ML043440231

Nonproprietary Revision 0 to NEDO-33112, Pressure-Temperature Curves for TVA Browns Ferry Unit 1.

ML043440231
Person / Time
Site:	Browns Ferry
Issue date:	01/31/2004
From:	Tilly L General Electric Co
To:	Office of Nuclear Reactor Regulation, Tennessee Valley Authority
References
FOIA/PA-2005-0108 NEDO-33112
Download: ML043440231 (139)

Text

ENCLOSURE 5 BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 1 TECHNICAL SPECIFICATIONS (TS) CHANGE TS 428 UPDATE OF PRESSURE-TEMPERATURE (P-T) CURVES NON-PROPRIETARY SUPPORTING INFORMATION ATTACHED REPORT NEDO-33112 This report is an edited, non-proprietary versions of the full report provided in Enclosure 4.

GE Nuclear Energy Engineering and Technology NEDO-33112 General Electric Company DRF 0000-0010-6492 175 Curtner Avenue Revision 0 San Jose, CA 95125 Class I January 2004 Pressure-Temperature Curves For TVA Browns Ferry Unit I Prepared by: L7 UirY L.J. Tilly, Senior Engineer Structural Analysis and Hardware Design Verified by: B131 Branfund B.J. Branlund, Principal Engineer Structural Analysis and Hardware Design Approved by: MUR Schraq M.R. Schrag, Manager Structural Analysis and Hardware Design

NEDO-331 12 IMPORTANT NOTICE This is a non-proprietary version of the document NEDC-33112P, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [f B.

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between TVA and GE, PO #00001704, Unit 1 Restart Program, effective 4/2103, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than TVA, or for any purpose other than that for which it is furnished by GE, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

Copyright, General Electric Company, 2004

GE Nuclear Energy NEDO-33112 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 1994 [1]; the P-T curves in this report represent 12 and 16 effective full power years (EFPY), where 16 EFPY represents the end of the 40 year license, and 12 EFPY is provided as a midpoint between the current EFPY and 16 EFPY. The 1995 Edition of the ASME Boiler and Pressure Vessel Code including 1996 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Cases N-640 and N-588, 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. ASME Code Case N-640 allows the use of Kjc of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. ASME Code Case N-588 allows the use of an alternative procedure for calculating the applied stress intensity factors of Appendix G for axial and circumferential welds. This report incorporates a fluence calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190. This fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 MWt.

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

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

• Closure flange region (Region A) iii

GE Nuclear Energy NEDO-331 12

• Core beltline region (Region B)

• Upper vessel (Regions A & B)

• Lower vessel (Regions B & C)

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 15'FIhr or less must be maintained at all times.

The P-T curves apply for both heatup and cooldown and for both the 1/4T and 314T 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, Ki, at 114T 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 12 and 16 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 12 and 16 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

iv

GE Nuclear Energy NEDO-3311 2 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 INITIAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 14 43 PRESSURE-TENPERATURE CURVE METHODOLOGY 20

5.0 CONCLUSION

S AND RECOMMENDATIONS 53

6.0 REFERENCES

76 v

GE Nuclear Energy NEDO-3311 2 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) vi

GE Nuclear Energy NEDO-331 12 TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE BROWNS FERRY UNIT 1 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] - 12 EFPY [15F/HR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] - 12 EFPY [I5°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 57 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 12 EFPY [15 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 58 FIGURE 5-4: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 12 EFPY [150 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 12 EFPY

[1 00°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 12 EFPY

[I00°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 61 FIGURE 5-7: BELTL1NE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 12 EFPY

[100 0FIHR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-8: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 12 EFPY

[I00F/HR OR LESS COOLANT HEATUP/COOLDOWN] 63 FIGURE 5-9: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 12 EFPY [1 000 F/HR OR LESS COOLANT HEATUPICOOLDOWN] 64 FIGURE 5-10: COMPOSITE CORE NOT CRITICAL [CURVE B] INCLUDING BOTTOM HEAD AND CORE CRITICAL P-T CURVES [CURVE C] UP TO 12 EFPY [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 65 FIGURE 5-11: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] - 16 EFPY

[15°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 66 FIGURE 5-12: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] - 16 EFPY [1 5F/HR OR LESS COOLANT HEATUP/COOLDOWN] 67 FIGURE 5-13: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 16 EFPY [15'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 68 FIGURE 5-14: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 16 EFPY [15F/HR OR LESS COOLANT HEATUP/COOLDOWN] 69 vii

GE Nuclear Energy NEDO-331 12 FIGURE 5-15: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] - 16 EFPY

[1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 70 FIGURE 5-16: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] -16 EFPY

[IO0DFIHR OR LESS COOLANT HEATUP/COOLDOWN) 71 FIGURE 5-17: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 16 EFPY

[1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 72 FIGURE 5-18: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 16 EFPY

[1 000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 73 FIGURE 5-19: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 16 EFPY [100 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 74 FIGURE 5-20: COMPOSITE CORE NOT CRITICAL [CURVE B] INCLUDING BOTTOM HEAD AND CORE CRITICAL P-T CURVES [CURVE C] UP TO 16 EFPY [1000F/HR OR LESS COOLANT HEATUP/COOLDOWN] 75 viii

GE Nuclear Energy NEDO-3311 2 TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR BROWNS FERRY UNIT I VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR BROWNS FERRY UNIT I NOZZLE AND WELD MATERIALS 12 TABLE 4-3: RTMT VALUES FOR BROWNS FERRY UNIT I APPURTENANCE AND BOLTING MATERIALS 13 TABLE 4-4: BROWNS FERRY UNIT I BELTLINE ART VALUES (12 EFPY) 18 TABLE 4-5: BROWNS FERRY UNIT I BELTLINE ART VALUES (16 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 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 55 ix

GE Nuclear Energy NEDO-33112

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 12 and 16 effective full power years (EFPY),

where 16 EFPY represents the end of the 40 year license, and 12 EFPY is provided as a midpoint between the current EFPY and 16 EFPY. The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. This report incorporates a fluence calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190. This fluence represents an Extended Power Uprate (EPU) for the rated power of 3952 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 1994 [1]. The 1995 Edition of the ASME Boiler and Pressure Vessel Code including 1996 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Cases N-640 and N-588 [4], 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. ASME Code Case N-640 allows the use of Kic from Figure A-4200-1 of AppendixA [17] in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. ASME Code Case N-588 allows the use of an alternative procedure for calculating the applied stress intensity factors of Appendix G for axial and circumferential welds. 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.

1

GE Nuclear Energy NEDO-331 12 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 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 12 and 16 EFPY are included in Section 4.2. The peak ID fluence values of 5.3 x 1017 n/cm2 (12 EFPY) and 7.06 x 1017 n/cm2 (16 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).

2

GE Nuclear Energy NEDO-33112 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 1994 [1]. A detailed description of the P-T curve bases is included in Section 4.3. The 1995 Edition of the ASME Boiler and Pressure Vessel Code including 1996 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Cases N-640 and N-588, 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. ASME Code Case N-640 allows the use of K1c from Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. ASME Code Case N-588 allows the use of an alternative procedure for calculating the applied stress intensity factors to consider attenuation to reference flaw orientation of Appendix G for circumferential welds. This Code Case also provides an alternative procedure for calculating the applied stress intensity factor for axial welds. Other features presented are:

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

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

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

In addition to beltline considerations, there are non-beltline discontinuity limits such as nozzles, penetrations, and flanges that influence the construction of P-T curves. The non-beltline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable Browns Ferry Unit 1 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].

3

GE Nuclear Energy NEDO-331 12 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 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).

4

GE Nuclear Energy NEDO-331 12 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 1 Technical Specification [5], is calculated for a water temperature of 680F.

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

5

GE Nuclear Energy NEDO-3311 2 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 ill, 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 600 F 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 (LST) 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 6

GE Nuclear Energy NEDO-331 12 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 1 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 1 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 20F per ft-lb energy difference from 50 ft-lb.

For example, for the Browns Ferry Unit I beltline plate heat C2868-2 in the lower-intermediate shell course; the lowest Charpy energy and test temperature from the CMTRs is 25 ft-lb at 100F. The estimated 50 ft-lb longitudinal test temperature is:

T50L = 100F + [(50 - 25) ft-lb

• 20F/ft-lb ] = 600F The transition from longitudinal data to transverse data is made by adding 301F to the 50 ft-lb longitudinal test temperature; thus, for this case above, TsoT = 600 F + 300F = 90 0F.

7

GE Nuclear Energy NEDO-33112 The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (TSOr 600F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is 00F. Thus, the initial RTNDT for plate heat C2868-2 is 30 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 1, Heat ZT2869, the NDT is 300 F and the lowest CVN data is 31 ft-lb at 40'F. The corresponding value of (T5oT- 60 0F) is:

(TSOT - 60 0F) = { [40 + (50 - 31) ft-lb

• 20F/ft-lb ] + 30 } - 600 F = 48 0F.

Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5r- 60 0F), which is 480 F.

In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNDT. For the bottom head lower torus plate heat of Browns Ferry Unit 1 (Heat C1412-3), the NDT is 400 F and the lowest CVN data was 27 ft-lb at 401F. The corresponding value of (TSOT - 601F) was:

(T5oT - 60 0F) = { [40 + (50 - 27) ft-lb

• 20F/ft-lb ] + 300F } - 601F = 560F.

Therefore, the initial RTNDT was 560F.

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 600F is the LST for the bolting materials. Some Charpy data for the Browns Ferry Unit I closure studs did not meet the 45 ft-lb, 25 MLE requirements at 100F. Therefore, the LST for the bolting material is 701F. The highest 8

GE Nuclear Energy NEDO-3311 2 RTNDT in the closure flange region is 23.1'F, for the vertical electroslag weld material in the upper shell. Thus, the higher of the LST and the RTNDT +60 0 F is 83.1OF, the bolt-up limit in the closure flange region.

The initial RTNDT values for the Browns Ferry Unit 1 reactor vessel (refer to Figure 4-1 for the Browns Ferry Unit 1 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.

9

GE Nuclear Energy NEDO-33112 TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL COURSE #5 S, SHELL COURSE #4 SHELL COURSE #3 TOP OF SHELL COURSE #2 ACTIVE FUEL (TAF) 366.3- AXIAL WELD (ESW)

GI RTH WELD BOTTOM OF SHL/ORE#

ACTIVE FUELSHLCORE#

(BAF) 216.3-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 beltline region.

Figure 4-1: Schematic of the Browns Ferry Unit I RPV Showing Arrangement of Vessel Plates and Welds 10

GE Nuclear Energy NEDO-331 12 Table 4-1: RTNDT Values for Browns Ferry Unit 1 Vessel Materials

,, - -; - -- , ., -Drop,;

,, s Ha Tep Charpy EIegy (T5 crr-60) ;Welght. RTNu Comonent - Heat Temp -tl C y -) -NDT (F)

PLATES & FORGINGS:

Top Head & Flange Shell Flange (MK48) 48-127-1 ALU 55 10 118 90 79 -20 10 10 Top Head Flange (MK209) 209-127-1 AMW 56 10 120 109 126 -20 10 10 Top Head Dollar (MK201) Not 201-122-2 C-1354-3 40 67 58 51 10 Available 40 Top Head Side Plates (MK202) 202-122-1 A0057-2 10 45 41 55 -2 10 10 202-122-2 A0057-2 10 37 30 42 20 10 20 202-122-5 C1182-1 10 41 60 50 -2 10 10 202-122-6 C1182-1 10 57 45 62 -10 10 10 202-139-5 C2737-2 10 68 90 86 -20 10 10 202-139-6 C2737-2 10 84 78 86 -20 10 10 Shell Courses Upper Shell Plates (MK60) 6-127-7 A0973-1 10 57 60 41 -2 10 10 6-127-12 C1942-2 10 46 52 52 -12 10 10 6-127-19 C2496-2 10 56 58 54 -20 10 10 Transiton Shell Plates (MK16) 15-127-2 C-2533-2 10 53 48 43 -6 10 10 15-127-5 A-0954-3 10 46 40 32 16 10 16 15-127-6 A-0954-3 10 41 52 45 -2 10 10 Upper Intermediate Shell Plates (MK59) 6-127-3 B5842-1 10 63 62 61 -20 10 10 6-127-8 A0954-1 10 52 60 55 -20 10 10 6-127-10 B5853-2 10 70 65 66 -20 10 10 Lower Intermediate Shell Plates (MK58) 6-139-19 C2884-2 10 33 55 34 14 0 14 6-139-20 C2868-2 10 46 55 25 30 0 30 6-139-21 C2753-1 10 39 58 57 2 -20 2 Lower Shell Segments (MK57) 6-127-2 B5864-1 10 84 73 62 -20 -20 -20 6-127-4 A1009-1 10 62 84 77 -20 -10 -10 6-127-1 A0999-1 10 56 59 66 -20 -20 -20 Bottom Head Bottom Head Upper Torus (MK2) 2-122-7 B5924-1 40 75 70 75 10 40 40 2-122-8 B5924-1 40 37 61 44 36 40 40 2-122-10 A0942-2 40 62 62 65 10 40 40 2-127-7 C2412-3 40 91 90 57 10 40 40 2-127-8 C2412-3 40 95 92 82 10 40 40 2-127-9 C2393-2 40 105 125 112 10 40 40 Bottom Head Lower Torus (MK4) 4-122-5 A0927-2 40 71 50 59 10 40 40 4-122-6 A0927-2 40 75 66 64 10 40 40 4-122-7 C1412-3 40 30 41 40 50 40 50 4-122-8 C1412-3 40 27 35 49 56 40 56 Bottom Head Dollar (MK1) 1-122-2 B5861-1 40 45 50 49 20 40 40 NOTE: These are minimum Charpy values.

11

GE Nuclear Energy NEDO-331 12 Table 4-2: RTNDT Values for Browns Ferry Unit 1 Nozzle and Weld Materials Componerrt CmontHost He lxJor Test mp CbarpyEnergy (1.4 .7.` (T.w-60) Weight

--NDT' RT.5 T RCF) r t .tb) . CJ

.. (NOT (

Nozzles:

N1 Recirc Outlet Nozzle (MK8) 8-127-1 E31VW 431H-1 40 109 86 90 10 40 40 8-139-2 AV1696 7J-6327 40 34 46 44 42 40 42 N2 Recirc Inlet Nozzle (MK7) 7-122-1 ZT2872 9709-1 40 65 54 58 10 30 30 7.122.7 ZT2869 9704-1 40 31 38 39 48 30 48 7-127-9 E25VW433H-9 40 93 103 110 10 40 40 7-122-10 ZT2872 9709-2 40 65 54 58 10 30 30 7-122-11 ZT28859711-1 40 34 37 41 42 30 42 7.122-12 ZT2885 9711-2 40 34 37 41 42 30 42 7-122-13 ZT2885-3 9710-2 40 64 48 36 39 30 39 7-122-16 ZT-2885 40 38 38 48 35 30 35 7.122-18 ZT2885 9712-2 40 54 52 49 12 30 30 7.122-19 ZT2869 9705-1 40 69 53 42 27 30 30 N3 Steam Outlet Nozzle (MK14) 14-127-1 E26VW 435H-1 40 97 77 94 10 40 40 14-127-2 E26VW 435H-2 40 86 84 76 10 40 40 14-127-3 E26VW 435H-3 40 102 92 105 10 40 40 14-127-4 E26VW 435H-4 40 119 94 94 10 40 40 N4 Feedwater Nozzle (MK10)10-127.1 E25VW 436H-1 40 98 97 92 10 40 40 10-127.2 E25VW 436H-2 40 124 98 105 10 40 40 10-127.3 E25VW 436H-3 40 99 84 92 10 40 40 10-127-4 E25VW436H.4 40 111 98 101 10 40 40 10-12745 E25VW 436H-S 40 114 114 110 10 40 40 10-127-6 E25VW 436H-6 40 117 111 112 10 40 40 NS Core Spray Nozzle (MK11) 11-111-1 BT2001-2 7098 40 48 32 42 46 40 46 11-111-2 BT2001-3 6945-1 40 54 36 49 38 40 40 N6 Top Head Instrumentation Nozzle (MK206) 206-139-1 &.4 BT2615.4 40 123 143 144 10 40 40 N7 Vent Nozzle (MK2C4) 204-127-1 ZT3043-3 40 102 130 117 10 40 40 N8 Jet Pump Instr. Nozzle (MKt9) 19-127-1& -2 ZT3043 40 107 112 113 10 40 40 N9 CRD HYD System Return Nozzle tMK13) 13-145-1 EV9793 7K46233A 40 81 50 91 10 40 40 N10 Core DP &Liquid Control Nozzle (MK17) 17-127-1 ZT3043 40 106 136 111 -20 40 40 N1i. N12. NI6 Instvrumentation Nozzle MK12) Inconel 12-127-1 through 6 8564 N13. N14 High &Low Pressure Seal Leak (MKI39) 139-127-1 &I 2 Not Available 40 40-N15 Drain Nozzle (MK22) 22-127-1 213099 40 42 44 39 32 40 40 WELDS:

Cylndrical SheN Axial Welds Electroslag Welds ESW 23.1 Geth Welds Shell I to Shell2 WF 154 (SAW) 406L44. Lot 8720 20

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

• Weld initial RTNDT values obtained from [13].

NOTE: These are minimum Charpy values.

12

GE Nuclear Energy NEDO-33112 Table 4-3: RTNDT Values for Browns Ferry Unit 1 Appurtenance and Bolting Materials

- Drop

-Test

- -Charpy Energy 4TT-60), :Weight RTwT

-Component Heat H. Tiemp 'NT - (F)

Miscellaneous Appurtenances:

Support Skirt Segment (MK24) 24-139-1 through-4 C3888-5 10 38 40 41 4 40 40 Shroud Support (MK51. MK52. MK53) Ahoy 600 Steam Dryer Support Bracket (MK131) Stainless Steel 131-127-1 through-4 00431 Core Spray Bracket (MK132) Stainless Steel 132-127.1 through -8 3342230 Dryer Hold Down Bracket (MK133) 133-127-1 through -4 EV-8446 40 38 42 37 36 40 40 Guide Rod Bracket (MK134) Stainless Steel 134-127-1 & -2 139506 Feedwater Sparger Brackets (MK135) Stainless Steel 135-127-1 through -12 00431 Stabilizer Bracket (MK196) 196-127-5 through -12 C6458-1 10 60 59 56 -20 40 40 Surveiltance Brackets (MK199 & MK200) Stainless Steel 199-127-1 through -3 342633-2 200-127-1 through -3 342633-2 Lifting Lugs (MK210) 210-122-1. -2. -3. & -6 A1210-3B 10 83 98 95 -20 40 40 CRD penetrations (MK101 - MK128) Alloy 600 101 through 128 Refueling Containment Skirt (MK71) 71-127-1 through -4 87478-4B _ 40-Componen - I :3.nergs

- - -t.: '.Te-stp

-T: ~l Charpy

• ..C;iryEnr~

Lateral -,

Expansion IS.

- I _ 1 -.- r%

_ I .. .I'. - -Z ,( - -

STUDS:.

Closure (MK61) 6730502 10 34 52 68 nWa 70 NUTS:

Closure (MK62) 6730502 10 34 52 68 n/a 70 23514 10 49 53 63 29 10 6780382 10 45 42 46 n/a 70 6790156 n/a n/a n/a n/a n/a 70 BUSHINGS:

Closure (MK63) T3798 10 61 69 73 51 10 M2513 10 64 65 67 40 10 M2514 10 66 56 70 42 10 EV9474 10 67 64 62 n/a 70 AV3107 10 63 70 72 n/a 70 WASHERS:

Cbosure (MK64 and MK65) 6730502 10 34 52 68 n/a 70 1 6780278 nta n/a n/a n/a n/a 70

• No Charpy or NDT values available on CMTR; obtained from Purchase Specification 21A111.

NOTE: These are minimum Charpy values.

13

GE Nuclear Energy NEDO-3311 2 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG1.99) provides the methods for determining the ART. The RG1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beltline plates and welds was performed and is summarized in Tables 4-4 and 4-5 for 12 and 16 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 8

where, ARTNDT = [CF]

• f (O2 -Oi01Og l

Margin = 2(a2 + %2)0.5 CF = chemistry factor from Tables 1 or 2 of RG1.99 f = AT fluence / 1019 Margin = 2(al2 + ;A2 ) 0.5 ayl =standard deviation on initial RTNDT, which is taken to be 00F (130 F for electroslag welds).

CT& =standard deviation on ARTNDT, 280F for welds and 170F for base material, except that ca need not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT The margin term as has constant values of 170F for plate and 280F for weld as defined in RG1.99. However, a, need not be greater than 0.5

• ARTNDT. Since the GEIBWROG 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 14

GE Nuclear Energy NEDO-33112 of a, is taken to be 0°F for the vessel plate and most weld materials, except that a, is 130F for the beltline electroslag weld materials and 10F for the beltline SAW girth weld material [13].

4.2.1.1 Chemistry The vessel beltline plate chemistries were obtained from [1] and the beltline weld material chemistries were obtained from [13].

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 (ca) is specified.

4.2.1.2 Fluence An EPU (Extended Power Uprate) flux for the vessel ID wall was calculated using methods consistent with Regulatory Guide 1.190, and is determined for the EPU rated power of 3952 MWt. The peak fast flux for the RPV inner surface, used for determination of the P-T curves, is 1.4e9 n/cm2 -s for EPU conditions.

For comparison, the calculated fast flux at the representative (Browns Ferry Unit 2 Cycle 7) capsule center is 8.85e8 n/cm2-s with a corresponding lead factor of 0.98; Browns Ferry Unit 2 is used as a representative capsule because Browns Ferry Unit 1 has not yet removed a capsule. This calculation was performed prior to Regulatory Guide 1.190 (RG1.190), using methodology similar to RG1.190. ((

)), the calculated fast flux at this capsule is 9.5e8 nlcm2 -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 12 EFPY Browns Ferry Unit 1 P-T curves are based upon a 32 EFPY fluence of 7.6e17 n/cm2 [1j.

15

GE Nuclear Energy NEDO-33112 16 EFPY Fluence Browns Ferry Unit I will begin EPU operation at approximately 6 EFPY, thereby operating for 10 EFPY at EPU conditions for a total of 16 EFPY. As can be seen above, use of the EPU flux of 1.4e9 n/cm2-s to determine the fluence for the entire 16 EFPY (representing the 40 year Browns Ferry Unit 1 license period) is conservative. The RPV peak ID fluence is calculated as follows:

1.4e9 n/cm2 -s

• 5.05e8 s = 7.06e17 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 p0"); hence the peak ID fluence used for these components is 5.72e17 n/cm2 . It was determined that the fluence calculated using the EPU flux (1.4e9 n/cm2-s) with the corresponding distribution factor of 0.81 bounds the fluence calculated using the pre-EPU flux (9.5e8 nlcm2-s) with the corresponding factor of 0.86 calculated during the 1995 Browns Ferry Unit 2 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 1 plant specific fluence and vessel thickness of 6.13". The 16 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

7.06e17 n/cm2

• exp(-0.24 * (6.13 /4)) = 4.89e17 n/cm2.

The 16 EFPY 1/4T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

5.72e17 n/cm2

• exp(-0.24 * (6.13 / 4)) = 3.96e17 n/cm 2.

12 EFPY Fluence The RPV peak ID fluence for 12 EFPY is scaled from the 16 EFPY calculation above:

7.06e17 n/cm2 * (12 / 16) = 5.3e17 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-16

GE Nuclear Energy NEDO-33112 intermediate girth weld based upon a peak / lower shell location ratio of 0.81 for EPU conditions; hence the peak ID fluence used for these components is 4.29e17 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 I plant specific fluence and vessel thickness of 6.13". The 12 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

5.3e17 n/cm 2

• exp(-0.24 * (6.13 / 4)) = 3.67e17 n/cm2 .

The 12 EFPY 1/4T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

4.29e17 n/cm 2

• exp(-0.24 * (6.13 / 4)) = 2.97e17 n/cm 2.

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 12 and 16 EFPY, respectively.

17

GE Nuclear Energy NEDO-331 12 Table 4-4: Browns Ferry Unit 1 Beltline ART Values (12 EFPY)

Lower.1temdlW Plaes and Axal Wald.

Thicknss inkwlflstbi 6.13 Ratio Pakf Loation . 1.00 12 EFPYPeak L.. Bueno 5.30E0.1 rn^or2 12 EFPYPeak 114T tuw 3.67E.17 rdcmn2 12 EFFY Pak114T tuence 367E.17 rntan2 Loww Plates and Axial Welds & Lower to LowEr4rtrrmdlat. Girlt Weld Thicrw" nskcidws. 6.13 RatioPeakf Location . 0.J1 12 EFPYPeak .0. iluwnc

• 429E-17 rVdr2 12 EFPYPeak 114T ftuec

• 2697E017 .rrtn2 12 EFPY Peak114T Rua

• 2 97El17 dnmfy'2 HEATOR ktail 114T 12 EFPY 12 EFPY 12 EFPY COMPONENT HEATt.OT %Cul %W CF RTal Fhwen a a KRa x. Ma9h Smt ART

_*F rdwm-2 *F F .F 'F PLATIES:

Lower skeft

-121.1 A0999.1 0.14 0.60 100 -20 2.97E017 22 0 t1 22 44 24 6-127-2 6586841 0.15 0.44 101 -20 2.97E017 22 0 11 22 44 24 6-1274 A1009-1 0.14 0.50 96 .10 2.97E*17 21 0 10 21 42 32 Loer~nteimedlat 6hll 6-139.19 C2884-2 0.12 0.53 62 14 3.67E+17 20 0 10 20 40 54 6-139-20 C2868-2 0.09 0.46 58 30 3.67E17 14 0 7 14 29 59 6-139-21 C2753-1 0.06 0.50 51 2 3.67E+17 13 0 6 13 25 27 WELDS:

AxIWeldb E5W _ 024 0.37 141 23.1 3.67E.17 35 13 17 43 76 101 WF154 40a44 027 0.60 164 20 2.970E17 40 10 20 45 65 105

• Cwistbwsobtined flromIIIad 1131-18

GE Nuclear Energy NEDO-331 12 Table 4-5: Browns Ferry Unit 1 Beltline ART Values (16 EFPY)

L r4rlerrmdlat. Plates end Axil Welds Thd'kness InInces &13 Rato P"kJ Locabon

• 1.00 16EFPY Peak I.. Suenos

• 7.06E17 rftcne2 16EFPY Pak 1 T uenc 4896E-17 rncrn2 4

16 EFPYPeak14T fwui

• 4890E-17 rgcr 2 Lverw Plates and Axial Welds & Loer to Lower4fntamedlate Girth Weld Thevse in Woc*es.6.13 RxiooPekl Locaion

• 0.81 16 EFPYPeek I.D.Ijncea

• 5.72E.17 rdonr2 1e EFPYPeak 1t4T due 3.96E.17 rdlon2 1e EFPYPe" 1/4T fluence= W*60E17 rr/an2 HEAT OR Ina 1/4T 16 EFPY 16 EFPY 16 EFPY COMPONENT HEATtOT %ctt CF RTndl Flue A RTrdt ,

• Shif ART

__ _ __ _ _F rvr"2 F _F_ .F F PLATES:

Lower Shell 6.127-1 A099`1 0.14 0o60 00 .20 3.960417 26 0 13 26 51 31 6.127-2 85884.1 0.15 O." 101 -20 3.96E017 26 0 13 26 52 32 6-127.4 A100i 1 0.14 0.50 96 .10 3.96E017 25 0 12 25 49 39 Lowe meedlae Shell 6-139.19 C2684-2 0.12 0.53 82 14 4.896E17 24 0 12 24 47 e1 6-139-20 C268-2 0.09 0e48 68 30 4.09E017 17 0 6 17 34 64 6e139-21 C2753-1 0.06 0.50 51 2 4890E.17 15 0 7 15 29 31 WELDS:

Axla Welds ESW - 024 0.37 141 23.1 4.690417 41 13 20 48 69 112 rlth WF154 406L44 027 0o60 184 20 39M6017 47 10 24 51 99 119 Cheies obcn rod fromI1anld 113J.

19

GE Nuclear Energy NEDO-3311 2 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) IOCFR50 Appendix G [8] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions to which a pressure-retaining component may be subjected over its service lifetime. The ASME Code (Appendix G of Section Xl [6]) forms the basis for the requirements of 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 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 GE Nuclear Energy NEDO-3311 2 are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 15*F/hr or less must be maintained at all times.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 314T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 314T 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, Kr at 114T 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 10CFR50 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 NEDO-3311 2 Table 4-6: Summary of the 10CFR50 Appendix G Requirements Operating Condition and Pressure Minimum Tempera ture Requirement I. Hydrostatic Pressure Test & Leak Test (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 + 600 F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNoT + 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 + 600 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 + 400 F or of a.1 pressure, with the water level within the normal range for power operation

2. At > 20% of preservice hydrotest Larger of ASME Limits + 400F or of pressure a.2 + 40'F or the minimum permissible temperature for the inservice system hydrostatic pressure test 0

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

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

requirements. The non-beltline and beltline region operating limits are evaluated according to procedures in 10CFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

GE Nuclear Energy NEDO-331 12

))

4.3.2 P-T Curve Methodology 4.3.2.1 Non-BeItline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.Oe17 n/cm 2) 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 BWR/6 stress analysis bounds for BWRl2 through BWR/5 designs, as will be demonstrated in the following evaluation. The analyses took into account mechanical loading and anticipated thermal transients. Transients considered include all normal and upset transients such as 1000 F/hr start-up and shutdown, SCRAM, and loss of feedwater heaters. 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 BWR/6 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 NEDO-3311 2 Table 4-7: Applicable BWRI4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification FW Nozzle 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 Discontinulty.Identification CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Drain Nozzle Shell**

Support Skirt*

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 inches). The analysis is considered appropriate for Browns Ferry Unit 1 as the plant specific geometric values are bounded by the generic GE Nuclear Energy NEDO-3311 2 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 1 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.

[ 1 4.3.2. 1.1 Pressure Test - Non-B eltlne, Curve A (Using Bottom Head)

In a (( )) finite element analysis (( )), the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K1.

The (( )) generic evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section XI Appendix G [6] and shown below. The results of that computation were K, = 143.6 ksi-in'" for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 840F. ((

The limit for the coolant temperature change rate is 150 Fihr or less.

GE Nuclear Energy NEDO-33112 1[]

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

was based on a thickness of 8.0 inches; hence, t"i = 2.83. The resulting value obtained was:

Mm = 1.85 for tR<2 Mm = 0.926 ..t- for 2<4i <3.464 = 2.6206 Mm = 3.21 for f >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 *cypm = (( 1] ksi-in"n Kib = (2/3) Mm - pb = (( )) ksi-in" The total K, is therefore:

K4 = 1.5 (Kim+ Kib) + Mm * (asm + (2/3)

• Csb) = 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 Kais based on the Kj, equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) = In ((K4 - 33.2) / 20.734] / 0.02 GE Nuclear Energy NEDO-33112 (T - RTNDT) = In [(144 - 33.2) / 20.734] /0.02 (T - RTNDT) = 840 F The generic curve was generated by scaling 143.6 ksi-in"7 by the nominal pressures and calculating the associated (T- RTNDT):

Pressure Test CRD Penetration K, and (T - RTNDT) as a Function Of Pressure Nominal Pressure K, T - RTNDT (psig) (ksi-in"') (0F) 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 560F, as shown in Tables 4-1 and 4-2. ((

GE Nuclear Energy NEDO-3311 2

))

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:

Kq cc a (ra) 1 7 (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t14. Thus, K, is proportional to R/(t)' 2. The generic curve value of R/(t)"2, based on the generic BWR/6 bottom head dimensions, is:

Generic: R / (t) 112 = 138/ (8) I2 = 49 inch' 2 (4-2)

The Browns Ferry Unit 1 specific bottom head dimensions are R = 125.7 inches and t =8 inches minimum [19], resulting in:

Browns Ferry Unit 1 specific: R / (t) 1I2 = 125.7 / (8)"n = 44 inch"1 (4-3)

Since the generic value of R/(t) 12 is larger, the generic P-T curve is conservative when applied to the Browns Ferry Unit I bottom head.

GE Nuclear Energy NEDO-33112 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 K,c 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 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 102'F GE Nuclear Energy NEDO-33112 The generic curve was generated by scaling 192 ksi-inla 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 Pressure K, T - RTNDT (psig) (ksi-in'2) (OF) 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 560F, 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 NEDO-3311 2

((

4.3.2.1.3 Pressure Test - Non-Beltline 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. ((

GE Nuclear Energy NEDO-33112

)) 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 XI). 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: c = 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 I= ( 2+ t 2)12 =2.36 inches t, = thickness of nozzle = 7.125 inches t, = thickness of vessel = 6.1875 inches rn = apparent radius of nozzle = r1 + 0.29 rj=7.09 inches ri = actual inner radius of nozzle = 6.0 inches rc = nozzle radius (nozzle corner radius) = 3.75 inches Thus, alrn = 2.36 / 7.09 = 0.33. The value F(a/rQ), 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, K,, is 1.5 a (na) "2 *F(a/rn):

Nominal K, = 1.5 - 34.97 - (.

• 2.36) 12 - 1.4 = 200 ksi-in 1 2r 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] for the pressure test condition:

(T - RTNDT) = In [(K, - 33.2) / 20.734 / 0.02 GE Nuclear Energy NEDO-3311 2 (T - RTNDT) = In [(200 - 33.2) / 20.734]1/0.02 (T - RTNDT) = 104.20F 111 1]

The generic pressure test P-T curve was generated by scaling 200 ksi-in"2 by the nominal pressures and calculating the associated (T - RTNDT), [f 1*

9 1 I I The highest RTNDT for the feedwater nozzle materials is 40 0F as shown in Table 4-2.

However, the RTNDT was increased to 51°F to consider the stresses in the bottom head lower torus together with the initial RTNDT as described below. The generic pressure test P-T curve is applied to the Browns Ferry Unit I feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 51 IF.

GE Nuclear Energy NEDO-3311 2 11 1]

Second, the P-T curve is dependent on the K, value calculated. The Browns Ferry Unit 1 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and Kaare shown below:

Vessel Radius, R, 125.7 inches Vessel Thickness, t, 6.125 inches Vessel Pressure, Pv 1563 psig GE Nuclear Energy NEDO-33112 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 a =

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

a = %/4( tn 2 + ty 2 )112 =2.32 inches t, = thickness of nozzle = 6.96 inches t, = thickness of vessel = 6.125 inches rn = apparent radius of nozzle = r, + 0.29 rC=6.9 inches r, = actual inner radius of nozzle = 6.0 inches rc = nozzle radius (nozzle corner radius) = 3.0 inches Thus, a/r, = 2.32 / 6.96 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an alrn of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (7ra) 12 - F(a/rn):

Nominal K, = 1.5 -35.04 - (s 2.32) '2-1.4 = 198.7 ksi-in"2

((

1))

4.3.2.1.4 Core Not Critical Heatup/Cooldown - Non-Beltline 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 (( )) 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

- 35.-

GE Nuclear Energy NEDO-3311 2 these was normal operation with cold40'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:

Kip = SF -a (7a)/2

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

GE Nuclear Energy NEDO-3311 2

((I 1]

Finite element analysis of a nozzle corner flaw was performed to determine appropriate values of F(aIrn) 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 (( )) design stress reports for the feedwater nozzle. The stresses considered are primary membrane, apm, and primary bending, apb. Secondary membrane, asm, and secondary bending, Csb, stresses are included in the total K, by using ASME Appendix G [6] methods for secondary portion, Kl,:

Kis = Mm (csm + (2/3) *asb) (4-5)

- 37.-

GE Nuclear Energy NEDO-33112 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. Kp and K1, are added to obtain the total value of stress intensity factor, Ki. 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 - RTNOT) = In [(Ki - 33.2) / 20.7341 / 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 ((

feedwater nozzle (( )) analysis, where feedwater injection of 400F 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 (<San) 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:

Gpm = 24.84 ksi csm = 16.19 ksi ay, = 45.0 ksi t= 6.1875 inches Gpb = 0.22 ksi asb = 19.04 ksi a = 2.36 inches rn = 7.09 inches tn = 7.125 inches GE Nuclear Energy NEDO-33112 In this case the total stress, 60.29 ksi, exceeds the yield stress, CT,, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 [15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the inside surface temperature is used.)

R = [ay, - apm + ((totalt - oys) 1 30)]1 (atota, - apm) (4-7)

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

<pm = 24.84 ksi asm = 9.44 ksi apb = 0.13ksi asb =11.I0ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

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

Mm = 1.85 for It<2 Mm = 0.926 rti for 2< fi <3.464 = 2.845 Mm = 3.21 for t>3.464 The value F(a/rQ), taken from Figure A5-1 of WRC Bulletin 175 for an alr, of 0.33, is therefore, F(a/r,) =1.4 K1p is calculated from Equation 4-4:

Kip = 2.0 *(24.84 + 0.13) * (n -2.36)" *1.4 Kip = 190.4 ksi-in'I K1, is calculated from Equation 4-5:

GE Nuclear Energy NEDO-33112 KS = 2.845 * (9.44 + 2/3 *11.10)

Kl, = 47.9 ksi-in'2 The total K, is, therefore, 238.3 ksi-in't.

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'F The (( )) 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 0F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (T 5 rai -

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 K,* (T - RTNDT)

(psig) (F) (ksi-in') (OF) 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 Ki.

GE Nuclear Energy NEDO-3311 2 The highest non-beltline RTNDT for the feedwater nozzle at Browns Ferry Unit 1 is 40'F as shown in Table 4-2. However, the RTNDT was increased to 51F to consider the stresses in the bottom head lower torus as previously discussed. The generic curve is applied to the Browns Ferry Unit 1 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 51 'F as discussed in Section 4.3.2.1.3.

1[1 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 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 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 NEDO-33112 thickness (tmin) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

am = PR /tmn (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 KIc, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

The relationship between Kic 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 = KIc = 20.734 exp[O.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, Kit, for a coolant heatup/cooldown rate, specified as 150F/hr for Browns Ferry Unit 1, 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 100°F/hr. The Kit calculation for a coolant heatup/cooldown rate of 1000F/hr is described in Section 4.3.2.2.3 below.

GE Nuclear Energy NEDO-33112 4.3.2.2.2 Calculations for the Beltline Region - Pressure Test This sample calculation is for a pressure test pressure of 1064 psig at 16 EFPY. The following inputs were used in the beltline limit calculation:

Adjusted RTNDT = Initial RTNDT + Shift A = 23 + 89 = 1120F (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 [6]

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

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

Mm = 1.85 for .tj< 2 Mm = 0.926 ft- for 2<1it3.464 = 2.29 Mm = 3.21 for 1t->3.464 GE Nuclear Energy NEDO-33112 The stress intensity factor for the pressure stress is Kim = Mm

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

Equation 4-9 can be rearranged, and 1.5 Klm 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 K1t= 1.71 for a 15'F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

(T - RTNDT) = ln[(1.5

• Kim + KI, - 33.2) / 20.734] / 0.02 (4-12)

= ln[(1.5

• 51.1 + 1.71 - 33.2) / 20.734] /0.02

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

T = 38.9 + 112 = 150.90F for P = 1064 psig at 16 EFPY For Browns Ferry Unit 1, the beltline girth weld is the limiting material for both 12 and 16 EFPY. However, because the calculated value of Kim is reduced for a girth weld due to implementation of Code Case N-588 (circumferentially oriented defect for circumferential welds), the axial weld bounds the P-T curve beltline region requirements. To demonstrate that by using Code Case N-588, the axial weld has the most limiting temperature for the P-T curves in the beltline region, the stress intensity calculations for both the axial and girth welds at 16 EFPY are presented.

GE Nuclear Energy NEDO-33112 Axial Weld Calculation:

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, tin = 2.48. The resulting value obtained was:

Mm= 1.85 for ft' <2 Mm = 0.926 Hi for 2<,t<3.464 = 2.29 Mm = 3.21 for t-1

>3.464 The stress intensity factor for the pressure stress is Kim = Mm

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

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

Using the K1c equation of Paragraph A-4200 in ASME Appendix A [17], KIm = 51.1, and KIt= 1.71 for a 150F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

(T - RTNDT) = ln[(1.5 - Kim + Kit- 33.2) / 20.734] /0.02 (4-12)

= In[(1.5

• 51.1 + 1.71 - 33.2) / 20.734] / 0.02

= 38.9 0F T can be calculated by adding the adjusted RTNDT:

T = 38.9 + 112 = 150.90F for P = 1064 psig at 16 EFPY GE Nuclear Energy NEDO-33112 Girth Weld Calculation:

The value of Mm for an inside circumferential postulated surface flaw from Paragraph G-2214.1 [6] was based on a thickness of 6.13 inches (the minimum thickness without cladding); hence, t'i = 2.48. The resulting value obtained was:

Mm = 0.89 for t _2 Mm = 0.443 -I for 2< Jic3.464 = 1.10 Mm = 1.53 for 4 >3.464 The stress intensity factor for the pressure stress is Klm = Mm -a. The stress intensity factor for the thermal stress, K,1, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 150F/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 K4, equation of Paragraph A-4200 in ASME Appendix A [17], Klm = 24.5, and Klt= 1.71 for a 150F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

(T - RTNDT) = ln[(1.5

• K4m + K1 t- 33.2) / 20.734] / 0.02 (4-12)

= In[(1.5 - 24.5 + 1.71 - 33.2) / 20.734] /0.02

= -68.6°F T can be calculated by adding the adjusted RTNDT:

T = -68.6 + 119 = 50.40 F for P = 1064 psig at 16 EFPY As stated above, based on the applied pressure and temperature stress intensity factors, the axial weld flaw bounds the P-T curve in the beltline region for 16 EFPY.

GE Nuclear Energy NEDO-331 12 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]:

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

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

The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient Mt from Figure G-2214-1 of ASME Appendix G [6] by the through-wall temperature gradient ATw, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G [6]. The relationship used to compute the through-wall AT, is based on one-dimensional heat conduction through an insulated flat plate:

a 2T(x,t) / a x2 = 1 / J (fT(x,t) / At) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and P is the thermal diffusivity.

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

GE Nuclear Energy NEDO-331 12

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

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

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

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

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

The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G [6]. Therefore, AT, calculated from Equation 4-15 is used with the appropriate Mt of Figure G-2214-1 of ASME Appendix G [6] to compute K12 for heatup and cooldown.

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

Bulletin 175 [15] for infinitely long cracks of 1/4T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.

4.3.2.2.4 Calculationsfor the Beltline Region Core Not Critical Heatup/Cooldown This Browns Ferry Unit 1 sample calculation is for a pressure of 1064 psig for 16 EFPY.

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

GE Nuclear Energy NEDO-33112 Coolant heatup/cooldown rate, normally 1000 F/hr G = 100 'F/hr Minimum vessel thickness, including clad thickness C = 0.526 ft (6.125" + 0.188" = 6.313")

Thermal diffusivity at 5500 F (most conservative value) p = 0.354 ft2/ 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 /2f3 (4-16)

= 100 - (0.526)21 (2 0.354) = 39 0 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, K1t = Ml

• AT = 11.39, can be calculated. The conservative value for thermal diffusivity at 550OF is used for all calculations; therefore, Kt is constant for all pressures. 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 Klm + Klt) - 33.2)/ 20.734] /0.02 (4-17)

= lnl(2 51.1 + 11.39-33.2)/20.734]/0.02

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

T = 67.8 + 112= 179.8 0F for P = 1064 psig at 16 EFPY GE Nuclear Energy NEDO-33112 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 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 114T 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 + 900F (the largest T-RTNDT for the flange at 1563 psig is 730F). For pressures below 312 psig, the flange curve is bounded by RTNDT + 60 (the largest T - RTNDT for the flange at 312 psig is 54OF); 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 1 at low pressures.

The approach used for Browns Ferry Unit 1 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 600F adder, and 2) inclusion of the additional 60OF requirement above the RTNDT provides the additional assurance that a 1/4T 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 23.1 OF, and the LST of the closure studs is 700F; therefore, the bolt-up temperature value used is the more conservative value of 83OF. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

GE Nuclear Energy NEDO-3311 2 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 + 90'F) and Curve B temperature no less than (RTNDT + 120F).

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 68 0F for the reason discussed below.

The shutdown margin, provided in the Browns Ferry Unit 1 Technical Specification [5], is calculated for a water temperature of 680 F. 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 830F limit for the upper vessel and beltline region and the 680F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel. When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

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 400F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 400 F for pressures above 312 psig.

GE Nuclear Energy NEDO-3311 2 Table 1 of 10CFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60'F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 830F, based on an RTNDT of 23.10 F. 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.

GE Nuclear Energy NEDO-3311 2

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 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]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 150 F/hr or less must be maintained at all times.

The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, K1r, at 114T to be less than that at 314T for a given metal temperature.

GE Nuclear Energy NEDO-3311 2 The following P-T curves were generated for Browns Ferry Unit 1:

• Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 12 and 16 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 12 and 16 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 12 and 16 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 (16 EFPY), for clarity and convenience of Browns Ferry Unit I personnel, two (2) sets of these curves are provided, each with a designation of EFPY (either 12 or 16) within the title. It should be understood that this designation of EFPY in non-beltline curves does not imply limitations with regard to EFPY.

The P-T curves are beltline limited above 720 psig for both Curve A and Curve B at 16 EFPY. At 12 EFPY, the P-T curves become beltline limited above 920 psig for Curve A and above 970 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 NEDO-33112 Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves

.'>'s'- - by 'c>,'-; F igu eb-flTablet?;-:"--

6,lt,-J.'

Cure 2- -

CuneDesc'iption4.  :,Numbers:f&  ; 6ie'sffor 1N

, t - , , , e i. Presentation

.1. Lft P 1 of the P-T' 12 EFPY Curves A Bottom Head Limits (CRD Nozzle) - 12 EFPY Figure 5-1 Table B-1 A Upper Vessel Limits (FW Nozzle) - 12 EFPY Figure 5-2 Table B-1 A Beltline Limits - 12 EFPY Figure 5-3 Table B-1 A Bottom Head and Composite Curve A - 12 EFPY* Figure 5-4 Table B-2 B Bottom Head Limits (CRD Nozzle) - 12 EFPY Figure 5-5 Table B-1 B Upper Vessel Limits (FW Nozzle) - 12 EFPY Figure 5-6 Table B-1 B Beltline Limits - 12 EFPY Figure 5-7 Table B-1 B Bottom Head and Composite Curve B- 12 EFPY* Figure 5-8 Table B-2 C Composite Curve C- 12 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 - 12 EFPY 16 EFPY Curves A Bottom Head Limits (CRD Nozzle) - 16 EFPY Figure 5-11 Table B-3 A Upper Vessel Limits (FW Nozzle) - 16 EFPY Figure 5-12 Table B-3 A Beltline Limits - 16 EFPY Figure 5-13 Table B-3 A Bottom Head and Composite Curve A- 16 EFPY* Figure 5-14 Table B-4 B Bottom Head Limits (CRD Nozzle) - 16 EFPY Figure 5-15 Table B-3 B Upper Vessel Limits (FW Nozzle) - 16 EFPY Figure 5-16 Table B-3 B Beltline Limits - 16 EFPY Figure 5-17 Table B-3 B Bottom Head and Composite Curve B-16 EFPY* Figure 5-18 Table B-4 C Composite Curve C- 16 EFPY" Figure 5-19 Table B4 B&C Composite Curve C** and Curve B*with Bottom Figure 5-20 Tables B-3 & 4 Head Curve -16 EFPY I.I

• 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 Limits (FW Nozzle), and Beltline Limits.

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUE IS 156'F FOR BOTTOM HEADI 1300 1200 HEATUP/COOLDOWN 1100 RATE OF COOLANT

< 15FIHR 0

z CL1000 a

D. 900 o

I-w O 700 ACCEPTABLE AREA OF OPERATION TO 0uJ50 w 0 THE RIGHT OF THIS 2600 CURVE z

o300 w

g2400 0.

300

=Soo 200 100 0

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

Figure 5-1: Bottom Head P-T Curve for Pressure Test [Curve A] - 12 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUE IS 51 *F FOR UPPER VESSEL 1300 HEATUP/COOLDOWN 1200 RATE OF COOLANT c 151F/HR 1100 ca C, 1000 0

w LU z

c0 900 0

I.-

tat w

w "3 800 ACCEPTABLE AREA OF o 700 OPERATION TO THE RIGHT OF THIS CURVE a 600 2

LU us 2400 w

cc 300 200 100 0

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 A] - 12 EFPY

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

GE Nuclear Energy NEDO-33112 1400 INITIAL RTndt VALUE IS 1300 23.1*F FOR BELTLINE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

1100 EEPY SHIFT (F) 12 78 D 1000 0

U.1 HEATUP/COOLDOWN IL 900 RATE OF COOLANT 0

I- < 15'F/HR oi co 800 Cn w

o 700 Lu ACCEPTABLE AREA OF w 600 OPERATION TO THE RIGHT OF THIS CURVE 2

7-v5 0: 400 0~

3L 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 12 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUES ARE 23.1VF FOR BELTLINE, 1300 51'F FOR UPPER VESSEL, AND 56F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (F)

10. 12 78 1000 HEATUP/COOLDOWN CL 900 0 RATE OF COOLANT I-to < 15F/HR 0 800 U,

w so o 700 ACCEPTABLE AREA OF a 600 OPERATION TO THE z

RIGHT OF THIS CURVE

- 500 w

us A:

i,> 400 w

0.

cL 300

_-UPPER VESSEL 200 AND BELTLINE LIMITS a- . 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 12 EFPY

[15SF/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-33112 1400 INITIAL RTndt VALUE IS 56OF FOR BOTTOM HEAD 1300 1200 HEATUPICOOLDOWN RATE OF COOLANT

< 100F/HR 1100 cm

i. 1000 a

w x

& 900 0

I-i ACCEPTABLE AREA OF 0 700 OPERATION TO THE U RIGHT OF THIS CURVE I 600 3

2 u-to 500 w

UU a: 400 c

w 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] - 12 EFPY 0 F/hr or less coolant heatup/cooldown]

[1 OO GE Nuclear Energy NEDO-331 12 1400

[INITIAL RTndt VALUE IS 1300 [51OF FOR UPPER VESSEL 1200 HEATUPICOOLDOWN RATE OF COOLANT c 1007FHR 1100 Is an 1000 0

'U

a. 900 0

-j Cl 800

'U ACCEPTABLE AREA OF o 700 OPERATION TO THE RIGHT OF THIS CURVE 0

3600 z

2500 12 400 a) 300 200

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

Figure 5-6: Upper Vessel P-T Curve for Core Not Critical [Curve B] - 12 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUE IS 1300 23.10F FOR BELTLINE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (F)

Is w

1100

__ I _I_ I I_ 12 78 C 1000 HEATUP/COOLDOWN w RATE OF COOLANT

< 100FIHR

0. 900 0

-j

,,, 800 (I)

III/

o 700 I-I ACCEPTABLE AREA OF OPERATION TO THE t: 600 RIGHT OF THIS CURVE z

m 500 LU M

Xo 400 co 0: f312 ISIG 300 200 _ 1-E-I_-I

-R_ _

100

_IBL.U.

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 12 EFPY

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

GE Nuclear Energy NEDO-33112 1400 INITIAL RTndt VALUES ARE 23.1F FOR BELTLINE, 1300 51F FOR UPPER VESSEL, AND 56*F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (1F)

In 12 78 an 1000 w HEATUP/COOLDOWN IL 900 RATE OF COOLANT 0 < 100*FIHR I-Soi

-j (n 800 w

o 700 I- ACCEPTABLE AREA OF

= 600 OPERATION TO THE z RIGHT OF THIS CURVE 3 500 w

us M

VW 400 w

300 UPPER VESSEL 200 AND BELTLINE LIMITS BOTTOM HEAD 100 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 B] up to 12 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUES ARE 1300 23.1"F FOR BELTLINE, 51'F FOR UPPER VESSEL, 1200 AND 566F FOR BOTTOM HEAD 1100 c,

on BELTLINE CURVE 0* 1000 ADJUSTED AS SHOWN:

w EFPY SHIFT (F) 12 78

c. 900 I-en 800 HEATUPICOOLDOWN RATE OF COOLANT

< 100-F/HR o 700 II-w 600 Z

ACCEPTABLE AREA OF

> 500 OPERATION TO THE w RIGHT OF THIS CURVE au 400 w

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

(.F)

Figure 5-9: Composite Core Critical P-T Curves [Curve C] up to 12 EFPY

[lOOF/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-33112 B C 1400 INITIAL RTndt VALUES ARE 23.1VF FOR BELTLINE.

1300 511 FOR UPPER VESSEL, AND 56'F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (OF)

Is 0 12 78 c' 1000 0

HEATUPICOOLDOWN z RATE OF COOLANT

0. 900 0 < 100IFIHR

-J en LU 800 0 700 ACCEPTABLE AREA OF z 600 OPERATION TO THE z RIGHT OF THIS CURVE 3 500 400 1n IL, Ce 300

-COMPOSITE 200 CURVE B BOTTOM HEAD .

100 CURVE B

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

Figure 5-10: Composite Core Not Critical [Curve B] including Bottom Head and Core Critical P-T Curves [Curve C] up to 12 EFPY [100OF/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUE IS 56'F FOR BOTTOM HEAD 1300 1200 HEATUP/COOLDOWN RATE OF COOLANT

< 15FIHR 1100 0.

-* 1000 u

'U

a. 900 0

I-to ci, 600 o 700 ACCEPTABLE AREA OF OPERATION TO THE RIGHT OF THIS R

a: 600 CURVE

_500

'U en 400

'u 300 200 100 30 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 5-11: Bottom Head P-T Curve for Pressure Test [Curve A] - 16 EFPY

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

GE Nuclear Energy NEDO-331 12 1400

[INITIAL RTndt VALUE IS I51 OF FOR UPPER VESSEL 1300 HEATUPICOOLDOWN 1200 RATE OF COOLANT

< 15'F/HR 1100 0.

0 CL 1 °°° X. 900 o

P-u0 800 Co aL ACCEPTABLE AREA OF OPERATION TO THE o 700 RIGHT OF THIS CURVE w 600 Z

j 500 Lu

°) 400 Lu 3t 300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 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] - 16 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 _ _ _ _ _ _

INITIAL RTndt VALUE 1300 - - 23.1'F FOR BELTLIN 1200 - BELTLINE CURVE ADJUSTED AS SHOWt 1100 _ _ - EFPY SHIFT (F) 1000 w

x 900 - - HEATUP/COOLDOWN 0 RATE OF COOLAN 0 -/ _ 15'F/HR 800 9n o 700 - - _

ACCEPTABLE AREA CDF 60 OPERATION TO THE 600 -- - RIGHT OF THIS CURV

500 _ _- _

s 400 - _ _

300 30 312 PSIG _

200 - _ _

BOL3TU-P BELTLINE LIMI 100 0 0 0 25 S0 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-13: Beltline P-T Curve for Pressure Test [Curve A] up to 16 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUES ARE 23.1F FOR BELTLINE, 1300 51 OF FOR UPPER VESSEL, AND 56*F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS SHOWN:

EFPY SHIFT (F) to 16 89 C. 1000 HEATUPICOOLDOWN

0. 900 RATE OF COOLANT 0

S 15F/HR

-z U,

o 700 (0 800 ACCEPTABLE AREA OF OPERATION TO THE U, RIGHT OF THIS CURVE 4-00 2500 2300 UPPER VESSEL AND BELTLINE LIMITS

--- 9--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 16 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 INITIAL RTndt VALUE IS 1300 56'F FOR BOTTOM HEAD 1200 HEATUPICOOLDOWN RATE OF COOLANT

< 100IFIHR 1100 Is

- 1000 w

0. 900 g-0n 800 w

w o 700 ACCEPTABLE AREA OF OPERATION TO THE U, 800 RIGHT OF THIS CURVE I3-600 2

z M 400 w

3)00 200

-BOTTOM HEAD LIMITS 100 0

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

Figure 5-15: Bottom Head P-T Curve for Core Not Critical [Curve B] - 16 EFPY

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

GE Nuclear Energy NEDO-3311 2 1400

[INITIAL RTndt VALUE IS 1300 l51OF FOR UPPER VESSELl 1200 HEATUP/COOLDOWN RATE OF COOLANT

< 100F/HR 1100 to

1000 w

IL 900 0

II-J co 800 co ACCEPTABLE AREA OF o 700 OPERATION TO THE RIGHT OF THIS CURVE I-I W 600 2

Ei 500 a) 400 Lu IS 300 200

-UPPER VESSEL LIMITS (Including Flange and FW 100 Nozzle Limits) 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 B] - 16 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 l INITIAL RTndt VALUE IS 1300 23.1 F FOR BELTLINE 1200 BELTLINE CURVE ADJUSTED AS SHOWN:

EEPY SHIFT OFF) 1100 16 89 an C 1000 HEATUP/COOLDOWN LU RATE OF COOLANT x < 100OF/HR

0. 900 o

I-l Xo 800 o 700 I-I ACCEPTABLE AREA OF OPERATION TO THE w 600 RIGHT OF THIS CURVE Z

I= 500 so 400 co 3L 300 200 100 0

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

Figure 5-17: Beltline P-T Curve for Core Not Critical [Curve BI up to 16 EFPY

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

GE Nuclear Energy NEDO-331 12 1400 _ _ _ _ _ _ _

INITIAL RTndt VALUES ARI 1300 --

1300

- -51'F

__ l tAND 23.1'F FOR BELTLINE, FOR UPPER VESSEL 1200 - - 56'F FOR BOTTOM HEAD 1100 _ - _ __l_

[ BELTLINE CURVES ADJUSTED AS SHOWN:

,0oo EFPY SHIFT (F) 16 89

~.900i

-z n 800<

X  :

/

lRATE HEATUP/COOLDOWN OF COOLANT 100-F/HR ll 0 700 ,.' -

600 60 _ACCEPTABLE AREA OF

_ _OPERATION TO THE

> 500 -4so -s toRIGHT - - OF THISCURVE w I 400 - - _- - - - _- A 0I -BOTTOM HEA I. ID /

300 68F 200 - - -I UPPERVESSELANI~

FLANGEBELTLINE LIMITS 1003F ...-. BOTTOM HEAD CUF 0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-18: Composite Core Not Critical P-T Curves [Curve B] up to 16 EFPY

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

GE Nuclear Energy NEDO-33112 1400 INITIAL RTndt VALUES ARE 1300 23.10F FOR BELTLINE.

51OF FOR UPPER VESSEL, 1200 AND 56F FOR BOTTOM HEAD 1100

-0 0.

D. 1000 BELTLINE CURVE ADJUSTED AS SHOWN:

<L EFPY SHIFT ('F) 0C 900 16 89 0

I I - - PI G ti I I I XtI IX I HEATUP/COOLDOWN RATE OF COOLANT

< 100F/HR o 700 w

a: 600 Miniu CnIclt I I Z

1:

M ACCEPTABLE AREA OF w

i 500 Tompratre,33F OPERATION TO THE RIGHT OF THIS CURVE w

400

_g_S __ _ _

I 3

300 200 I I>. 1

-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

('F)

Figure 5-19: Composite Core Critical P-T Curves [Curve C] up to 16 EFPY (IO000Fhr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-3311 2 B C 1400 INITIAL RTndt VALUES ARE 23.1VF FOR BELTLINE, 1300 51VF FOR UPPER VESSEL, AND 56*F FOR BOTTOM HEAD 1200 BELTLINE CURVES 1100 ADJUSTED AS Is on SHOWN:

0. FM SHIFT OEF)

- 1000 16 89 C

0. 900 0 HEATUP/COOLDOWN I.-

RATE OF COOLANT U' 800 c 100F/HR U'

Uj ACCEPTABLE AREA OF 3600 OPERATION TO THE z RIGHT OF THIS CURVE 0 700 2

Lu 500

3 LU CO 400 IL 300

- COMPOSITE 200 CURVE B 100 I I I

.... Ad BOTTOM HEAD CURVE B COMPOSITE I 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 16 EFPY [100F/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-33112

6.0 REFERENCES

1. GW Contreras, 'Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3: Pressure Temperature Operating Limits", GENE, San Jose, CA, June 1994 (GE-NE-523-A65-0594, DRF 137-0010-7, Revision 1).

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

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

4. a) "Alternative Reference Fracture Toughness for Development of P-T Limit Curves Section Xl, Division 1," Code Case N-640 of the ASME Boiler & Pressure Vessel Code, Approval Date February 26, 1999.

b) "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels Section Xl, Division 1", Code Case N-588 of the ASME Boiler & Pressure Vessel Code, Approval Date December 12, 1997.

5. Technical Specifications For Browns Ferry Nuclear Plant, Unit 1.

6. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section III or Xl of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

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.

GE Nuclear Energy NEDO-33112

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).

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 1 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. a) "Evaluation of RTNDT, USE, and Chemical Composition of Core Region Electroslag Welds for Quad Cities Units 1 and 2", Framatome Technologies, Lynchburg, Virginia, January 1996 (BAW-2259).

b) Letter, TE Abney (TVA) to US NRC, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - Generic Letter (GL) 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity - Response to NRC Request for Additional Information (TAC Nos.

MA1179, MAI180, and MA181), September 8, 1998.

14. 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. ((

3))

17. 'Analysis of Flaws", Appendix A to Section Xl of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

GE Nuclear Energy NEDO-33112

18. ((

19. Bottom Head and Feedwater Nozzle Dimensions:

a. Babcock & Wilcox Company Drawing 122859E, Revision 10, 'Lower 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, 1995 Edition with Addenda through 1996.

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

GE Nuclear Energy NEDO-331 12 APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy NEDO-33112 1[

Il I I I I .I l l I l l l 0I VI

_-I i 1 II

__ _ _ _ _ _ _ _ _ I __ _ I _ _ _ _

1]

A-2

GE Nuclear Energy NEDO-3311 2 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 60°F. Also Inconel discontinuities require no fracture toughness evaluations.

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

Appendix E).

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

_ _ evaluation.

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

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

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.

MK131 - 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.

MK1 33 - 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 SA 82 F304 1,21, 22 Appurtenances made from Stainless Flange) Steel require no fracture toughness evaluation.

MK1 35 - 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 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.

' 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 NEDO-33112 APPENDIX A

REFERENCES:

1. Vessel Drawings and Materials:

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

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

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

. Drawing #122855E, Revision 14, '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. J. Valente (TVA) to Dale Porter (GE), 'Browns Ferry Nuclear Plant (BFN) -

Pressure-Temperature Curves Design Input Request (DIR) - Transmittal of DIR Rev. 0", July 17, 2003 (TVA RIMS No. W83 030717 001).

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-1 1 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).

A-4

GE Nuclear Energy NEDO-33112

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

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. Control Rod Nozzles:

• Drawing #122883E, Revision 5, 'Control Rod Nozzles, Unit #1", Babcock &

Wilcox Company, Mt. Vernon, Indiana (GE VPF #1805-065).

• 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 #1805-1 08).

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

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

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

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

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

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

24. Drawing #122863E, Revision 6, "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).

A-5

GE Nuclear Energy NEDO-3311 2

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

A-6

GE Nuclear Energy NEDO-33112 APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy NEDO-33112 TABLE B-1. Browns Ferry Unit I P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 0F/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 BOTTOM UPPER-- 2 EFPY -- BOTTOM:-, UPPER v 12 EFPY HEAD'. VESSEL BELTLINE HEAD CURVE:- VESSEL.U to; BELTLINE PRESSURE CURVE A CURVE A- CURVE A - -' B CURVEBusCU EB (0F)

(PSIG); (FF) -  : (F) -. (F) (F) Ax

. (0F) 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 85.2 83.0 140 68.0 83.0 83.0 68.0 88.4 83.0 150 68.0 83.0 83.0 68.0 91.2 83.0 160 68.0 83.0 83.0 68.0 93.9 83.0 170 68.0 83.0 83.0 68.0 96.5 83.0 180 68.0 83.0 83.0 68.0 98.9 83.0 190 68.0 83.0 83.0 68.0 101.2 83.0 200 68.0 83.0 83.0 68.0 103.3 83.0 210 68.0 83.0 83.0 68.0 105.3 83.0 220 68.0 83.0 83.0 68.0 107.3 83.0 B-2

GE Nuclear Energy NEDO-3311 2 TABLE B-1. Browns Ferry Unit 1 P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 15 0F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7

. BOTTOM>.. UPPER' -12EFPY%-.:-BOTTOM.-- UPPER, 12 EFPY HEAD VESSEL B LiLINE HEAD CURVE :-' VESSEL- BELTLINE PRESSURE

• CURVEA'j CU RVE Ar CURVEA:. - -CURVE:B -: CURVEB (PSIG) (F) (OF),: -:(F) (OF)>-°-

230 68.0 83.0 83.0 68.0 109.1 83.0 240 68.0 83.0 83.0 68.0 110.9 83.0 250 68.0 83.0 83.0 68.0 112.6 83.0 260 68.0 83.0 83.0 68.0 114.2 83.0 270 68.0 83.0 83.0 68.0 115.8 83.0 280 68.0 83.0 83.0 68.0 117.3 83.0 290 68.0 83.0 83.0 68.0 118.8 83.0 300 68.0 83.0 83.0 68.0 120.2 83.0 310 68.0 83.0 83.0 68.0 121.5 83.0 312.5 68.0 83.0 83.0 68.0 121.9 83.0 312.5 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 B-3

GE Nuclear Energy NEDO-331 12 TABLE B-1. Browns Ferry Unit 1 P-T Curve Values for 12 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

. 'BO1TOM`- UPPERk>'!,-`12 EFPY- a- ETOMrlZ UAPR'VK12

! EFPY -

HEAD. VESSEL. - BELT-NEs HEADICURVE* VESSEL -.BELTLINE PRESSUREi: -,-:-EA. C,CyEA CURVE .i- RB SIG)", -(OF 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 70.5 143.0 143.0 490 68.0 113.0 113.0 72.8 143.0 143.0 500 68.0 113.0 113.0 75.0 143.0 143.0 510 68.0 113.0 113.0 77.2 143.0 143.0 520 68.0 113.0 113.0 79.2 143.2 143.0 530 68.0 113.0 113.0 81.2 144.0 143.0 540 68.0 113.0 113.0 83.1 144.8 143.0 550 68.0 113.0 113.0 84.9 145.6 143.0 560 68.0 113.0 113.0 86.7 146.4 143.0 570 68.0 113.0 113.0 88.4 147.1 143.0 580 68.0 113.0 113.0 90.0 147.9 143.0 590 68.0 113.0 113.0 91.6 148.6 143.0 600 68.0 113.0 113.0 93.2 149.1 143.0 610 68.0 113.0 113.0 94.7 149.6 143.0 620 68.0 113.0 113.0 96.1 150.0 143.0 630 68.0 113.0 113.0 97.5 150.4 143.0 640 68.0 113.0 113.0 98.9 150.8 143.0 650 68.2 113.0 113.0 100.2 151.2 143.0 660 69.9 113.0 113.0 101.5 151.7 143.0 670 71.6 113.0 113.0 102.8 152.1 143.0 680 73.2 113.0 113.0 104.1 152.5 143.0 690 74.7 113.0 113.0 105.3 152.9 143.0 700 76.2 113.0 113.0 106.4 153.3 143.0 B-4

GE Nuclear Energy NEDO-3311 2 TABLE B-1. Browns Ferry Unit 1 P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 15 0F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7

, BOTTOMS'> UPPER -- 12PEFPY-. -. BOTTOM,.UPPERr::, 12 EFPY

- - .HEA: VESEL`.`:- BELTLINE:.;.HEADCURVE-VESSE. BELTLINE PRESSURES I-CURVEA CURVEA.' CURVE:AB ,URVEBC CURVElB (SIG - , F) 710 77.7 113.0 113.0 107.6 153.7 143.0 720 79.1 113.0 113.0 108.7 154.1 143.0 730 80.5 113.3 113.0 109.8 154.5 144.0 740 81.8 114.1 113.0 110.9 154.9 144.9 750 83.1 115.0 113.0 112.0 155.2 145.9 760 84.4 115.8 113.0 113.0 155.6 146.8 770 85.6 116.6 113.0 114.0 156.0 147.7 780 86.8 117.3 113.0 115.0 156.4 148.6 790 88.0 118.1 113.0 116.0 156.8 149.4 800 89.2 118.9 113.3 116.9 157.1 150.3 810 90.3 119.6 114.6 117.9 157.5 151.1 820 91.4 120.4 115.9 118.8 157.9 151.9 830 92.5 121.1 117.2 119.7 158.2 152.8 840 93.5 121.8 118.4 120.6 158.6 153.6 850 94.6 122.5 119.6 121.4 158.9 154.3 860 95.6 123.2 120.7 122.3 159.3 155.1 870 96.6 123.9 121.9 123.1 159.6 155.9 880 97.5 124.6 123.0 124.0 160.0 156.6 890 98.5 125.3 124.0 124.8 160.3 157.4 900 99.4 125.9 125.1 125.6 160.7 158.1 910 100.4 126.6 126.1 126.4 161.0 158.8 920 101.3 127.2 127.2 127.1 161.4 159.5 930 102.1 127.9 128.2 127.9 161.7 160.2 940 103.0 128.5 129.1 128.7 162.0 160.9 950 103.9 129.1 130.1 129.4 162.4 161.6 B-5

GE Nuclear Energy NEDO-33112 TABLE B-1. Browns Ferry Unit 1 P-T Curve Values for 12 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

',BOTOOMm 12 EPPER - PUPPERN - 12 EFPY

, of, -~'*-. HEAD J

.O7iVESSEL<b-;BELTULNE HEAD CURVE;" VESSELS: - BELTLINE PRESSURE.CURVEV ,ACURVE'B,':CURVE B I'. - (OF) 960 104.7 129.7 131.0 130.1 162.7 162.3 970 105.6 130.3 132.0 130.9 163.0 162.9 980 106.4 130.9 132.9 131.6 163.4 163.6 990 107.2 131.5 133.8 132.3 163.7 164.2 1000 108.0 132.1 134.6 133.0 164.0 164.8 1010 108.7 132.7 135.5 133.6 164.3 165.5 1020 109.5 133.2 136.3 134.3 164.6 166.1 1030 110.3 133.8 137.2 135.0 165.0 166.7 1040 111.0 134.4 138.0 135.6 165.3 167.3 1050 111.7 134.9 138.8 136.3 165.6 167.9 1060 112.4 135.5 139.6 136.9 165.9 168.5 1064 112.7 135.7 139.9 137.2 166.0 168.7 1070 113.2 136.0 140.4 137.5 166.2 169.1 1080 113.9 136.5 141.1 138.2 166.5 169.7 1090 114.6 137.1 141.9 138.8 166.8 170.2 1100 115.2 137.6 142.6 139.4 167.1 170.8 1105 115.6 137.8 143.0 139.7 167.3 171.1 1110 115.9 138.1 143.4 140.0 167.4 171.4 1120 116.6 138.6 144.1 140.6 167.7 171.9 1130 117.2 139.1 144.8 141.2 168.0 172.4 1140 117.9 139.6 145.5 141.7 168.3 173.0 1150 118.5 140.1 146.2 142.3 168.6 173.5 1160 119.1 140.6 146.9 142.9 168.9 174.0 1170 119.8 141.1 147.5 143.4 169.2 174.6 1180 120.4 141.6 148.2 144.0 169.5 175.1 B-6

GE Nuclear Energy NEDO-33112 TABLE B-1. Browns Ferry Unit 1 P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 15 "F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM -UPPER 4'- 12 EFPY,-:- BOTTOM UPPER -. 12 EFPY

, -..  :;'-,HEADj'. . - VESSEL '. .BELTLINE. HEAD CURVE; VESSEL:..: BELTLINE PRESSURE' -,-,CURVE A., i, CURVE A .' CURVE BA-; ,  : CURVE B:;.CURVEB (PSIG).* -; (CF, (F ) ."; (0F)"'- (° F);'-. (0F) 1190 121.0 142.1 148.9 144.5 169.7 175.6 1200 121.6 142.5 149.5 145.1 170.0 176.1 1210 122.2 143.0 150.2 145.6 170.3 176.6 1220 122.8 143.5 150.8 146.2 170.6 177.1 1230 123.3 143.9 151.4 146.7 170.9 177.6 1240 123.9 144.4 152.0 147.2 171.2 178.1 1250 124.5 144.8 152.6 147.7 171.4 178.6 1260 125.0 145.3 153.2 148.2 171.7 179.0 1270 125.6 145.7 153.8 148.7 172.0 179.5 1280 126.1 146.2 154.4 149.2 172.2 180.0 1290 126.7 146.6 155.0 149.7 172.5 180.5 1300 127.2 147.0 155.6 150.2 172.8 180.9 1310 127.7 147.5 156.1 150.7 173.1 181.4 1320 128.3 147.9 156.7 151.2 173.3 181.8 1330 128.8 148.3 157.2 151.6 173.6 182.3 1340 129.3 148.7 157.8 152.1 173.8 182.7 1350 129.8 149.1 158.3 152.6 174.1 183.2 1360 130.3 149.6 158.9 153.0 174.4 183.6 1370 130.8 150.0 159.4 153.5 174.6 184.0 1380 131.3 150.4 159.9 153.9 174.9 184.4 1390 131.8 150.8 160.5 154.4 175.1 184.9 1400 132.3 151.2 161.0 154.8 175.4 185.3 B-7

GE Nuclear Energy NEDO-33112 TABLE B-2. Browns Ferry Unit 1 Composite P-T Curve Values for 12 EFPY 0

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 and 5-10

-. "BOTTOM' AUPPER RPV 8J&;BOTTOM 'JUPPER.RPV:&,URPER RPV &'-

i -;."HEAD, ELINEAT HE BELTLlNEATBELTUNEATr

."12 EPY~,EFPY--

PRESSURE.t"CURVE'A, "

';CURVE A 7. CURVE B3 CURVE -,CRV P (PSiG) s F): (oj .(F - '.(n - ) - ,

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 91.0 70 68.0 83.0 68.0 83.0 98.2 80 68.0 83.0 68.0 83.0 104.2 90 68.0 83.0 68.0 83.0 109.3 100 68.0 83.0 68.0 83.0 113.8 110 68.0 83.0 68.0 83.0 117.9 120 68.0 83.0 68.0 83.0 121.7 130 68.0 83.0 68.0 85.2 125.2 140 68.0 83.0 68.0 88.4 128.4 150 68.0 83.0 68.0 91.2 131.2 160 68.0 83.0 68.0 93.9 133.9 170 68.0 83.0 68.0 96.5 136.5 180 68.0 83.0 68.0 98.9 138.9 190 68.0 83.0 68.0 101.2 141.2 200 68.0 83.0 68.0 103.3 143.3 210 68.0 83.0 68.0 105.3 145.3 220 68.0 83.0 68.0 107.3 147.3 B-8

GE Nuclear Energy NEDO-33112 TABLE B-2. Browns Ferry Unit 1 Composite P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 "F/hr for Curves B & Cand 15 °F/hr for Curve A For Figures 54, 5-8, 5-9 and 5-10 BOTTOM UPPER RPV &' BOTOM-AUPPER RRV&.8 UPPER RPV9&.

HE - -BELTINE HEAD B-E ELTLNE AT IT4a'-,.a-, '. *.*12 EFY-w, . ,S PRESSUlRECURVECURVE URVEA~.~CUJRVE B-yCRVE-ElBZ-, b-~CURE C 4 (IPSIG)'--- (FT F~*h (F .O)a1.~ F 230 68.0 83.0 68.0 109.1 149.1 240 68.0 83.0 68.0 110.9 150.9 250 68.0 83.0 68.0 112.6 152.6 260 68.0 83.0 68.0 114.2 154.2 270 68.0 83.0 68.0 115.8 155.8 280 68.0 83.0 68.0 117.3 157.3 290 68.0 83.0 68.0 118.8 158.8 300 68.0 83.0 68.0 120.2 160.2 310 68.0 83.0 68.0 121.5 161.5 312.5 68.0 83.0 68.0 121.9 161.9 312.5 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 NEDO-3311 2 TABLE B-2. Browns Ferry Unit 1 Composite P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B& C and 15 0 F/hr for Curve A For Figures 5-4, 5-8, 5-9 and 5-10

• .. . BO.T;.-;-.BOTTOM-U;.sUPPER.RPV&- BOTTOMW-UPPER!RPV,&`;-UPPER RPV &

-, HEAD "'BELTINE AT- HEAD-- BELTLINEATs -<BELTLINE-AT-

' 2 FYEFPYW 2:

PRESSURE' ,CURVEA> 9.'CURVE-A '-CURVE B6.'CURVE B ,P CURVE C.

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 70.5 143.0 183.0 490 68.0 113.0 72.8 143.0 183.0 500 68.0 113.0 75.0 143.0 183.0 510 68.0 113.0 77.2 143.0 183.0 520 68.0 113.0 79.2 143.2 183.2 530 68.0 113.0 81.2 144.0 184.0 540 68.0 113.0 83.1 144.8 184.8 550 68.0 113.0 84.9 145.6 185.6 560 68.0 113.0 86.7 146.4 186.4 570 68.0 113.0 88.4 147.1 187.1 580 68.0 113.0 90.0 147.9 187.9 590 68.0 113.0 91.6 148.6 188.6 600 68.0 113.0 93.2 149.1 189.1 610 68.0 113.0 94.7 149.6 189.6 620 68.0 113.0 96.1 150.0 190.0 630 68.0 113.0 97.5 150.4 190.4 640 68.0 113.0 98.9 150.8 190.8 650 68.2 113.0 100.2 151.2 191.2 660 69.9 113.0 101.5 151.7 191.7 670 71.6 113.0 102.8 152.1 192.1 680 73.2 113.0 104.1 152.5 192.5 690 74.7 113.0 105.3 152.9 192.9 700 76.2 113.0 106.4 153.3 193.3 B-1 0

GE Nuclear Energy NEDO-3311 2 TABLE B-2. Browns Ferry Unit 1 Composite P-T Curve Values for 12 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 and 5-10 BOTTOMr- UPPER RPV,&. BOTTOM UPPER RPV & -.-UPPER RPV &

"HEAD, BBELTL INE AT'!

TLINETA E BELTLN T BET 12 EFPY-:',- 2EFPY."'

it>'_12 PRESSURE-' CCURVE C 710 77.7 113.0 107.6 153.7 193.7 720 79.1 113.0 108.7 154.1 194.1 730 80.5 113.3 109.8 154.5 194.5 740 81.8 114.1 110.9 154.9 194.9 750 83.1 115.0 112.0 155.2 195.2 760 84.4 115.8 113.0 155.6 195.6 770 85.6 116.6 114.0 156.0 196.0 780 86.8 117.3 115.0 156.4 196.4 790 88.0 118.1 116.0 156.8 196.8 800 89.2 118.9 116.9 157.1 197.1 810 90.3 119.6 117.9 157.5 197.5 820 91.4 120.4 118.8 157.9 197.9 830 92.5 121.1 119.7 158.2 198.2 840 93.5 121.8 120.6 158.6 198.6 850 94.6 122.5 121.4 158.9 198.9 860 95.6 123.2 122.3 159.3 199.3 870 96.6 123.9 123.1 159.6 199.6 880 97.5 124.6 124.0 160.0 200.0 890 98.5 125.3 124.8 160.3 200.3 900 99.4 125.9 125.6 160.7 200.7 910 100.4 126.6 126.4 161.0 201.0 920 101.3 127.2 127.1 161.4 201.4 930 102.1 128.2 127.9 161.7 201.7 940 103.0 129.1 128.7 162.0 202.0 950 103.9 130.1 129.4 162.4 202.4 B-1I1

GE Nuclear Energy NEDO-3311 2 TABLE B-2. Browns Ferry Unit 1 Composite P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 15 0F/hr for Curve A For Figures 5-4, 5-8, 5-9 and 5-10 BOTTOM-; UPPER RPRV&-- BOTT0M-?,WUPPER.RPV.&-'< UPPER RPV &

-:;-' HEAD BELTLINE-AT' 'l AHEAD' -- BELT E-AT ELTLINEAT P ~~~~-'12 EFPY!r' ~ ~ 2EP~J -A2EP PAR PRESSURE .~~CURVE'A E

..~-'- C~'~URVE A---: - '- CURVEB S~RVE B wv. UV (PSIG)TF - (0F) - (F) --  : (°F' 960 104.7 131.0 130.1 162.7 202.7 970 105.6 132.0 130.9 163.0 203.0 980 106.4 132.9 131.6 163.6 203.6 990 107.2 133.8 132.3 164.2 204.2 1000 108.0 134.6 133.0 164.8 204.8 1010 108.7 135.5 133.6 165.5 205.5 1020 109.5 136.3 134.3 166.1 206.1 1030 110.3 137.2 135.0 166.7 206.7 1040 111.0 138.0 135.6 167.3 207.3 1050 111.7 138.8 136.3 167.9 207.9 1060 112.4 139.6 136.9 168.5 208.5 1064 112.7 139.9 137.2 168.7 208.7 1070 113.2 140.4 137.5 169.1 209.1 1080 113.9 141.1 138.2 169.7 209.7 1090 114.6 141.9 138.8 170.2 210.2 1100 115.2 142.6 139.4 170.8 210.8 1105 115.6 143.0 139.7 171.1 211.1 1110 115.9 143.4 140.0 171.4 211.4 1120 116.6 144.1 140.6 171.9 211.9 1130 117.2 144.8 141.2 172.4 212.4 1140 117.9 145.5 141.7 173.0 213.0 1150 118.5 146.2 142.3 173.5 213.5 1160 119.1 146.9 142.9 174.0 214.0 1170 119.8 147.5 143.4 174.6 214.6 1180 120.4 148.2 144.0 175.1 215.1 B-12

GE Nuclear Energy NEDO-33112 TABLE B-2. Browns Ferry Unit 1 Composite P-T Curve Values for 12 EFPY Required Coolant Temperatures at 100 °FIhr for Curves B & C and 15 "F/hr for Curve A For Figures 5-4, 5-8, 5-9 and 5-10 BOTTOM UPPER RPV 8' BOTTOM UPPER RPV &; UPPER RPV &-

*.: >-.-HEAD, ,-..BELTLINEAT. .. HEAD. BELTLINE AT&'" ., BELTLINE AT'.

-. -- - S .EFP PRESSUREX -' CURVEWA -- CURVE A CURVE .iCURVE B 5Y.,, .CURVEC --

G---

IPj :'(F) ~'.S'- :( 0 ) . (@F .(F~)? ° '^)~ (by-(F2' 1190 121.0 148.9 144.5 175.6 215.6 1200 121.6 149.5 145.1 176.1 216.1 1210 122.2 150.2 145.6 176.6 216.6 1220 122.8 150.8 146.2 177.1 217.1 1230 123.3 151.4 146.7 177.6 217.6 1240 123.9 152.0 147.2 178.1 218.1 1250 124.5 152.6 147.7 178.6 218.6 1260 125.0 153.2 148.2 179.0 219.0 1270 125.6 153.8 148.7 179.5 219.5 1280 126.1 154.4 149.2 180.0 220.0 1290 126.7 155.0 149.7 180.5 220.5 1300 127.2 155.6 150.2 180.9 220.9 1310 127.7 156.1 150.7 181.4 221.4 1320 128.3 156.7 151.2 181.8 221.8 1330 128.8 157.2 151.6 182.3 222.3 1340 129.3 157.8 152.1 182.7 222.7 1350 129.8 158.3 152.6 183.2 223.2 1360 130.3 158.9 153.0 183.6 223.6 1370 130.8 159.4 153.5 184.0 224.0 1380 131.3 159.9 153.9 184.4 224.4 1390 131.8 160.5 154.4 184.9 224.9 1400 132.3 161.0 154.8 185.3 225.3 B-13

GE Nuclear Energy NEDO-33112 TABLE B-3. Browns Ferry Unit 1 P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 °Flhr 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

• BOTTOM - .UPE,,;1 EFP BOTM;kPER3:16E HEAD. VESSELf HBELTLINE PRESSURE CURVE AS ; CURVE AN, -4' CURVEA.w CURVEB '4CURVEB CURVE B (PSiG) (IFS . .'v ((0F.- =FF) -  ;

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 85.2 83.0 140 68.0 83.0 83.0 68.0 88.4 83.0 150 68.0 83.0 83.0 68.0 91.2 83.0 160 68.0 83.0 83.0 68.0 93.9 83.0 170 68.0 83.0 83.0 68.0 96.5 83.0 180 68.0 83.0 83.0 68.0 98.9 83.0 190 68.0 83.0 83.0 68.0 101.2 83.0 200 68.0 83.0 83.0 68.0 103.3 83.0 210 68.0 83.0 83.0 68.0 105.3 83.0 220 68.0 83.0 83.0 68.0 107.3 83.0 230 68.0 83.0 83.0 68.0 109.1 83.0 B-14

GE Nuclear Energy NEDO-33112 TABLE B-3. Browns Ferry Unit 1 P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 15 0F/hr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20

- BOTTOM UPPER 16 EFPY. -E BOTTOM.-- UPPER:--- .16 EFPY HEA VESLBCLN~~ED'.~VSSEV'-~'-bELTLINE PRESSURE-~,' CURE URV V ? ~~CRVE .B :.,1~CURVE.BW-~- CURVE B 2406803.- 0 6 1 240 68.0 83.0 83.0 68.0 110.9 83.0 250 68.0 83.0 83.0 68.0 112.6 83.0 260 68.0 83.0 83.0 68.0 114.2 83.0 270 68.0 83.0 83.0 68.0 115.8 83.0 280 68.0 83.0 83.0 68.0 117.3 83.0 290 68.0 83.0 83.0 68.0 118.8 83.0 300 68.0 83.0 83.0 68.0 120.2 83.0 310 68.0 83.0 83.0 68.0 121.5 83.0 312.5 68.0 83.0 83.0 68.0 121.9 83.0 312.5 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-15

GE Nuclear Energy NEDO-33112 TABLE B-3. Browns Ferry Unit 1 P-T Curve Values for 16 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 BOTTOM. UPPERI§ 16EFPY. BOTTOM%.-r.UPPER 16EFPY 16..

'7 EA~~ ~.VSSL~ 2 ELLIEK

- YHEADWZ VESE~; LTLINE PRS EC'URVEA" .CRE0 AZ. S_

CURVEALCURVEBiJ.CUR v~,.$; 'AE CURVE B 480 68.0 113.0 113.0 70.5 143.0 143.0 490 68.0 113.0 113.0 72.8 143.0 143.0 500 68.0 113.0 113.0 75.0 143.0 143.0 510 68.0 113.0 113.0 77.2 143.0 143.0 520 68.0 113.0 113.0 79.2 143.2 143.0 530 68.0 113.0 113.0 81.2 144.0 143.0 540 68.0 113.0 113.0 83.1 144.8 143.0 550 68.0 113.0 113.0 84.9 145.6 143.0 560 68.0 113.0 113.0 86.7 146.4 143.0 570 68.0 113.0 113.0 88.4 147.1 143.0 580 68.0 113.0 113.0 90.0 147.9 143.0 590 68.0 113.0 113.0 91.6 148.6 143.0 600 68.0 113.0 113.0 93.2 149.1 143.0 610 68.0 113.0 113.0 94.7 149.6 143.0 620 68.0 113.0 113.0 96.1 150.0 143.1 630 68.0 113.0 113.0 97.5 150.4 144.3 640 68.0 113.0 113.0 98.9 150.8 145.5 650 68.2 113.0 113.0 100.2 151.2 146.7 660 69.9 113.0 113.0 101.5 151.7 147.8 670 71.6 113.0 113.0 102.8 152.1 148.9 680 73.2 113.0 113.0 104.1 152.5 149.9 690 74.7 113.0 113.0 105.3 152.9 151.0 700 76.2 113.0 113.0 106.4 153.3 152.0 710 77.7 113.0 113.0 107.6 153.7 153.0 720 79.1 113.0 113.0 108.7 154.1 154.0 730 80.5 113.3 114.1 109.8 154.5 155.0 B-16

GE Nuclear Energy NEDO-33112 TABLE B-3. Browns Ferry Unit 1 P-T Curve Values for 16 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 BOTTOM: UPPER..:, -16EFP.Yz;:.V i.-BOTTOM. - UPPER, 166EFPY

HEAD,--

- ~VESSEL- BELTLiNE;)< HEADE ~.< VESSEL-

-' BELTLINE PRESSURE- CUVEA  :; ,:.'CURVE CURVE 'A A'f~~CURVE BURVE ' B CURVE B (PSIG ()  ;- * -; . (.F  ? F) 740 81.8 114.1 115.7 110.9 154.9 155.9 750 83.1 115.0 117.2 112.0 155.2 156.9 760 84.4 115.8 118.7 113.0 155.6 157.8 770 85.6 116.6 120.2 114.0 156.0 158.7 780 86.8 117.3 121.6 115.0 156.4 159.6 790 88.0 118.1 123.0 116.0 156.8 160.4 800 89.2 118.9 124.3 116.9 157.1 161.3 810 90.3 119.6 125.6 117.9 157.5 162.1 820 91.4 120.4 126.9 118.8 157.9 162.9 830 92.5 121.1 128.2 119.7 158.2 163.8 840 93.5 121.8 129.4 120.6 158.6 164.6 850 94.6 122.5 130.6 121.4 158.9 165.3 860 95.6 123.2 131.7 122.3 159.3 166.1 870 96.6 123.9 132.9 123.1 159.6 166.9 880 97.5 124.6 134.0 124.0 160.0 167.6 890 98.5 125.3 135.0 124.8 160.3 168.4 900 99.4 125.9 136.1 125.6 160.7 169.1 910 100.4 126.6 137.1 126.4 161.0 169.8 920 101.3 127.2 138.2 127.1 161.4 170.5 930 102.1 127.9 139.2 127.9 161.7 171.2 940 103.0 128.5 140.1 128.7 162.0 171.9 950 103.9 129.1 141.1 129.4 162.4 172.6 960 104.7 129.7 142.0 130.1 162.7 173.3 970 105.6 130.3 143.0 130.9 163.0 173.9 980 106.4 130.9 143.9 131.6 163.4 174.6 990 107.2 131.5 144.8 132.3 163.7 175.2 B-17

GE Nuclear Energy NEDO-33112 TABLE B-3. Browns Ferry Unit 1 P-T Curve Values for 16 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 BOTTOMW UPPERS :-16 EFPY :BO16OM >UPPER.'..:.6EFPY

- -iPMAM -"- ,.BE.

- HEAD .--- VESSEL - BELTUENE -:HEA~ `VESSEl- t BELTLINE PRESSURE- -CURVE A- CURVEk--.;

C- CUREVE.B~;kCURVEEA; B- -.

CURVE B (PSIG):: ( Oy-:-°)-; °? r(p< ';° t(F 1000 108.0 132.1 145.6 133.0 164.0 175.8 1010 108.7 132.7 146.5 133.6 164.3 176.5 1020 109.5 133.2 147.3 134.3 164.6 177.1 1030 110.3 133.8 148.2 135.0 165.0 177.7 1040 111.0 134.4 149.0 135.6 165.3 178.3 1050 111.7 134.9 149.8 136.3 165.6 178.9 1060 112.4 135.5 150.6 136.9 165.9 179.5 1064 112.7 135.7 150.9 137.2 166.0 179.7 1070 113.2 136.0 151.4 137.5 166.2 180.1 1080 113.9 136.5 152.1 138.2 166.5 180.7 1090 114.6 137.1 152.9 138.8 166.8 181.2 1100 115.2 137.6 153.6 139.4 167.1 181.8 1105 115.6 137.8 154.0 139.7 167.3 182.1 1110 115.9 138.1 154.4 140.0 167.4 182.4 1120 116.6 138.6 155.1 140.6 167.7 182.9 1130 117.2 139.1 155.8 141.2 168.0 183.4 1140 117.9 139.6 156.5 141.7 168.3 184.0 1150 118.5 140.1 157.2 142.3 168.6 184.5 1160 119.1 140.6 157.9 142.9 168.9 185.0 1170 119.8 141.1 158.5 143.4 169.2 185.6 1180 120.4 141.6 159.2 144.0 169.5 186.1 1190 121.0 142.1 159.9 144.5 169.7 186.6 1200 121.6 142.5 160.5 145.1 170.0 187.1 1210 122.2 143.0 161.2 145.6 170.3 187.6 1220 122.8 143.5 161.8 146.2 170.6 188.1 1230 123.3 143.9 162.4 146.7 170.9 188.6 B-18

GE Nuclear Energy NEDO-3311 2 TABLE B-3. Browns Ferry Unit 1 P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 15 0F/hr for Curve A For Figures 5-11, 5-12, 5-13, 5-15, 5-16, 5-17 & 5-20 BOTTOM UPPER.-- 16EFPY. -:BOTTOMW . -UPPER;,.1 A. 16EFPY

>;- 7 - HEAD- VESSEL.:- BELTLINE -: AHD SEL ELTLINE PRESSURE ',- -'CURVEAS' CURVEW- A : CR EiB,-4 i CURVEBBstC

,PIG .,V'_,'F., _, ,.,'.' ,.;) .> ,, (t ) I ,F 1240 123.9 144.4 163.0 147.2 171.2 189.1 1250 124.5 144.8 163.6 147.7 171.4 189.6 1260 125.0 145.3 164.2 148.2 171.7 190.0 1270 125.6 145.7 164.8 148.7 172.0 190.5 1280 126.1 146.2 165.4 149.2 172.2 191.0 1290 126.7 146.6 166.0 149.7 172.5 191.5 1300 127.2 147.0 166.6 150.2 172.8 191.9 1310 127.7 147.5 167.1 150.7 173.1 192.4 1320 128.3 147.9 167.7 151.2 173.3 192.8 1330 128.8 148.3 168.2 151.6 173.6 193.3 1340 129.3 148.7 168.8 152.1 173.8 193.7 1350 129.8 149.1 169.3 152.6 174.1 194.2 1360 130.3 149.6 169.9 153.0 174.4 194.6 1370 130.8 150.0 170.4 153.5 174.6 195.0 1380 131.3 150.4 170.9 153.9 174.9 195.4 1390 131.8 150.8 171.5 154.4 175.1 195.9 1400 132.3 151.2 172.0 154.8 175.4 196.3 B-1 9

GE Nuclear Energy NEDO-33112 TABLE B4. Browns Ferry Unit 1 Composite P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 15 OF/hr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20

~ABOTTOM

,:..~* ~UPPER RPV'&-.BO190TOMI QPPER 'RP-'V! -%UPPERRPV,&:

~.~EAD BELLIN AT 'HAD BETLIE~rY.~ELTLINEATA`

PRESSURE- CURVEA CURVEA'-CURVE B.&"CURVE'Bi <CURVEC)

.(OSG&iF), I*(F)@

'irtt;(F;(n:,0-0 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 91.0 70 68.0 83.0 68.0 83.0 98.2 80 68.0 83.0 68.0 83.0 104.2 90 68.0 83.0 68.0 83.0 109.3 100 68.0 83.0 68.0 83.0 113.8 110 68.0 83.0 68.0 83.0 117.9 120 68.0 83.0 68.0 83.0 121.7 130 68.0 83.0 68.0 85.2 125.2 140 68.0 83.0 68.0 88.4 128.4 150 68.0 83.0 68.0 91.2 131.2 160 68.0 83.0 68.0 93.9 133.9 170 68.0 83.0 68.0 96.5 136.5 180 68.0 83.0 68.0 98.9 138.9 190 68.0 83.0 68.0 101.2 141.2 200 68.0 83.0 68.0 103.3 143.3 210 68.0 83.0 68.0 105.3 145.3 220 68.0 83.0 68.0 107.3 147.3 B-20

GE Nuclear Energy NEDO-33112 TABLE B4. Browns Ferry Unit 1 Composite P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 lF/hr for Curves B & C and 15 F/hr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20 BOTTOM UPPER RPV .-BOTTOMi>..UPPERRPV&.i:UPPERRPV&

HE BELTLINEAT...

..D.- ,HED ~BELLlENE.AT BELTLINEAT.

-, ill', ~ M 6EEPY t-.6:EE.,.P,, EFI? ,.

PRESSURE-CURVE.A 'CURVE-As.. CURVE BUi BE;CURVE C'- .

RCURVE (PI)() ~ )- F)~ ~~~)v- (?F)-

230 68.0 83.0 68.0 109.1 149.1 240 68.0 83.0 68.0 110.9 150.9 250 68.0 83.0 68.0 112.6 152.6 260 68.0 83.0 68.0 114.2 154.2 270 68.0 83.0 68.0 115.8 155.8 280 68.0 83.0 68.0 117.3 157.3 290 68.0 83.0 68.0 118.8 158.8 300 68.0 83.0 68.0 120.2 160.2 310 68.0 83.0 68.0 121.5 161.5 312.5 68.0 83.0 68.0 121.9 161.9 312.5 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 68.0 143.0 183.0 460 68.0 113.0 68.0 143.0 183.0 B-21

GE Nuclear Energy NEDO-33112 TABLE B-4. Browns Ferry Unit 1 Composite P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 15 'F/hr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20 BOTTOM- UPPERRPV &. BOTTOM- UPPERRPV& - :UPPERRPV&

,,HEAD BELTLINE-AT;e HEAD, ;: BELTUNE:Afw .- BELTLiNEAT.

16'EFPY: I&EE" Y PRESSURE CURVE A :CURVE'A. ',-CURVE B ;- CURVEBZ CURVE'C--!-

(PSiG) F (OF) ' (0 Fy

-* --- (f:

470 68.0 113.0 68.0 143.0 183.0 480 68.0 113.0 70.5 143.0 183.0 490 68.0 113.0 72.8 143.0 183.0 500 68.0 113.0 75.0 143.0 183.0 510 68.0 113.0 77.2 143.0 183.0 520 68.0 113.0 79.2 143.2 183.2 530 68.0 113.0 81.2 144.0 184.0 540 68.0 113.0 83.1 144.8 184.8 550 68.0 113.0 84.9 145.6 185.6 560 68.0 113.0 86.7 146.4 186.4 570 68.0 113.0 88.4 147.1 187.1 580 68.0 113.0 90.0 147.9 187.9 590 68.0 113.0 91.6 148.6 188.6 600 68.0 113.0 93.2 149.1 189.1 610 68.0 113.0 94.7 149.6 189.6 620 68.0 113.0 96.1 150.0 190.0 630 68.0 113.0 97.5 150.4 190.4 640 68.0 113.0 98.9 150.8 190.8 650 68.2 113.0 100.2 151.2 191.2 660 69.9 113.0 101.5 151.7 191.7 670 71.6 113.0 102.8 152.1 192.1 680 73.2 113.0 104.1 152.5 192.5 690 74.7 113.0 105.3 152.9 192.9 700 76.2 113.0 106.4 153.3 193.3 710 77.7 113.0 107.6 153.7 193.7 720 79.1 113.0 108.7 154.1 194.1 B-22

GE Nuclear Energy NEDO-33112 TABLE B4. Browns Ferry Unit I Composite P-T Curve Values for 16 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 BOTTOM UPPER RPV &. BOTTOM. .UPPER RPV.&.; UPP.ER RPV &

=

* HEADV; BELTNE AT.tHAD.BE IBELTLINE AT

.

- EF

... .  :..; 16EFPYW

____-16 § 16-EFPY- ---

PRESSURE CURVEA-CCURVEAX rl-BCURVE Be CURVEIB V CURVEC-.

.F -

730 80.5 114.1 109.8 155.0 195.0 740 81.8 115.7 110.9 155.9 195.9 750 83.1 117.2 112.0 156.9 196.9 760 84.4 118.7 113.0 157.8 197.8 770 85.6 120.2 114.0 158.7 198.7 780 86.8 121.6 115.0 159.6 199.6 790 88.0 123.0 116.0 160.4 200.4 800 89.2 124.3 116.9 161.3 201.3 810 90.3 125.6 117.9 162.1 202.1 820 91.4 126.9 118.8 162.9 202.9 830 92.5 128.2 119.7 163.8 203.8 840 93.5 129.4 120.6 164.6 204.6 850 94.6 130.6 121.4 165.3 205.3 860 95.6 131.7 122.3 166.1 206.1 870 96.6 132.9 123.1 166.9 206.9 880 97.5 134.0 124.0 167.6 207.6 890 98.5 135.0 124.8 168.4 208.4 900 99.4 136.1 125.6 169.1 209.1 910 100.4 137.1 126.4 169.8 209.8 920 101.3 138.2 127.1 170.5 210.5 930 102.1 139.2 127.9 171.2 211.2 940 103.0 140.1 128.7 171.9 211.9 950 103.9 141.1 129.4 172.6 212.6 960 104.7 142.0 130.1 173.3 213.3 970 105.6 143.0 130.9 173.9 213.9 980 106.4 143.9 131.6 174.6 214.6 B-23

GE Nuclear Energy NEDO-33112 TABLE B-4. Browns Ferry Unit I Composite P-T Curve Values for 16 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 BOTTOM UPPER RPV.& ;.BOTTOM ,- UPPER RPV , URPERRRPV &

• , '-;HEAD . BELTNE AT .HEAD BELTLINE AT¢ SrBELTUNEAT.

- 2'. .. ,  ;^ -; 16 EFW.y'- f.. , '. 16FPm ,16FY~.

PRESSUREi'.CURVE A "ICURVEA&'-i;iCURVE BE .,RVEC C B (PSIG)+ '.Ž- (F)] { - (°F) - '1 '(F). ~-4 I 4 > p,(;t w -

990 107.2 144.8 132.3 175.2 215.2 1000 108.0 145.6 133.0 175.8 215.8 1010 108.7 146.5 133.6 176.5 216.5 1020 109.5 147.3 134.3 177.1 217.1 1030 110.3 148.2 135.0 177.7 217.7 1040 111.0 149.0 135.6 178.3 218.3 1050 111.7 149.8 136.3 178.9 218.9 1060 112.4 150.6 136.9 179.5 219.5 1064 112.7 150.9 137.2 179.7 219.7 1070 113.2 151.4 137.5 180.1 220.1 1080 113.9 152.1 138.2 180.7 220.7 1090 114.6 152.9 138.8 181.2 221.2 1100 115.2 153.6 139.4 181.8 221.8 1105 115.6 154.0 139.7 182.1 222.1 1110 115.9 154.4 140.0 182.4 222.4 1120 116.6 155.1 140.6 182.9 222.9 1130 117.2 155.8 141.2 183.4 223.4 1140 117.9 156.5 141.7 184.0 224.0 1150 118.5 157.2 142.3 184.5 224.5 1160 119.1 157.9 142.9 185.0 225.0 1170 119.8 158.5 143.4 185.6 225.6 1180 120.4 159.2 144.0 186.1 226.1 1190 121.0 159.9 144.5 186.6 226.6 1200 121.6 160.5 145.1 187.1 227.1 1210 122.2 161.2 145.6 187.6 227.6 1220 122.8 161.8 146.2 188.1 228.1 B-24

GE Nuclear Energy NEDO-33112 TABLE B-4. Browns Ferry Unit I Composite P-T Curve Values for 16 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & Cand 15 eF/hr for Curve A For Figures 5-14, 5-18, 5-19 & 5-20 BOTTOM UPPERRPV& .'BOTTOM. UPPER RPV&8 UPPER RPV&

HEAD BELTLINEAT;:+ HEAD- <.BELTLINE AT?: BELTLINEAT 16 EFPY,:- .C16 EFPY ^.: 16EFP Y PRESSURE CURVE-A CURVE A' CURVE B ' ,

• s CURVE B . - .CURVE C (PSIG): (F) -l 0

( F) (°F)  :(@F) -- (

1230 123.3 162.4 146.7 188.6 228.6 1240 123.9 163.0 147.2 189.1 229.1 1250 124.5 163.6 147.7 189.6 229.6 1260 125.0 164.2 148.2 190.0 230.0 1270 125.6 164.8 148.7 190.5 230.5 1280 126.1 165.4 149.2 191.0 231.0 1290 126.7 166.0 149.7 191.5 231.5 1300 127.2 166.6 150.2 191.9 231.9 1310 127.7 167.1 150.7 192.4 232.4 1320 128.3 167.7 151.2 192.8 232.8 1330 128.8 168.2 151.6 193.3 233.3 1340 129.3 168.8 152.1 193.7 233.7 1350 129.8 169.3 152.6 194.2 234.2 1360 130.3 169.9 153.0 194.6 234.6 1370 130.8 170.4 153.5 195.0 235.0 1380 131.3 170.9 153.9 195.4 235.4 1390 131.8 171.5 154.4 195.9 235.9 1400 132.3 172.0 154.8 196.3 236.3 B-25

GE Nuclear Energy NEDO-3311 2 APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear Energy NEDO-331 12 C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

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

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

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

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

C-2

GE Nuclear Energy NEDO-3311 2 C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by <151F 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 NEDO-331 12 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 100 0F in one hour period heatup)

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

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

C4

GE Nuclear Energy NEDO-3311 2 APPENDIX C

REFERENCES:

1. T.A. Caine, T 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 NEDO-3311 2 APPENDIX D GE SIL 430 D-1

GE Nuclear Energy NEDO-33112 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 212 0 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 2120 F for Tech Must comply with SIL 251 Spec 100°F/hr heatup to avoid vessel stratification.

and cooldown rate.

Altemate measurement When above 2120F need to above 212 0F. allow for temperature variations (up to 10-150 F lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GE Nuclear Energy NEDO-3311 2 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 100IF/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 I 00F for BWR/6s and 1450F 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 NEDO-33112 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 BWR/6s.

Top head outside Information only. Very slow to respond to RPV surface T/Cs coolant changes. Not avail-able on BWRI6s.

D-4

GE Nuclear Energy NEDO-3311 2 TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Bottom head outside 1 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 NEDO-3311 2 Product

Reference:

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

B.H. Eldridge, Mgr. D.L. Allred, Manager Service Information Customer Service Information and Analysis Notice:

SlLs pertain only to GE BWRs. GE prepares SlLs 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 SILs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SILs.

D-6

GE Nuclear Energy NEDO-331 12 APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy NEDO-33112 IOCFR50, 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 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/cm 2. 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/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 "o":

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 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 # 1 193.3" [4]

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

Girth Weld between Shell Ring #2 and Shell Ring #3 391.5" [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 [2,3], 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 NEDO-3311 2 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 -25" above TAF. Therefore, if it can be shown that the peak fluence at these locations is less than 1.0e17 n/cm 2, it can be safely concluded that all nozzles and welds, other than those included in Tables 4-4 and 4-5, are outside the beltline region of the reactor vessel.

Based on the axial flux profile for EPU, which bounds the pre-EPU axial flux profile, the RPV fluence drops to less than 1.0e17 n/cm 2 at -5" below the BAF and at -7" above TAF. The beltline region considered in the development of the P-T curves is adjusted to include the additional 7" above the active fuel region and the additional 5" below the active fuel region. This adjusted beltline region extends from 211.3" to 373.3" above reactor vessel V0" for 16 EFPY.

Based on the above, it is concluded that none of the Browns Ferry Unit I 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 NEDO-33112 APPENDIX E

REFERENCES:

1. J. Valente (TVA) to Dale Porter (GE), uBrowns Ferry Nuclear Plant (BFN) -

Pressure-Temperature Curves Design Input Request (DIR) - Transmittal of DIR Rev. 0", July 17, 2003 (TVA RIMS No. W83 030717 001).

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

3. Drawing #254185F, Revision 11, NGeneral 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).

E-4

GE Nuclear Energy NEDO-3311 2 APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

F-1

GE Nuclear Energy NEDO-3311 2 Paragraph lV.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 16 EFPY. Calculations of 16 EFPY USE, using Regulatory Guide 1.99, Revision 2 [2] methods and BWROG Equivalent Margin Analyses [3, 4] 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 1 beltline region. Therefore, Browns Ferry Unit 1 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 16 EFPY peak 114T fluence as provided in Tables 4-4 and 4-5. Surveillance capsule data is not available for Browns Ferry Unit 1.

Based on the results presented in Tables F-1 and F-2, the USE EMA values for the Browns Ferry Unit 1 reactor vessel beltline materials remain within the limits of Regulatory Guide 1.99, Revision 2 and 10CFR50 Appendix G for 16 EFPY of operation.

F-2

GE Nuclear Energy NEDO-3311 2

- Table F-I Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit I For 16 EFPY (including Extended Power Uprate)

BWRI3-6 PLATE Surveillance Plate USE:

%Cu = N/A 11t Capsule Fluence = N/A 1" Capsule Measured % Decrease = N/A (Charpy Curves) 1t Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limiting Beitline Plate (Heat B5864-1) USE:

%Cu = 0.15 16 EFPY 1/4T Fluence = 3.96 x 1017 nIcm 2 R.G. 1.99 Predicted % Decrease = 11.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 11.5%

• 21%, so vessel plates are bounded by equivalent margin analysis F-3

GE Nuclear Energy NEDO-3311 2 Table F-2 Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit I For 16 EFPY (including Extended Power Uprate)

BWR/2-6 WELD Surveillance Weld USE:

%Cu = N/A 1"' Capsule Fluence = N/A 1 st Capsule Measured % Decrease = N/A (Charpy Curves) 1s' Capsule R.G. 1.99 Predicted % Decrease = N/A (R.G. 1.99, Figure 2)

Limiting Beftline Weld (Heat 406L44) USE:

%Cu = 0.27 16 EFPY 114T Fluence = 3.96 x I017 n/cm2 R.G. 1.99 Predicted % Decrease = 20 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 20% S 34%, so vessel welds are bounded by equivalent margin analysis F-4

GE Nuclear Energy NEDO-33112 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 BWR/2 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.

F-5