0 83.0 50 68.0 83.0 68.0 83.0 83.0 60 68.0 83.0 68.0 83.0 89.0 70 68.0 83.0 68.0 83.0 96.2 80 68.0 83.0 68.0 83.0 102.2 90 68.0 83.0 68.0 83.0 107.3 100 68.0 83.0 68.0 83.0 111.8 110 68.0 83.0 68.0 83.0 115.9 120 68.0 83.0 68.0 83.0 119.7 130 68.0 83.0 68.0 83.2 123.2 140 68.0 83.0 68.0 86.4 126.4 150 68.0 83.0 68.0 89.2 129.2 160 68.0 83.0 68.0 91.9 131.9 170 68.0 83.0 68.0 94.5 134.5 180 68.0 83.0 68.0 96.9 136.9 190 68.0 83.0 68.0 99.2 139.2 200 68.0 83.0 68.0 101.3 141.3 210 68.0 83.0 68.0 103.3 143.3 220 68.0 83.0 68.0 105.3 145.3 B-20

GE Nuclear Energy GE-NE-0000-0013-31 93-02a TABLE B-4. Browns Ferry Unit 3 Composite P-T Curve Values for 28 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 & 5-20 230 68.0 83.0 68.0 107.1 147.1 240 68.0 83.0 68.0 108.9 148.9 250 68.0 83.0 68.0 110.6 150.6 260 68.0 83.0 68.0 112.2 152.2 270 68.0 83.0 68.0 113.8 153.8 280 68.0 83.0 68.0 115.3 155.3 290 68.0 83.0 68.0 116.8 156.8 300 68.0 83.0 68.0 118.2 158.2 310 68.0 83.0 68.0 119.5 159.5 312.5 68.0 83.0 68.0 119.9 159.9 312.5 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-21

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B4. Browns Ferry Unit 3 Composite P-T Curve Values for 28 EFPY Required Coolant Temperatures at 100 F/hr for Curves B & C and 15 0F/hr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20 470 68.0 113.0 68.0 146.5 186.5 480 68.0 113.0 68.5 148.4 188.4 490 68.0 113.0 70.8 150.2 190.2 500 68.0 113.0 73.0 151.9 191.9 510 68.0 113.0 75.2 153.6 193.6 520 68.0 113.0 77.2 155.2 195.2 530 68.0 113.0 79.2 156.8 196.8 540 68.0 113.0 81.1 158.3 198.3 550 68.0 113.0 82.9 159.8 199.8 560 68.0 113.0 84.7 161.3 201.3 570 68.0 113.0 86.4 162.7 202.7 580 68.0 113.0 88.0 164.0 204.0 590 68.0 113.0 89.6 165.4 205.4 600 68.0 113.0 91.2 166.7 206.7 610 68.0 115.2 92.7 167.9 207.9 620 68.0 117.8 94.1 169.1 209.1 630 68.0 120.3 95.5 170.3 210.3 640 68.5 122.7 96.9 171.5 211.5 650 70.2 124.9 98.2 172.7 212.7 660 71.9 127.1 99.5 173.8 213.8 670 73.6 129.2 100.8 174.9 214.9 680 75.2 131.2 102.1 175.9 215.9 690 76.7 133.1 103.3 177.0 217.0 700 78.2 134.9 104.4 178.0 218.0 B-22

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B4. Browns Ferry Unit 3 Composite P-T Curve Values for 28 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 EAD

......- JE1JUN~AT

............... :.*: I EA 710 79.7 136.7 105.6 179.0 219.0 720 81.1 138.4 106.7 180.0 220.0 730 82.5 140.1 107.8 181.0 221.0 740 83.8 141.7 108.9 181.9 221.9 750 85.1 143.2 110.0 182.9 222.9 760 86.4 144.7 111.0 183.8 223.8 770 87.6 146.2 112.0 184.7 224.7 780 88.8 147.6 113.0 185.6 225.6 790 90.0 149.0 114.0 186.4 226.4 800 91.2 150.3 114.9 187.3 227.3 810 92.3 151.6 115.9 188.1 228.1 820 93.4 152.9 116.8 188.9 228.9 830 94.5 154.2 117.7 189.8 229.8 840 95.5 155.4 118.6 190.6 230.6 850 96.6 156.6 119.4 191.3 231.3 860 97.6 157.7 120.3 192.1 232.1 870 98.6 158.9 121.1 192.9 232.9 880 99.5 160.0 122.0 193.6 233.6 890 100.5 161.0 122.8 194.4 234.4 900 101.4 162.1 123.6 195.1 235.1 910 102.4 163.1 124.4 195.8 235.8 920 103.3 164.2 125.1 196.5 236.5 930 104.1 165.2 125.9 197.2 237.2 940 105.0 166.1 126.7 197.9 237.9 950 105.9 167.1 127.4 198.6 238.6 B-23

GENuclear Energy GE-NE-0000-0013-31 93-02a TABLE B4. Browns Ferry Unit 3 Composite P-T Curve Values for 28 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 & 5-20 106.7 128.1 239.3 970 107.6 169.0 128.9 199.9 239.9 980 108.4 169.9 129.6 200.6 240.6 990 109.2 170.8 130.3 201.2 241.2 1000 110.0 171.6 131.0 201.8 241.8 1010 110.7 172.5 131.6 202.5 242.5 1020 111.5 173.3 132.3 203.1 243.1 1030 112.3 174.2 133.0 203.7 243.7 1040 113.0 175.0 133.6 204.3 244.3 1050 113.7 175.8 134.3 204.9 244.9 1060 114.4 176.6 134.9 205.5 245.5 1064 114.7 176.9 135.2 205.7 245.7 1070 115.2 177.4 135.5 206.1 246.1 1080 115.9 178.1 136.2 206.7 246.7 1090 116.6 178.9 136.8 207.2 247.2 1100 117.2 179.6 137.4 207.8 247.8 1105 117.6 180.0 137.7 208.1 248.1 1110 117.9 180.4 138.0 208.4 248.4 1120 118.6 181.1 138.6 208.9 248.9 1130 119.2 181.8 139.2 209.4 249.4 1140 119.9 182.5 139.7 210.0 250.0 1150 120.5 183.2 140.3 210.5 250.5 1160 121.1 183.9 140.9 211.0 251.0 1170 121.8 184.5 141.4 211.6 251.6 1180 122.4 185.2 142.0 212.1 252.1 B-24

GE Nuclear Energy GE-NE-0000-0013-3193-02a TABLE B4. Browns Ferry Unit 3 Composite P-T Curve Values for 28 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 1190 123.0 185.9 142.5 212.6 252.6 1200 123.6 186.5 143.1 213.1 253.1 1210 124.2 187.2 143.6 213.6 253.6 1220 124.8 187.8 144.2 214.1 254.1 1230 125.3 188.4 144.7 214.6 254.6 1240 125.9 189.0 145.2 215.1 255.1 1250 126.5 189.6 145.7 215.6 255.6 1260 127.0 190.2 146.2 216.0 256.0 1270 127.6 190.8 146.7 216.5 256.5 1280 128.1 191.4 147.2 217.0 257.0 1290 128.7 192.0 147.7 217.5 257.5 1300 129.2 192.6 148.2 217.9 257.9 1310 129.7 193.1 148.7 218.4 258.4 1320 130.3 193.7 149.2 218.8 258.8 1330 130.8 194.2 149.6 219.3 259.3 1340 131.3 194.8 150.1 219.7 259.7 1350 131.8 195.3 150.6 220.2 260.2 1360 132.3 195.9 151.0 220.6 260.6 1370 132.8 196A 151.5 221.0 261.0 1380 133.3 196.9 151.9 221.4 261.4 1390 133.8 197.5 152.4 221.9 261.9 1400 134.3 198.0 152.8 222.3 262.3 B-25

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear Energy GE-NE-00O0-0013-31 93-02a CA 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 t 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-0000-0013-3193-02a 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 Heatup/Cooldown 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 LIMITS 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-0000-0013-3193-02a 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 I 00OF in one hour period heatup)

• Shutdown (coolant temperature change of less than or equal to 100F in one hour period cooldown)

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

C-4

GE Nuclear Energy GE-NE-0000-0013-3193-02a 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-000-001 3-31 93-02a

'APPENDIX D GE SIL 430 D-1

GE Nuclear Energy GE-N E-0000-0013-31 93-02a

'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 Use Limitations Steam dome saturation Primary measurement Must convert saturated temperature as determined above 2120 F for Tech steam pressure to from main steam Instrument Spec 1000F/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 2121F for Tech Must comply with SIL 251 Spec 1001F/hr heatup to avoid vessel stratification.

and cooldown rate.

Alternate measurement When above 2120 F need to above 2120F. allow for temperature variations (up to 10-15OF lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations 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).

RHR heat exchanger Alternate measurement Must have previously inlet coolant for Tech Spec 100 0F/hr correlated RHR inlet temperature cooldown rate when In coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow. Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta Ts will be ndicated coolant temperature. Delta T limit is 100F for BWR/6s and 1450 F for earlier BWRs.

Primary measurement to Must have drain line comply with Tech Spec flow. Use to verify brittle fracture compliance with Tech limits during cooldown. Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).

Alternate Information Must compensate for outside only measurement for metal temperature lag bottom head inside/ during heatup/cooldown.

outside metal surface Should have drain line flow.

temperatures.

D-3

GE Nuclear Energy GE-NE-000-001 3-31 93-02a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Closure head flanges Primary measurement for Use for metal (not coolant) outside surface T/Cs BWRI6s to comply with temperature. Install Tech Spec brittle fracture temporary T/Cs for metal temperature limit alternate measurement, if for head bolt-up. required.

One of two primary measure-ments for BWRI6s for hydro test.

RPV flange-to-shell Primary measurement for Use for metal (not coolant) junction outside BWRs earlier than 6s to temperature. -Response surface T/Cs comply with Tech Spec faster than closure head brittle fracture metal flange T/Cs.

temperature limit for head bolt-up.

One of two primary Use RPV closure head flange measurements for BWRs outside surface as alternate earlier than 6s for measurement.

hydro test. Preferred In lieu of closure head flange T/Cs If available.

RPV shell outside Information only. Slow to respond to RPV surface T/Cs coolant changes. Not available on BWRI6s.

Top head outside Information only. Very slow to respond to RPV surface T/Cs coolant changes. Not avail-able on BWR/6s.

D-4

GE Nuclear Energy GE-NE-00-0013-3 193-02a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Bottom head outside I of 2 primary measurements Should verify that vessel surface T/Cs to comply with stratification is not Tech Spec brittle fracture present for vessel hydro.

metal temperature (see SIL No. 251).

limit for hydro test.

Primary measurement to Use during heatup to verify comply with Tech Spec compliance with Tech Spec brittle fracture metal metal temperature/reactor temperature limits pressure curves.

during heatup.

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.

D-5

GE Nuclear Energy GE-NE0000-001 3-31 93-02a 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:

SILs 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 SILs 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 SLs. GE assumes no responsibility for liability or damage, which may result from the use of Information contained in SILs.

D-6

GE Nuclear Energy GE-NE-0000-001 3-31 93-02a APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy GE-NE0000-0013-3193-02a 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 1.0e7 n/cm 2, 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 "0":

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 Ni in Shell # 1 161.5" [2,3]

Top of Recirculation Outlet Nozzle N1 InShell # 1 188.0" [4]

Centerline of Recirculation nlet Nozzle N2 in Shell # 1 181.0" [2,3]

Top of Recirculation Inlet Nozzle N2 InShell # 1 193.3" [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 beltline 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 [2,31, 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-0000-0013-31 93-02a 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 .Oel 7 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 [6] 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

-11" above TAF. The beltline region considered in the development of the P-T curves is adjusted to include the additional 11 above the active fuel region and the additional 8' below the active fuel region. This adjusted beltline region extends from 208.3 to 377.3" above reactor vessel "0"for 28 EFPY.

Based on the above, it Is concluded that none of the Browns Ferry Unit 3 reactor vessel plates, nozzles, or welds, other than those Included in Tables 4-4 and 4-5, are in the beltline region.

E-3

GE Nuclear Energy GE-NE-0000-0013-3193-02a 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 Vesser, General Electric Company, GENE, San Jose, California.

3. Drawing #25469F, Revision 8, General Outline", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974050).

4. Drawing #131839E, Revision 4, Shell Segment Assembly Course #1 and

1. 40, Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-031).

5. Drawing #25471F, Revision 9, Vessel Sub-Assembly", Babcock & Wilcox Company, Mt. Vernon, Indiana (GE VPF #1974-052).

6. S. Wang, "Project Task Report, Tennessee Valley Authority Browns Ferry Unit 2 and Unit 3 Extended Power Uprate, Task T0313: RPV Flux Evaluation", GE-NE, San Jose, CA, March 2002 (GE-NE-A22-00125-19-01, Revision 0)(GE Proprietary Information).

E-4 IA

GE Nuclear Energy GE-NE-000-001 3-31 93-02a APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

F-1

GE Nuclear Energy GE-NE-000-0013-3193-02a Paragraph IV.B of IOCFR50 Appendix G [1] sets limits on the upper shelf energy of the beltline materials. The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be 28 EFPY. Calculations of 28 EFPY USE, using Regulatory Guide 1.99, Revision 2 [2J 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 3 beltline region. Therefore, Browns Ferry Unit 3 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 InTables F-1 and F-2. These calculations are based upon the 28 EFPY peak 1/4T fluence as provided in Tables 4-4 and 4-5. Surveillance capsule data is not available for Browns Ferry Unit 3.

Based on the results presented In Tables F-1 and F-2, the USE EMA values for the Browns Ferry Unit 3 reactor vessel beltline materials remain within the limits of Regulatory Guide 1.99, Revision 2 and IOCFR50 Appendix G for 28 EFPY of operation.

F-2

GE Nuclear Energy GE-NE0000-0013-31 93-02a Table F-I Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 3 For 28 EFPY (including Extended Power Uprate)

BWR/3-6 PLATE Surveillance Plate (Heat C3188-2) USE:

%Cu = 0.10 18 Capsule Fluence = Not Tested 81Capsule Measured  % Decrease = N/A (Charpy Curves) 81 Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limiting Beltline Plate (Heat C3222-2) USE:

%Cu = 0.15 28 EFPY 1/4T Fluence = 6.9 x 1017 n/cm2 R.G. 1.99 Predicted % Decrease = 13 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 13% 5 21%, so vessel plates are bounded by equivalent margin analysis F-3

GE Nuclear Energy GE-NE-0000-0013-3193-02a Table F-2 Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 3 For 28 EFPY (including Extended Power Uprate)

BWRI2-6 WELD Surveillance Weld (Electroslaol USE:

%Cu = 0.11 1 "tCapsule Fluence = Not Tested 1" Capsule Measured % Decrease = NIA (Charpy Curves) 1a Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limitina Beitline Weld (Electroslaa) USE:

%Cu = 0.24 28 EFPY 1/4T Fluence = 8.6 x 1017 n/Cm 2 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% s 34%, so vessel welds are bounded by equivalent margin analysis F-4

GE Nuclear Energy GE-NE-0000-0013-31 93-02a APPENDIX F

REFERENCES:

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

2. Radiation Embrittlement 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. L.A. England (BWR Owners' Group) to Daniel G. McDonald (USNRC), BWR 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 License Renewal Rule (10 CFR 54.21) , October 18, 2001.

, F-5