RS-03-010, Supporting Documentation for Amendment to Technical Specifications Section 3.4.11, RCS Pressure and Temperature (P/T) Limits. Attachment 4, Table of Contents - Appendix F

Supporting Documentation for Amendment to Technical Specifications Section 3.4.11, RCS Pressure and Temperature (P/T) Limits. Attachment 4, Table of Contents - Appendix F

ML030410050
Person / Time
Site:	LaSalle
Issue date:	01/31/2003
From:	Jury K Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-03-010 GE-NE-0000-0003-5526-01a, Rev 0
Download: ML030410050 (125)

Text

GE NuclearEnergy Engineering and Technology GE-NE-0000-0003-5526-01a General Electric Company Revision 0 175 Curtner Avenue Class I San Jose, CA 95125 June 2002 Pressure-Temperature Curves For Exelon LaSalle Unit 2 Prepared by:

L.iSenioqngineer Structural Mechanics and Materials Verified by:

B.D. Frew, Principal Engineer Structural Mechanics and Materials Approved by:

B.J. Branlund, Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy GE-NE-0000-0003-5526-01a 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 Exelon and GE, Fluence Analysis, effective 11/14/01, 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 Exelon, 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, 2002

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GE Nuclear Energy GE-NE-0000-0003-5526-01a 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 2000 [1]. The P-T curve methodology includes the

-following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 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 K~c of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. This report incorporates a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER

[14b], and is in compliance with Regulatory Guide 1.190.

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

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

"* Closure flange region (Region A)

"* Core beltline region (Region B)

"* Upper vessel (Regions A & B)

"* Lower vessel (Regions B & C)

For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 1000F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the

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GE Nuclear Energy GE-NE-0000-0003-5526-01 a 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 20°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 beitline curves this approach has added conservatism because irradiation effects cause the allowable toughness, KIr, at 1/4T to be less than that at 3/4T for a given metal temperature.

Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beitline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beltline (at 20 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions. A composite P-T curve was also generated for the Core Critical condition at 20 EFPY.

GE Nuclear Energy GE-NE-0000-0003-5526-O la TABLE OF CONTENTS

1.0 INTRODUCTION

I 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITIAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 13 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 18

5.0 CONCLUSION

S AND RECOMMENDATIONS 50

6.0 REFERENCES

65

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GE Nuclear Energy GE-NE-0000-0003-5526-01a 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 EVALUATION FOR UPPER SHELF ENERGY (USE)

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GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE OF FIGURES FIGURE 4-1 SCHEMATIC OF REACTOR VESSEL ARRANGEMENT SHOWING PLATES AND WELDS 10 FIGURE 4-2. CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 30 FIGURE 4-3. FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 36 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 53 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [20°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 54 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 20 EFPY

[20'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 55 FIGURE 5-4: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 32 EFPY

[2 0 'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] []00°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 57 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 58 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 20 EFPY

[100'F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-8: BELTLINE P-T CURVES FOR CORE NOT CRITICAL [CURVE B] UP TO 32 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 FIGURE 5-9: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 20 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 61 FIGURE 5-10: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 32 EFPY

[20°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-11: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 32 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 63 FIGURE 5-12: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 32 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64

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GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR LASALLE UNIT 2 VESSEL MATERIALS 1 TABLE 4-2: RTNr VALUES FOR LASALLE UNIT 2 NOZZLE & WELD MATERIALS 12 TABLE 4-3: LASALLE UNIT 2 BELTLINE ART VALUES (20 EFPY) 16 TABLE 4-4: LASALLE UNIT 2 BELTLINE ART VALUES (32 EFPY) 17 TABLE 4-5:

SUMMARY

OF THE 10CFRS0 APPENDIX G REQUIREMENTS 20 TABLE 4-6: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 22 TABLE 4-7: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 22 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 52

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GE Nuclear Energy GE-NE-0000-0003-5526-01 a

1.0 INTRODUCTION

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

Complete P-T curves were developed for 20 and 32 effective full power years (EFPY).

The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. The P-T curves incorporate a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190.

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 2000 [1]. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640 [4]. 2) The use of the Mm calculation in the 1995 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 Kc of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. 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 is documented in Section 4.1.

Adjusted Reference Temperature (ART) is the reference temperature when including irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 [7] provides the methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 20 and 32 EFPY are included in Section 4.2. The 32 EFPY peak ID fluence value of GE Nuclear Energy GE-NE-0000-0003-5526-01 a 1.09 x 1018 n/cm2 used in this report is 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 the discontinuities.

Guidelines and requirements for operating and temperature monitoring are included in Appendix C. GE SIL 430, a GE service information letter regarding Reactor Pressure Vessel Temperature Monitoring is included in Appendix D. Appendix E demonstrates that all reactor vessel nozzles (other than the LPCI nozzle) are outside the beltline region. Finally, Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 2.0 SCOPE OF THE ANALYSIS The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curves in this report incorporate a fluence [14a] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190. A detailed description of the P-T curve bases is included in Section 4.3. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 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 of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. 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 LaSalle Unit 2 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].

Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize heating requirements prior to pressure testing. Operating and temperature monitoring GE Nuclear Energy GE-N E-0000-0003-5526-O1 a 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 (other than the LPCI nozzle) are outside the beltline region. Finally, Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE).

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:

For end-of-license (32 EFPY) fluence an 80% capacity factor is used to determine the EFPY for a 40-year plant life. The 80% capacity factor is based on the objective to have BWR's available for full power production 80% of the year (refueling outages, etc. -20%

of the year).

The shutdown margin is calculated for a water temperature of 680 F, as defined in the LaSalle Unit 2 Technical Specification, Section 1.1.

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 and are summarized as follows:

a. Test specimens shall be longitudinally oriented CVN specimens.

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

c. Pressure tests shall be conducted at a temperature at least 60°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 60°F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion is 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.

I OCFR50 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 GE Nuclear Energy GE-NE-0000-0003-5526-01 a must be supplemented in an approved manner. GE developed methods for analytically 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 LaSalle Unit 2 vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDT calculations for vessel plate, weld, HAZ, and forging, and bolting material LST are summarized in the remainder of this section.

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 LaSalle Unit 2 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 2°F per ft-lb energy difference from 50 ft-lb.

For example, for the LaSalle Unit 2 beltline plate heat C9404-2 in the lower-intermediate shell course, the lowest Charpy energy and test temperature from the CMTRs is 29 ft-lb at 400F. The estimated 50 ft-lb longitudinal test temperature is:

TSOL = 40°F + [(50 - 29) ft-lb *20F/ft-lb] = 820 F The transition from longitudinal data to transverse data is made by adding 30°F to the 50 ft-lb transverse test temperature; thus, for this case above, T-0T = 82 0F + 30°F = 112 0 F GE Nuclear Energy GE-N E-0000-0003-5526-01 a The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (Tso-r 60 0F). Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is 100 F. Thus, the initial RTNDT for plate heat C9404-2 is 520 F.

For the LaSalle Unit 2 beltline weld heat 3P4966 with flux lot 1214 (contained in the lower-intermediate shell), the CVN results are used to calculate the initial RTNDT. The 50 ft-lb test temperature is applicable to the weld material, but the 30°F adjustment to convert longitudinal data to transverse data is not applicable to weld material. Heat 3P4966 has a lowest Charpy energy of 28 ft-lb at 10°F as recorded in weld qualification records. Therefore, TsoT = 10F + [(50 - 28) ft-lb

• 2°F/ft-lb] = 54°F The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (TSoT - 60°F). For LaSalle Unit 2, the dropweight testing to establish NDT was not recorded for most weld materials. GE procedure requires that, when no NDT is available for the weld, the resulting RTNDT should be -50°F or higher. The value of (T5oT - 60 0F) in this example is -6°F; therefore, the initial RTNDT was -6°F.

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 feedwater nozzle at LaSalle Unit 2 (Heat Q2Q25W), the NDT is -20°F and the lowest CVN data is 28 ft-lb at -200 F. The corresponding value of (Ts0T - 60 0F) is:

(T5oT - 60°F) = {[-20 + (50 - 28) ft-lb "2°F/ft-lb] + 30 0 F} - 60OF = -60F.

GE Nuclear Energy GE-NE-0000-0003-5526-01a Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (TsoT- 60 0F), which is -60F.

In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNOT. For the lower torus shell of LaSalle Unit 2 (Heat C9306-2), the NDT was not available and the lowest CVN data was 33 ft-lb at 40°F. The corresponding value of (TS0T - 60 0F) was:

(T50T - 60 0F) = {[40 0F + (50 - 33) ft-lb

• 2°F/ft-lb] + 300F} - 60°F = 440 F.

Therefore, the initial RTNDT was 440 F.

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 60°F (as discussed in Section 4.3.2.3) is the LST for the bolting materials. Charpy data for the LaSalle Unit 2 closure studs do not meet the 45 ft-lb, 25 MLE requirement at 10°F. Therefore, the LST for the bolting material is 70°F. The highest RTNDT in the closure flange region is 260F, for the vessel upper shell materials. Thus, the higher of the LST and the RTNDT +600 F is 86°F, the boltup limit in the closure flange region.

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

GE Nuclear Energy GE-NE-0000-0003-5526-0 a TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL #4 0

SHELL #3

• /I 0/////////

LPCI NOZZLE TOP OF BELTLINE REGION 376.3125" TOP OF ACTIVE FUEL SHELL #2 (TAF) 366.3125" ELDS

,BF ELDAB SHELL #1 BOTTOM OF ACTIVE FUEL (BAF) 216.3125" BOTTOM OF BELTLINE REGION 206.3125" BOTTOM HEAD SUPPORT SKIRT Notes: (1) Refer to Tables 4-1 and 4-2 for reactor vessel components and their heat identifications.

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

Figure 4-1: Schematic of the LaSalle Unit 2 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-NE-0000-0003-5526-01 a Table 4-1: RTNDT Values for LaSalle Unit 2 Vessel Materials IPH TEST TEM, CHARPY ENERGY JDO (Ts.r-60) DROP COMPONENT HEAT TEMP. RTNDT (F

(FT-LB) ("F)WEIGHT RTND CF) (IF) NDT (F)

PLATES & FORGINGS:

Top Head & Flange:

Top Head: Torus Plate B3269-1 10 75 72 71 -20 10 Torus Plate B3269-2 10 50 60 50 -20 10 Dollar Plate C9195-3 40 30 30 33 50 50 Top Head Flange BWK-446 10 70 122 142 -20 20 Shell Flange BRC424 10 103 110 105 -20 10 Shell Courses:

Upper Shell C9678-1 10 47 66 66 -14 10 Mk-24 A8453-1 10 44 27 42 26 26 C9507-1 10 51 53 56 -20 10 Upper Int. Shell C9569-1 40 73 69 64 10 40 Mk-23 C9481-2 40 46 55 62 18 40 C9602-2 40 56 45 58 20 40 Low-Int. Shell C9404-2 40 48 44 29 52 10 Mk-22 C9481-1 40 103 61 85 10 -30 C9601-2 40 85 93 74 10 -30 Lower Shell C9425-1 40 39 43 48 32 0 Mk-21 C9425-2 40 44 40 49 30 -30 C9434-2 40 91 58 72 10 -10 Bottom Head: C9306-2 40 36 38 33 44 44 C9514-2 40 50 59 71 10 40 C9621-1 40 63 64 72 10 40 C9245-1 40 61 53 62 10 40 Support Skirt: A8699-3 40 30 31 37 50 50 A8879-1 B 40 30 30 33 50 50 A8418-4 40 34 33 30 50 50 C9491-IB 40 77 74 68 10 40 STUDS: LST Studs 82552 10 44 45 40 70 Nuts 10134-48 10 51 52 43 70

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GE Nuclear Energy GE-NE-0000-0003-5526-O1 a Table 4-2: RTNDT Values for LaSalle Unit 2 Nozzle & Weld Materials 1

• TEST DRO COMPONENT HEAT TEMP.

T(OF ESI C CKARPY ENERGY (FT-LB) Te)IDO Ts-0

('F) WEIGHT DRP NDT RTNOT

('F)

NOZZLES:

Recirculation Outlet Nozzle, N1 Q2Q32W 40 54 39 40 32 40 40 Recirculation Inlet Nozzle. N2 02Q33W 40 54 62 49 12 40 40 Q2Q25W 40 82 50 77 10 40 40 Q2Q36W 40 94 87 105 10 40 40 02Q42W 40 82 98 98 10 40 40 Steam Outlet Nozzle, N3 Q2Q30W 40 45 49 48 20 40 40 Q2032W 40 62 49 58 12 40 40 Feedwater Nozzle, N4 02033W -20 35 37 43 -20 -20 -20 02Q25W -20 38 35 28 -6 -20 -6 02Q29W -20 63 52 38 -26 -20 -20 LP Core Spray Nozzle, N5 02025W -20 40 26 36 -2 -20 -2 HP Core Spray Nozzle, N16 02029W -20 38 42 53 -26 -20 -20 RHR/LPCI Nozzle, N6 02036W -20 44 37 28 -6 -20 -6 Q2042W -20 57 49 38 -26 -20 -20 Head Spray Nozzle, N7 02Q33W -20 64 70 93 -50 -20 -20 Vent Nozzle, N8 02019W 40 69 65 58 10 40 40 Jet Pump Instrumentation Nozzle, N9 02026W 40 29 30 41 52 52 52 CRD Hyd. System Return Nozzle, N10 02Q23W -10 39 30 43 0 -10 0 Drain Nozzle. N15 265M-1 -10 55 34 42 -8 -8 -8 WELDS:

Vertical Welds:

BA, BB, BC 3P4000 10 86 87 90 -50 -50 -50 BD, BE, BF 3P4966 10 28 84 63 -6 -50 -6 BG. BJ, BH &BK, BM, BN 4P4784 10 71 73 73.5 -50 -50 -50 Girth Welds:

AA.AC 3P4966 10 28 84 63 -6 -50 -6 AB 5P6771 10 57 55 42 -34 -50 -34 AD 5P6214B 10 37 54 47 -24 -50 -24 Bottom Head Assembly Welds:

DA, DB, DC, DD, DE, DF 3P4000 10 86 87 90 -50 -50 -50 Top Head Assembly Welds:

DM, DN, DP. DH, DJ, DK 3P4966 10 28 84 63 -6 -50 -6 AG 5P6214B 10 37 54 47 -24 -50 -24 GE Nuclear Energy GE-NE-0000-0003-5526-01 a 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 (Rev 2) provides the methods for determining the ART. The Rev 2 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 beitline plates and welds was made and summarized in Table 4-3 for 20 EFPY and Table 4-4 for 32 EFPY.

4.2.1 Regulatory Guide 1.99, Revision 2 (Rev 2) 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 Rev 2, the SHIFT equation consists of two terms:

SHIFT = ARTNDT + Margin where, ARTNDT = [CF]*f (0.28-0.10logt Margin = 2(al 2 + 0. )0.5 CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = 1/4Tfluence / 1019 2

Margin = 2(cra + y2)0.5 0., = standard deviation on initial RTNDT, which is taken to be 0°F.

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

ART = Initial RTNDT + SHIFT The margin term aA has constant values in Rev 2 of 170F for plate and 28*F for weld.

However, a,&need not be greater than 0.5

• ARTNDT. Since the GE/BWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in GE Nuclear Energy GE-N E-0000-0003-5526-01 a Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value of a, is taken to be 0°F for the vessel plate and weld materials.

4.2.1.1 Chemistry The vessel beltline chemistries for LaSalle Unit 2 were obtained from several sources.

The vessel plate copper values were obtained from the plate manufacturer [5a] and the nickel values were obtained from the CMTRs [12]. Submerged arc weld properties were obtained from separate evaluations [13a, 13b, and 13c]. The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of Rev 2, to determine a chemistry factor (CF) per Paragraph 1.1 of Rev 2 for welds and plates, respectively.

4.2.1.2 Fluence A LaSalle Unit 2 flux for the vessel IDwall [14a] was calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in SER [14b], and is in compliance with Regulatory Guide 1.190.The flux as documented in [14] is determined for the currently licensed power of 3489 MWt using a conservative power distribution and is conservatively used from the beginning to the end of the licensing period (32 EFPY).

The peak fast flux for the RPV inner surface from Reference 14 is 1.08e9 n/cm2-s. The peak fast flux for the RPV inner surface determined from surveillance capsule flux wires removed during the outage following Fuel Cycle 6 at 6.98 EFPY and at a full power of 3323 MWt is 5.22e8 n/cm 2-s [5b]. Linearly scaling the Reference 5 flux by 1.05 to the currently licensed power of 3489 MWt results in an estimated flux of 5.48e8 n/cmý-s.

Therefore, the Reference 14 flux bounds the flux determined from the surveillance capsule flux wire results by 197%.

The time period 32 EFPY is 1.01e9 sec, therefore the RPV IDsurface fluence is as follows: RPV ID surface fluence = 1.08e9 n/cm2 -s*1.01e9 s = 1.09e18 n/cm2 . This fluence applies to the lower-intermediate plates and welds. The fluence is adjusted for the lower plates and welds and the girth weld based upon a peak I lower shell location ratio of 0.88 (at an elevation of 277" above vessel "0"); hence the peak ID surface fluence for these components is 9.59e17 n/cm 2. Similarly, the fluence is adjusted for the GE Nuclear Energy GE-NE-0000-0003-5526-01a LPCI nozzle based upon a peak / LPCI nozzle location ratio of 0.244 (at an elevation of 355" above vessel "0"and at 45°. 1350, and 2250 azimuths); hence the peak IDsurface fluence used for this component is 2.66e17 n/cm2 .

4.2.2 Limiting Beltline Material The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, Rev 2 was applied to compute ART. For LaSalle Unit 2, the LPCI nozzle is the limiting material for the beltline region for 32 EFPY as discussed in Section 4.3.2.2.2. At 20 EFPY, the P-T curves are not beltline limited.

Table 4-3 lists values of beltline ART for 20 EFPY and Table 4-4 lists the values for 32 EFPY. Sections 4.3.2.2.2 and 4.3.2.2.3 provide a discussion of the limiting material.

GE Nuclear Energy GE-NE-0000-0003-5526-01a Table 4-3: LaSalle Unit 2 Beltline ART Values (20 EFPY)

Lower-4murdIte IPlat" ad WehdiED,3=. 31 Thlickness in indhes 619 Ratio Peak/Location , 1.00 32 EFPY Peak I.D. iluence , 1.09E-18 nlom^2 32 EFPY Peak 1/4 T fluenoe 7.5E+17 nVanI2 20 EFPY Peak 114T fluence= 4.7E+17 rdan^2 Loer rlata and Welds A. M M Girth Wdd AS Thickness in hinces 8.19 Rado Peak/ Locatn = 0.88 32 EFPY Peak i.D. fluence

• 9.59E+17 nicmt2 Ele.allon - 27r 32 EFP`Y Peak 1/4 T Ikience

• 6.6E017 n/cmn2 20 EFPYPeakl/4Tfluence. 4.10E17 n/tmi2 3.10 i nI=.

Thickeaa in inches 6.19 Ratio Peak/Location - 0244 32 EFPY Peak I.D. Iluence , 2.66E17 rVcm^2 Eilevllon -355 32 EFPY Peak 114 T luenoe = 1.6E017 rVo2 20 EFPYPeakl4Tfluenoe= 1.10E17 n/cmW2 hnital 114T 20 EFPY 20 EFPY 20 EFPY COMPONENT HEAT OR HEATALOT %Cu %m CF RTmr FRuence a RTmT a, aA margin Shift ART

"*F ,VcMA2 -F F *F -F PLATES:

Lower She.

21-1 C9425-2 0.120 0.510 81 30 4.1E017 21 0 11 21 43 73 21-2 C9425-1 0.120 0.510 81 32 4.1E+17 21 0 11 21 43 75 21-3 C9434-2 0.090 0.510 58 10 4.1E+17 15 0 8 15 31 41 Lower4ntenmedlate shell 22-1 C9481-1 0.110 0.500 73 10 4.7E+17 21 0 10 21 41 51 22-2 C9404-2 0.070 0.490 44 52 4.7E+17 12 0 6 12 25 77 22-3 C9601-2 0.120 0.500 81 10 4.7E+17 23 0 11 23 46 58 WELDS:

Lower Vertical BA.BB,BC 3P4000 13933 0.020 0.930 27 -. 0 4.1E+17 7 0 4 7 14 -36 Lower-lntermedlate Vertical BD.BE.BF 3P4966 / 1214 0.026 0.920 41 -6 4.7E+17 12 0 6 12 23 17 Girth AB 5P677110342 0.040 0.940 54 -34 4.1E.17 14 0 7 14 28 .6 LPCI 02036W 0.220 0.830 177 .6 1.1+E17 21 0 11 21 43 37 GE Nuclear Energy GE-NE-0000-0003-5526-01 a Table 4-4: LaSalle Unit 2 Beltline ART Values (32 EFPY)

Lm-.ttrmedhAt. IPlateand WeddBD.33 311 Thickness in inchmes 6.19 Ratio Peak/ Location

• 1.00 32 EFPY Peak 1.0.fluence

• 1092.18 wNcm'"2 32 EFPY Peak 14 T fluence 7.5E+17 32 EFPYPeak1/4Tfluence, 7.5E17 n(cmA2 Lower Pln andWeld BA"S3. W. Glrth WeldAN Thickness In inches 6.19 Ratio Peak/Location- 0.88 32 EFPY Peak I.D. Ituence - 9.59E+17 Flevation - 277 32 EFPY Peak 114 T fluence

• 8.BE+17 n/cnn'2 32 EFPYPeakl14Tfiuence. 8.6E*17 n/crrn2 1.1,C Nozzle Thicknmesin indies. 6.19 Ratio Peak/ Location - 0.244 32 EFPY Peak I.0. fluence - 2.66E+17 n/cm`2 Elevation -355" 32 EFPY Peak 1/4 T fluence.. 1.8E+17 32 EFPY Peak 1/4 T fluence - 1.8$E17 N'cm"2 Initial 114T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %M CF RTa, Fluence A RT. 7 a, a. Margin Shift ART

-F rVCM"2 F "F "F -F PLATES:

Lower Shen 21-1 C9425-2 0.120 0.510 81 30 6.6E+17 27 0 14 27 55 85 21-2 C9425-1 0.120 0.510 81 32 6.6E-17 27 0 14 27 55 87 21-3 C9434-2 0090 0.510 58 10 6.6E+17 20 0 10 20 39 49 Lower4-ntenedlate Shell 22-1 C9481-1 0.110 0.500 73 10 7.5E÷17 28 0 13 28 53 63 22-2 C9404-2 0.070 0.490 44 52 7.5E+17 16 0 8 16 32 84 22-3 C9601-2 0.120 0.500 81 10 7.5E+17 29 0 15 29 59 69 WELDS:

Lower Vertical BA, BB, BC 3P4000/ 3933 0020 0930 27 -50 8.6E÷17 9 0 5 9 18 -32 Lower-hltarmedlatl vertical BO, BE. BF 3P4966 1214 0.026 0.920 41 -6 7.5E+17 15 0 7 15 30 24 GMrth AS 5P6771 /0342 0.040 0.940 54 -34 6.6E-17 18 0 9 18 37 3 LP21 02036W 0.220 10.830 177 -6 1.82.17 20 0 14 29 58 1 52 GE Nuclear Energy GE-NE-0000-0003-5526-O1a 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 Appendix G [8] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions that a pressure-retaining component may be subjected to over its service lifetime. The ASME Code (Appendix G of Section XI of the ASME Code [6]) forms the basis for the requirements of 10CFR50 Appendix G. The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

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

"* Closure flange region (Region A)

"* Core beltline region (Region B)

"* Upper vessel (Regions A & B)

"* Lower vessel (Regions B & C)

The closure flange region includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTNDT. The remaining portion 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 1OOF/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also GE Nuclear Energy GE-NE-0000-0003-5526-01a 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 the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20 °F/hr or less must be maintained at all times.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, K4r, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

The applicable temperature is the greater of the 10CFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-5:

GE Nuclear Energy GE-NE-0000-0003-5526-01a Table 4-5: Summary of the 10CFR50 Appendix G Requirements Operating Condition and Pressure Minimum Temperature Requirement

1. 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 + 60 0F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 90°F I1. Normal operation (heatup and cooldown),

including anticipated operational occurrences

a. Core not critical - Curve B

1. At < 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 60 0F*

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

b. Core critical - Curve C

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

2. At > 20% of preservice hydrotest Larger of ASME Limits + 40°F or of pressure a.2 + 40=F or the minimum permissible temperature for the inservice system hydrostatic pressure test 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 IOCFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

GE Nuclear Energy GE-NE-0000-0003-5526-01 a GE PROPRIETARY INFORMATION DELETED 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0e17 n/cm2 ) to cause any significant shift of RTNDT (see Appendix E). Non-beltline components include nozzles, 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-beitline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include 100°F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWR/6 components: the feedwater nozzle (FW) and the CRD penetration (bottom GE Nuclear Energy GE-N E-0000-0003-5526-01 a head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-6 and 4-7.

Table 4-6: Applicable BWR/5 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification FW Nozzle LPCI 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 Instrumentation Nozzle Shell CRD and Bottom Head (Bonly)

Top Head Nozzles (Bonly)

Recirculation Outlet Nozzle (Bonly)

Table 4-7: Applicable BWR/5 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity Identification CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Shefl**

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, since separate bottom head P-T curves are provided to monitor the bottom head.

GE Nuclear Energy GE-NE-0000-0003-5526-01a 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 LaSalle Unit 2 as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4. The generic value was adapted to the conditions at LaSalle Unit 2 by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.

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

4.3.2.1.1 PressureTest - Non-Beitline, 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, K4. The 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 Rsi-in1 2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 84°F.

GE Nuclear Energy GE-N E-0000-0003-5526-O1 a The limit for the coolant temperature change rate Is 20°Flhr or less.

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

was based on a thickness of 8.0 inches; hence, tV2 = 2.83. The resulting value obtained was:

Mm = 1.85 for _<2 M, = 0.926 ft for 2<f <3.464 = 2.6206 Mm = 3.21 for N >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 *apm = ksi-in"2 KIb = (2/3) Mmpb = ksi-in"2 The total K,is therefore:

K = 1.5 (K*m+ Kib) + Mm" (aSm + (2/3)" Osb) = 143.6 ksi-inl2 GE Nuclear Energy GE-NE-0000-0003-5526-01 a This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNoT) for a specific K,is based on the K, equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) = In [(K1 - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 84°F The generic curve was generated by scaling 143.6 ksi-in"2 by the nominal pressures

.. W and calculating the associated (T - RTNDT):

The highest RTNDT for the bottom head plates and welds is 440 F, as shown in Tables 4-1 and 4-2.

GE Nuclear Energy GE-NE-0000-0003-5526-01 a Second, the P-T curve is dependent on the calculated K,value, and the K,value is proportional to the stress and the crack depth as shown below:

K* cc c (7ra)"* (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, i, is proportional to RI(t)f 2 . The generic curve value of R/(t)"2 , based on the generic BWR/6 bottom head dimensions, is:

Generic: R / (t)t2 = 138 / (8)"12 = 49 inch1 2 (4-2)

The LaSalle Unit 2 specific bottom head dimensions are R = 126.7 inches and t =7.13 inches minimum [19], resulting in:

LaSalle Unit 2 specific: R / (t)'12 = 126.7 / (7.13)12 = 47.5 inch"r (4-3)

-26 -

GE Nuclear Energy GE-NE-0000-0003-5526-01a Since the generic value of R/(t)'" is larger, the generic P-T curve is conservative when applied to the LaSalle Unit 2 bottom head.

4.3.2.1.2 Core Not CriticalHeatup/Cooldown- Non-Beltilne 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 K4 value for the core not critical condition is (143.6 / 1.5)

• 2.0 = 191.5 ksi-in"2 .

GE Nuclear Energy GE-NE-0000-0003-5526-01 a h..

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

(T - RTNDT) = In [(14 - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 102°F The generic curve was generated by scaling 192 ksi-in'" by the nominal pressures and calculating the associated (T - RTNDT):

Core Not Critical CRD Penetration K,and (T - RTNDT) as a Function of Pressure Nominal Pressure  !( T - RTNDT (psig) (ksi-in" 2) (-F) 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 440 F, as shown in Tables 4-1 and 4-2.

As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Tables 4-6, 4-7, and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe

-28 -

GE Nuclear Energy GE-NE-0000-0003-5526-01 a than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.

GE Nuclear Energy GE-NE-0000-0003-5526-01a Figure 4-2. CRD Penetration Fracture Toughness Limiting Transients GE Nuclear Energy GE-NE-0000-0003-5526-01 a 4.3.2.1.3 PressureTest - Non-Beltline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, KI, 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 r2 for an applied pressure of 1563 psig preservice hydrotest pressure.

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

To evaluate the results, K,is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or 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 K1 is shown below using the BWR/6, 251-inch dimensions:

Vessel Radius, R, 126.7 inches Vessel Thickness, t, 6.1875 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psig. 126.7 inches / (6.1875 inches) = 32,005 psi.

The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.97 ksi. The factor F (a/rn) from Figure A5-1 of WRC-1 75 is 1.4 where:

a = %(t, 2 + tv 2)1/2 =2.36 inches t, = thickness of nozzle = 7.125 inches t, = thickness of vessel = 6.1875 inches r, = apparent radius of nozzle = ri + 0.29 rc=7.09 inches r1 = actual inner radius of nozzle = 6.0 inches rc = nozzle radius (nozzle corner radius) = 3.75 inches GE Nuclear Energy GE-N E-0000-0003-5526-01 a Thus, airm = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an aift of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (na)1 2 . F(a/rn):

2 Nominal K,= 1.5 34.97. (x. 2.36)12 .1.4 = 200 ksi-in1 The method to solve for (T - RTNDT) for a specific K,is based on the Kk, equation of Paragraph A-4200 in ASME Appendix A [17] for the pressure test condition:

(T - RTNDT) = In [(K1- 33.2) / 20.734] / 0.02 (T - RTNoT) = In [(200 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 104.2°F 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),

GE Nuclear Energy GE-NE-0000-0003-5526-01a The highest RTNDT for the feedwater nozzle materials is 40°F as described below. The generic pressure test P-T curve is applied to the LaSalle Unit 2 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 400 F.

GE Nuclear Energy GE-NE-0000-0003-5526-01a Second, the P-T curve is dependent on the K,value calculated. The LaSalle Unit 2 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and KIare shown below:

Vessel Radius, Rv 126.7 inches Vessel Thickness, t, 6.19 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psig. 126.7 inches / (6.19 inches) = 31,992 psi. The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.96 ksi. The factor F (a/r) from Figure A5-1 of WRC-175 is determined where:

a I=(t 2 + tv 2 )112 =2.36 inches tn = thickness of nozzle = 7.125 inches t, = thickness of vessel = 6.19 inches rn = apparent radius of nozzle = r1 + 0.29 r,=6.8 inches ri = actual inner radius of nozzle = 6.0 inches r, = nozzle radius (nozzle comer radius) = 2.75 inches Thus, alrn = 2.36 / 6.8 = 0.35. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an air, of 0.35, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (na)7". F(a/rn):

Nominal K,= 1.5- 34.96 - (7 - 2.36)"2 .1.4 = 199.9 ksi-in" 2 GE Nuclear Energy GE-NE-0000-0003-5526-01 a 4.3.2.1.4 Core Not CriticalHeatup/Cooldown- Non-Beitline Curve B (Using FeedwaterNozzle/Upper Vessel Region)

The feedwater nozzle was selected to represent non-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter 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 these was normal operation with cold 40°F feedwater injection, which is equivalent to hot standby, see 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)1/2

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

GE Nuclear Energy GE-NE-0000-0003-5526-01 a Figure 4-3. Feedwater Nozzle Fracture Toughness Limiting Transient Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/rn) 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, a.r, and primary bending, ap. Secondary membrane, am, and secondary bending, ao1 , stresses are included in the total KI by using ASME Appendix G [6] methods for secondary portion, Ks:

KI, = Mm (asm + (2/3)" asb) (4-5)

GE Nuclear Energy GE-NE-0000-0003-5526-01 a In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [15].

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

Once KIwas 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 Kc equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

(T - RTNDT) = In [(K1 - 33.2) / 20.734] / 0.02 (4-6)

Example Core Not Critical HeatuplCooldown Calculation for Feedwater Nozzle/Upper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the feedwater nozzle analysis, where feedwater injection of 40°F into the vessel while at operating conditions (551.4°F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle corner stresses were obtained from finite element analysis [ ]. To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 inches was used in the evaluation.

However, a thickness of 7.5 inches is not conservative for the pressure stress evaluation. Therefore, the pressure stress (apm) was adjusted for the actual vessel thickness of 6.1875 inches (i.e., 0 pm = 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:

apm = 24.84 ksi as, = 16.19 ksi a = 45.0 ksi tv = 6.1875 inches apb = 0.22 ksi a*b = 19.04 ksi a = 2.36 inches r, = 7.09 inches t, = 7.125 inches GE Nuclear Energy GE-NE-0000-0003-5526-01 a In this case the total stress, 60.29 ksi, exceeds the yield stress, ac,, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 [15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the temperature assumed for the crack root is the inside surface temperature.)

R = [ay - om + ((ototai - cy) / 30)] / (aOota - apr) (4-7)

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

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

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

Mm = 1.85 for N/tV<2 Mm = 0.926 fti for 2<4:<3.464 = 2.845 Mm = 3.21 for ft">3.464 The value F(a/r,), taken from Figure A5-1 of WRC Bulletin 175 for an aft, of 0.33, is therefore, F (a / rn) = 1.4 Kip is calculated from Equation 4-4:

Kip = 2.0 (24.84 + 0.13) . (7 . 2.36)'2 . 1.4 Kip = 190.4 ksi-in'12 GE Nuclear Energy GE-NE-0000-0003-5526-01a K15 is calculated from Equation 4-5:

Kis = 2.845. (9.44 + 2/3- 11.10) 2 K1 s = 47.9 ksi-in 1 The total K,is, therefore, 238.3 ksi-in1'2.

The total K4 is substituted into Equation 4-6 to solve for (T - RTNDT):

(T - RTNDT) = In [(238.3- 33.2) / 20.7341 / 0.02 (T- RTNDT)= 115°F The curve was generated by scaling the stresses used to determine the K1; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40°F water injected into the hot reactor vessel nozzle. In the base case that yielded a K,value of 238 ksi-in'12, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4°F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tswraon - 40) / (551.4 - 40). From K, the associated (T - RTNDT) can be calculated:

GE Nuclear Energy GE-NE-0000-0003-5526-01a Core Not Critical Feedwater Nozzle K,and (T - RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp. R K* (T - RTNDT)

(psig) (OF) (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 K4.

The highest non-beltline RTNDT for the feedwater nozzle at LaSalle Unit 2 is 40°F as shown in Tables 4-1 and 4-2 and previously discussed. The jet pump instrumentation nozzle is not limiting, as previously discussed. The generic curve is applied to the LaSalle Unit 2 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 40°F as discussed in Section 4.3.2.1.3.

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

GE Nuclear Energy GE-NE-0000-0003-5526-01a 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 100°F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits.

4.3.2.2.1 Beitline Region - PressureTest The methods of ASME Code Section XI, Appendix G [6] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum thickness (trn) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

c'm = PR / trin (4-8)

The stress intensity factor, Km, 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 K1c equation of Paragraph A-4200 in ASME Appendix A [17]

for the pressure test condition:

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

GE Nuclear Energy GE-NE-0000-0003-5526-01a 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, Kt, for a coolant heatup/cooldown rate of 20 °F/hr 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 K1t calculation for a coolant heatup/cooldown rate of 100lF/hr is described in Section 4.3.2.2.3 below.

4.3.2.2.2 Calculationsfor the Beitline Region - PressureTest This sample calculation is for a pressure test pressure of 1105 psig at 32 EFPY. The following inputs were used in the beltline limit calculation:

Adjusted RTNDT = Initial RTNDT + Shift A = 32 + 55 = 87°F (Based on ART values in Section 4.2)

Vessel Height H = 870.5 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to inside of clad) R = 126.5 inches Minimum Vessel Thickness (without clad) t = 6.19 inches Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1105 psi + (H - B) 0.0361 psi/inch = P psig (4-10)

= 1105 + (870.5 - 216.3) 0.0361 = 1129 psig Pressure stress:

a = PR/t (4-11)

= 1.129 - 126.5 / 6.19 = 23.1 ksi GE Nuclear Energy GE-NE-0000-0003-5526-01 a The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on a thickness of 6.19 inches (the minimum thickness without cladding);

hence, t" 2 = 2.49. The resulting value obtained was:

Mm = 1.85 for ftV<2 am = 0.926 f-"for 2< Ft<3.464 = 2.30 Mm = 3.21 for ft >3.464 The stress intensity factor for the pressure stress is Kim = Mm " a. The stress intensity factor for the thermal stress, K,, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20°F/hr instead of 100°F/hr.

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

Using the K1c equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 53.1, and Kit = 2.58 for a 20°F/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)

= ln[(1.5. 53.1 + 2.58 - 33.2) / 20.734] / 0.02

= 43.0°F T can be calculated by adding the adjusted RTNDT:

T = 43.0 + 87 = 130°F for P = 1105 psig For LaSalle Unit 2, the LPCI nozzle is the limiting material for the beltline region for 32 EFPY. The beltline pressure test P-T curves provided in Section 5.0 of this report are calculated in the same manner as the Feedwater Nozzle pressure test P-T curves as described in Section 4.3.2.1.3. The initial RTNDT for the LPCI nozzle materials is -6 0 F as shown in Table 4-2. The generic pressure test P-T curve is applied to the LaSalle Unit 2 Feedwater Nozzle curve by shifting the P vs. (T - RTNDT) values in Section 4.3.2.1.3 to reflect the ART value of 52 0 F. The 20 EFPY beltline pressure test

-43 -

GE Nuclear Energy GE-NE-0000-0003-5526-01a P-T curves are non-beitline limited and the beltline material calculations are performed as described in this section.

4.3.2.2.3 Beltline Region - Core Not CriticalHeatup/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 XI Appendix G [6]:

KIc = 2.0- KIm +K1 t (4-13) where Kim is primary membrane Kdue to pressure and KIt is radial thermal gradient K due to heatup/cooldown.

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

The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient 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 ATf is based on one-dimensional heat conduction through an insulated flat plate:

a 2T(x,t) / a x2 = 1 / p (aT(x,t) / 8t) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and 13 is the thermal diffusivity.

GE Nuclear Energy GE-NE-0000-0003-5526-01 a The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that 8l(x,t) / at = dT(t) / dt = G, where G is the coolant heatup/cooldown rate, normally 100OF/hr. The differential equation is integrated over x for the following boundary conditions:

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

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

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

T = Gx 2 / 20 - GCx / 3 + To (4-15)

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

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

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

Bulletin 175 [15] for infinitely long cracks of 1/4T and 1/8T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.

For LaSalle Unit 2, the LPCI nozzle is the limiting material for the beltline region for 32 EFPY. The beltline core not critical P-T curves provided in Section 5.0 of this report are calculated in the same manner as the Feedwater Nozzle core not critical P-T curves as described in Section 4.3.2.1.4. The initial RTNDT for the LPCI nozzle materials is -6°F as shown in Table 4-2. The generic core not critical P-T curve is applied to the LaSalle Unit 2 Feedwater Nozzle curve by shifting the P vs. (T - RTNDT) values in Section 4.3.2.1.4 to reflect the ART value of 52°F. The 20 EFPY beltline core not critical P-T curves are non-beltline limited and the beltline material calculations are performed as described in this section.

-45 -

GE Nuclear Energy GE-NE-0000-0003-5526-01a 4.3.2.2.4 Calculationsfor the Beltine Region Core Not Critical Heatup/Cooldown This sample calculation is for a pressure of 1105 psig for 32 EFPY. The core not critical heatup/cooldown curve at 1105 psig uses the same Kim as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. Inaddition, there is a K,term for the thermal stress.

The additional inputs used to calculate Kit are:

Coolant heatup/cooldown rate, normally 100°F/hr G = 100 °F/hr Minimum vessel thickness, including clad thickness C = 0.552 ft (6.625 inches)

(the maximum vessel thickness is conservatively used)

Thermal diffusivity at 550°F (most conservative value) 13 = 0.354 ftl/ 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 / 21 (4-16)

= 100. (0.552)/ (2 0.354) = 43°F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.30) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kit = Mt- AT = 12.9, can be calculated. Kim has the same value as that calculated in Section 4.3.2.2.2.

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

(T - RTNDT) = ln[((2* Kim + Kit)- 33.2)/20.734]/0.2 (4-17)

= In[(2

• 53.1 + 12.9- 33.2) / 20.734] /0.02

= 71.1 OF GE Nuclear Energy GE-N E-0000-0003-5526-01 a T can be calculated by adding the adjusted RTNDT:

T=71.1+87=158.1°F forP= 1105psig 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. 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 LaSalle Unit 2 at low pressures.

The approach used for LaSalle Unit 2 for the bolt-up temperature was based on a 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 60°F adder, and 2) inclusion of the additional 60 0 F requirement above the RTNDT provides the additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Tables 4-1 and 4-2, the limiting initial RTNDT for the closure flange region is represented by both the top head and vessel shell flange materials at 26°F, and the LST of the closure studs is 70°F; therefore, the bolt-up temperature value used is 86°F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

10CFR50 Appendix G, paragraph IV.A.2 [8] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 900F) and Curve B temperature no less than (RTNDT + 1200 F).

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 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°F for the reason discussed below.

The shutdown margin, provided in the LaSalle Unit 2 Technical Specification, is calculated for a water temperature of 68°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 86°F limit for the upper vessel and beltline region and the 68°F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel.

When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

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 I of [8] requires that core critical P-T limits be 40°F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40°F for pressures above 312 psig.

Table I of 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 86°F, based on an RTNDT of 26°F. In GE Nuclear Energy GE-NE-0000-0003-5526-01 a 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 1105 psig). The requirement of closure region RTNDT + 160°F does cause a temperature shift in Curve C at 312 psig.

GE Nuclear Energy GE-NE-0000-0003-5526-01 a

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) 0 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 IOO0 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 20F/hr or less must be maintained at all times.

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

GE Nuclear Energy GE-NE-0000-0003-5526-01a The following P-T curves were generated for LaSalle Unit 2.

"* Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition.

" Separate P-T curves were developed for the upper vessel, beltline (at 20 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

" A composite P-T curve was also generated for the Core Critical condition at 20 and 32 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.

Using the flux from Reference 14 the P-T curves are not beltline limited through 1400 psig for curve A and curve B for 20 EFPY. The P-T curves are beltline (LPCI nozzle) limited above 760 psig for curve A and 550 psig for curve B for 32 EFPY.

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

GE Nuclear Energy GE-NE-0000-0003-5526-01 a Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure Table Numbers Curve Curve Description Numbers for for Presentation of Presentation of

_the P-T Curves the P-T Curves Curve A Bottom Head Limits (CRD Nozzle) Figure 5-1 B-1 & B-3 Upper Vessel Limits (FW Nozzle) Figure 5-2 B-1 & B-3 Beltline Limits for 20 EFPY Figure 5-3 B-3 Beltline Limits for 32 EFPY Figure 5-4 B-1 Curve B I Bottom Head Limits (CRD Nozzle) Figure 5-5 B-1 & B-3 Upper Vessel Limits (FW Nozzle) Figure 5-6 B-1 & B-3 Beitline Limits for 20 EFPY Figure 5-7 B-3 Beltline Limits for 32 EFPY Figure 5-8 B-1 Curve C Composite Curve for 20 EFPY** Figure 5-9 B-4 A, B, & C Composite Curves for 32 EFPY Bottom Head and Composite Curve A Figure 5-10 B-2 for 32 EFPY*

Bottom Head and Composite Curve B Figure 5-11 B-2 for 32 EFPY*

Composite Curve C for 32 EFPY** Figure 5-12 B-2

• 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 GE-NE-0000-0003-5526-01 a 1400 1300 1200 1100 1000 z

9. 900 0 iINITIAL RTndt VALUE ISI 49LF FOR BOTOM HEADJ 0 800 0700 HEATUP/COOLDOWN RATE OF COOLANT

< 20"F/HR

• 600

• 500 0

ILl 300 200 100 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]

[20 °F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0003-5526-O1a 1400 1300 1200 1100 1000 0 9800 ' 'IINITIAL RTndt VALUE IS I

_______40_F FOR UPPER VESSEL w

U) 800 U)

U) 0 700 HEATUP/COOLDOWN Uj RATE OF COOLANT S600 ----- 20"F/HR S500 f 400 300 200 -FLANGE REGION 16F

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

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

[20 0 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0003-5526-01a 1400 1300 1200 INITIAL RTndt VALUE IS 52°F FOR BELTLINE 1100 C.

a 1000 R

-I 900

-I BELTLINE CURVE to 800 ADJUSTED AS SHOWN:

EFPY SHIFT (F) 20 25 700 U) HEATUP/COOLDOWN 600 RATE OF COOLANT

< 200F/HR

,I 500 400 uJ 300 200

"-BELTLINELIMITS 100 0

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

(*F)

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

[201F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-O000-0003-5526-01 a 1400 1300 1200 INITIAL RTndt VALUE IS

-6°F FOR LPCI NOZZLE 1100 Note: The LPCI Nozzle is the limiting material for the beltline region.

IL 900 BELTLINE CURVE ADJUSTED AS SHOWN:

10 EFPY SHIFT (°F) o)

U 700 32 58 Uj I-1 HEATUP/COOLDOWN 1% 600

-I RATE OF COOLANT Z < 20F/IHR

_j 500 w

~400 w

300 200 -BELTLINE LIMITS 100 0

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

(*F)

Figure 5-4: Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY

[20°F/hr or less coolant heatup/cooldown]

-56 -

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 1400 1300 1200 1100 0.. 1000 a.

$A 900 INITIAL RTndt VALUE IS I 49°F FOR BOTTOM HEAD w

0 800 700 600 500 w

0) 400 300 200 100 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]

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

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 1400 1300 1200 1100 1000 ILu x

"IL 900 0

I. 40"F FOR UPPER VESSELI

.1 I INITIAL RTndt VALUE IS I

.Il "U) 0 LU 800 w o700 HEATUP/COOLDOWN RATE OF COOLANT I < 100°FIHR

, 600 500 F I.

LU CVn400 (n

300 3712 PSI G ___

I a 200

!~ *-UPPER VESSEL I i LIMITS (Including 100 I FLANGE REGION 86°F Flange and FW I Nozzle Umits) 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]

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

GE Nuclear Energy GE-N E-0000-0003-5526-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 52°F FOR BELTLINE 1100

• F a 1000 BELTLINE CURVE 9(

z ADJUSTED AS SHOWN:

I

-j

="900 I j W EFPY 20 SHIFT (°F) 25 C

1,1 800 O 700 I_ _

HEATUP/COOLDOWN 31 SIG RATE OF COOLANT

• 600 < 100°FIHR

_ _ _ _ I, _

-500 LU 1QF5

________J

___ _ __ILU uo 400 U) 3U l 86F 300 200

-BELTLINE LIMITS 100 0

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

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

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

GE Nuclear Energy GE-NE-0000-0003-5526-01a 1400 1300 I I 1200 T INITIAL RTndt VALUE IS II I-6=F FOR LPCI NOZZLE 1100 + [ Note: The LPCI Nozzle is the limiting material for the beitline region.

1000 a.

0 900 , t i BELTLINE CURVE ADJUSTED AS SHOWN:

0 EFPY SHIFT (-F) 800 32 58 700 r I-l w Ii HEATUP/COOLDOWN RATE OF COOLANT u'J < 100°F/HR 600 I-l 500 _ - _

400 _ _ - - -

0, a.

300 _ j-31--S 200 "rl

,RS0 - BELTLINE LIMITS 100 T F F 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(*F)

Figure 5-8: Beltline P-T Curves for Core Not Critical [Curve B] up to 32 EFPY

[1OO0 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 1400 INITIAL RTndt VALUES ARE 1300 52°F FOR BELTLINE, 40°F FOR UPPER VESSEL, 1200 AND 49°F FOR BOTTOM HEAD 1100

=31000 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

0. 900 20 25 I 800 HEATUP/COOLDOWN RATE OF COOLANT

_<100°F/HR 600 w

300 600 2

SBELTLINE AND 100 NON-BELTLINE LIMITS 0

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

(°F)

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

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

GE Nuclear Energy GE-NE-0000-0003-5526-01a 1400 1300 1200 INITIAL RTndt VALUES ARE

-6°F FOR LPCI NOZZLE, 1100 40"F FOR UPPER VESSEL, AND 49°F FOR BOTTOM HEAD S1000 Note: The LPCI Nozzle is the limiting material for the 0 900 beitline region.

800 I

V-.

0O 700 BELTLINE CURVES EOS600 ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

S500 32 58 w

o 400 HEATUP/COOLDOWN U RATE OF COOLANT

< 20F/HR 300

-UPPER VESSEL 200 AND BELTLINE LIMITS

...... BOTTOM HEAD 100 CURVE 0

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

Figure 5-10: Composite Pressure Test P-T Curves [Curve A] up to 32 EFPY

[20 °F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 1400 -. -

1300 -~ _ ."_':_

1200 r INITIAL RTndt VALUES I/ ARE

_ "-6°FFOR LPCI NOZZLE, 1100 t 409F FOR UPPER S VESSEL,

_ ____ __ __ I___

____I IAND "01000 49F FOR BOTTOM HEAD

-0 i Note: The LPCI Nozzle is S900 - tthe limiting material for the beltline region.

cao 80 0 _ __

0 700 _ BELTLINE CURVES i IADJUSTED AS SHOWN:

oi[ IEFPY SHIFT (F) a: 600 .." I 32 58

-j0 HEAD 68T HEATUP/COOLDOWN RATE OF COOLANT 1a_400 _

100°F/HR 300 200 [ FAND

._UPPER LIMITS VESSEL BELTLINE 100 __ REGION ...... BOTTOM HEAD SIO CURVE 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-11: Composite Core Not Critical P-T Curves [Curve B] up to 32 EFPY

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

GE Nuclear Energy GE-NE-O000-0003-5526-01 a 1400 1300 INITIAL RTndt VALUES ARE

-6°F FOR LPCI NOZZLE, 1200 40"F FOR UPPER VESSEL, AND 1100 49"F FOR BOTTOM HEAD "1000 Note: The LPCI Nozzle is 4 the limiting material for the uJ beltline region.

M. 900 0

iI

-J

'f 800 o 700 w

BELTLINE CURVE ADJUSTED AS SHOWN:

4 EFPY SHIFT (°F)

Uj 5 00 32 58 w

S4500 HEATUP/COOLDOWN RATE OF COOLANT

u. < 100°F/HR a.

300 200 I--BELTLINED NON-BELTLINE 100 LIMITS 0

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

(*F)

Figure 5-12: Composite Core Critical P-T Curves [Curve C] up to 32 EFPY

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

GE Nuclear Energy GE-N E-0000-0003-5526-01 a

6.0 REFERENCES

1. Carey, R.G., "Pressure-Temperature Curves for ComEd LaSalle Unit 2," GE-NE, San Jose, CA, May 2000, (GE-NE-B13-02057-00-05Rl, Revision 1)(GE Proprietary).

2. GE Drawing Number 761 E581, "Reactor Vessel Thermal Cycles," GE-NED, San Jose, CA, Revision 1 (GE Proprietary).

3. GE Drawing Number 15888136, "Reactor Vessel Nozzle Thermal Cycles,"

GE-NED, San Jose, CA, Revision 6 (GE Proprietary).

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

5. a) T.A. Caine, "LaSalle County Station Units 1 and 2 Fracture Toughness Analysis per 10CFR50 Appendix G", GE-NE, San Jose, CA, March 1988 (SASR 88-10).

9 b) E.W. Sleight, "LaSalle Unit 2 RPV surveillance Materials Testing and Analysis,"

GE-NE, San Jose, CA, February 1996, (GE-NE-B1301786-01, Revision 0).

6. "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section III or XI 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 GE-NE-0000-0003-5526-01 a

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 CMTR's:

LaSalle Unit 2 -QA Records & RPV CMTR's LaSalle Unit 2 GE PO# 205-AE020, Manufactured by CBIN.

13. a) Letter from L. Loflin (Shearon Harris Nuclear Power) to NRC dated September 8, 1989, transmitting BAW-2083, "Analysis of Capsule U, Carolina Power & Light Company, Shearon Harris Unit No. 1, Reactor Vessel Material Surveillance Program", August 1989.

b) "Carolina Power & Light Company, Shearon Harris Unit No. 1, Reactor Vessel Radiation Surveillance Program", WCAP-10502, May 1984.

c) Letter from R. M. Krich to the NRC, "Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity - Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 - LaSalle County Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374 - Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265," Commonwealth Edison Company, Downers Grove, IL., July 30, 1998.

GE Nuclear Energy GE-NE-0000-0003-5526-01a

14. a) Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation," GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)(GE Proprietary Information).

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

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

Welding Research Council Bulletin 175, August 1972.

16.

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

18.

19. Bottom Head and Feedwater Nozzle Dimensions:

a) CBIN Drawing, GE Number VPF 3073-1-7, "Vessel Outline," GE-APED, San Jose, CA, Revision 7.

b) GE Drawing Number VPF 3073-52, "Feedwater Nozzle", GE-NED, San Jose, CA, Revision 7.

20.

21. "Materials - Properties," Part D to Section IIof the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

-67 -

GE Nuclear Energy GE-NE-0000-0003-5526-01 a APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-NE-0000-0003-5526-01a A-2

GE Nuclear Energy GE-N E-0000-0003-5526-01 a 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 0F. Nozzles and appurtenances made from Alloy 600 (Inconel) do not require fracture toughness analysis. Components that do not require a fracture toughness evaluation are listed below:

Nozzle or Appurtenance Nozzle or Appurtenance Material Reference Remarks Identification Nl1 Core Differential Pressure & Alloy 600 Thickness is < 2.5" and made of Liquid Poison - Penetration Alloy 600; therefore, no further

< 2.5" fracture toughness evaluation Is required.

N15 Drain- Penetration < 2.5" - SA-508 Cl. 1 1.5.9 & The discontinuity of the CRD Bottom Head (Heat 265M-1) 1.5.21 nozzle listed in Table A-1 bounds this discontinuity; RTNDT=-8"F therefore, no further fracture toughness evaluation is required.

N17 Seal Leak Detection - Alloy 600 1.5.9 & Not a pressure boundary Penetration -1" 1.5.28 component; therefore, requires no fracture toughness evaluation.

Top Head Lifting Lugs SA-533 GR. B 1.5.9 & Not a pressure boundary CL. 1 1.5.14 component and loads only occur on this component when the reactor is shutdown during an outage. Therefore, no fracture toughness evaluation is required.

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

A-3

GE Nuclear Energy GE-NE-0000-0003-5526-01 a APPENDIX A

REFERENCES:

1.5. RPV Drawings 1.5.1. CBI #32, Rev. 5, "Top Head Assembly," (GE VPF # 3073-032, Rev. 5) 1.5.2. CBI #30, Rev. 3, "Top Head Flange Assembly," (GE VPF # 3073-030, Rev. 3) 1.5.3. CBI #26, Rev. 8, "Shell Flange Assembly w/ N17 Nozzle,"

(GE VPF # 3073-026, Rev. 9) 1.5.4. CBI #21, Rev. 2, "#1 Shell Ring Assembly," (GE VPF # 3073-021, Rev. 4) 1.5.5. CBI #22, Rev. 3, "#2 Shell Ring Assembly," (GE VPF # 3073-022, Rev. 4) 1.5.6. CBI #23, Rev. 2, "#3 Shell Ring Assembly," (GE VPF # 3073-023, Rev. 4) 1.5.7. CBI #24, Rev. 3, "#4 Shell Ring Assembly," (GE VPF # 3073-024, Rev. 4) 1.5.8. CBI #13, Rev. 5, "Bottom Head Assembly," (GE VPF # 3073-013, Rev. 6) 1.5.9. CBI #R1 3, Rev. 7, "Vessel, Nozzle & Outside Bracket As-Built Dimensions," (GE VPF # 3073-104, Rev. 8) 1.5.10. CBI #58, Rev. 5, "RHRPLPCI Mode Nozzle N6," (GE VPF # 3073-058, Rev. 5) 1.5.11. CBI #69, Rev. 4, "Instrumentation Nozzle N12," (GE VPF # 3073-069, Rev. 4) 1.5.12. CBI #19, Rev. 4, "Shroud Support Assembly," (GE VPF # 3073-019, Rev. 5) 1.5.13. CBI #17, Rev. 2, "Shroud Support Stubs," (GE VPF # 3073-017, Rev. 2) 1.5.14. CBI #40, Rev. 2, "Top Head Lift Lugs," (GE VPF # 3073-040, Rev. 3) 1.5.15. CBI #51, Rev. 8, "N3 Nozzle," (GE VPF # 3073-051, Rev. 8) 1.5.16. CBI #52, Rev. 7, "N4 Nozzle," (GE VPF # 3073-052, Rev. 7) 1.5.17. CBI #55, Rev. 6, "N5 Nozzle," (GE VPF # 3073-055, Rev. 7) 1.5.18. CBI #61, Rev. 2, "N7 Nozzle," (GE VPF # 3073-061, Rev. 3) 1.5.19. CBI #63, Rev. 5, "N9 Nozzle," (GE VPF # 3073-063, Rev. 5) 1.5.20. CBI #65, Rev. 85, "N10 Nozzle," (GE VPF # 3073-065, Rev. 5) 1.5.21. CBI #72, Rev. 4, "N1 5 Nozzle," (GE VPF # 3073-072, Rev. 4) 1.5.22. CBI #51, Rev. 8, "N3 Nozzle," (GE VPF # 3073-051, Rev. 8) 1.5.23. CBI #73, Rev. 5, "N16 Nozzle," (GE VPF # 3073-073, Rev. 5) 1.5.24. CBI #76, Rev. 2, "N18 Nozzle," (GE VPF # 3073-076, Rev. 3)

A-4

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 1.5.25. CBI #80, Rev. 2, "Stabilizer Brackets,' (GE VPF# 3073-080, Rev. 2) 1.5.26. CBI #9, Rev. 7, "Support Skirt Knuckle," (GE VPF # 3073-009, Rev. 7) 1.5.27. CBI #62, Rev. 2, "N8 Nozzle,' (GE VPF # 3073-062, Rev. 3) 1.5.28. GE Drawing 732E143, Rev. 16, "Purchase Part, Reactor Vessel,"

GE-NED, San Jose, CA.

1.5.29. CBI #46, Rev. 5, "Ni Nozzle," (GE VPF # 3073-046, Rev. 5) 1.5.30. CBI #48, Rev. 6, "N2 Nozzle," (GE VPF # 3073-048, Rev. 6) 1.6. Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation', GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)

(GE Proprietary).

A-5

GE Nuclear Energy GE-NE-0000-0003-5526-01a APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTIINE PRESSURE CURVE A CURVEA CURVEA CURVEB CURVE B CURVE B (PSIG) (OF) (OF) (F) (OF) (F) (F) 0 68.0 86.0 86.0 68.0 86.0 86.0 10 68.0 86.0 86.0 68.0 86.0 86.0 20 68.0 86.0 86.0 68.0 86.0 86.0 30 68.0 86.0 86.0 68.0 86.0 86.0 40 68.0 86.0 86.0 68.0 86.0 86.0 50 68.0 86.0 86.0 68.0 86.0 86.0 60 68.0 86.0 86.0 68.0 86.0 86.0 70 68.0 86.0 86.0 68.0 86.0 86.0 80 68.0 86.0 86.0 68.0 86.0 86.0 90 68.0 86.0 86.0 68.0 86.0 86.0 100 68.0 86.0 86.0 68.0 86.0 86.0 110 68.0 86.0 86.0 68.0 86.0 86.0 120 68.0 86.0 86.0 68.0 86.0 86.0 130 68.0 86.0 86.0 68.0 86.0 86.2 140 68.0 86.0 86.0 68.0 86.0 89.4 150 68.0 86.0 86.0 68.0 86.0 92.2 160 68.0 86.0 86.0 68.0 86.0 94.9 170 68.0 86.0 86.0 68.0 86.0 97.5 180 68.0 86.0 86.0 68.0 87.9 99.9 190 68.0 86.0 86.0 68.0 90.2 102.2 200 68.0 86.0 86.0 68.0 92.3 104.3 210 68.0 86.0 86.0 68.0 94.3 106.3 220 68.0 86.0 86.0 68.0 96.3 108.3 230 68.0 86.0 86.0 68.0 98.1 110.1 240 68.0 86.0 86.0 68.0 99.9 111.9 B-2

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (°F) (°F) (OF) (°F) (OF) (OF) 250 68.0 86.0 86.0 68.0 101.6 113.6 260 68.0 86.0 86.0 68.0 103.2 115.2 270 68.0 86.0 86.0 68.0 104.8 116.8 280 68.0 86.0 86.0 68.0 106.3 118.3 290 68.0 86.0 86.0 68.0 107.8 119.8 300 68.0 86.0 86.0 68.0 109.2 121.2 310 68.0 86.0 86.0 68.0 110.5 122.5 312.5 68.0 86.0 86.0 68.0 110.9 122.9 312.5 68.0 116.0 116.0 68.0 146.0 146.0 320 68.0 116.0 116.0 68.0 '146.0 146.0 330 68.0 116.0 116.0 68.0 146.0 146.0 340 68.0 116.0 116.0 68.0 146.0 146.0 350 68.0 116.0 116.0 68.0 146.0 146.0 360 68.0 116.0 116.0 68.0 146.0 146.0 370 68.0 116.0 116.0 68.0 146.0 146.0 380 68.0 116.0 116.0 68.0 146.0 146.0 390 68.0 116.0 116.0 68.0 146.0 146.0 400 68.0 116.0 116.0 68.0 146.0 146.0 410 68.0 116.0 116.0 68.0 146.0 146.0 420 68.0 116.0 116.0 68.0 146.0 146.0 430 68.0 116.0 116.0 68.0 146.0 146.0 440 68.0 116.0 116.0 68.0 146.0 146.0 450 68.0 116.0 116.0 68.0 146.0 146.0 460 68.0 116.0 116.0 68.0 146.0 146.0 470 68.0 116.0 116.0 68.0 146.0 146.0 480 68.0 116.0 116.0 68.0 146.0 146.0 490 68.0 116.0 116.0 68.0 146.0 146.0 B-3

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (*F) (OF) (OF) (F) (OF) (OF) 500 68.0 116.0 116.0 68.0 146.0 146.0 510 68.0 116.0 116.0 68.0 146.0 146.0 520 68.0 116.0 116.0 68.2 146.0 146.0 530 68.0 116.0 116.0 70.2 146.0 146.0 540 68.0 116.0 116.0 72.1 146.0 146.0 550 68.0 116.0 116.0 73.9 146.0 146.6 560 68.0 116.0 116.0 75.7 146.0 147.4 570 68.0 116.0 116.0 77.4 146.0 148.1 580 68.0 116.0 116.0 79.0 146.0 148.9 590 68.0 116.0 116.0 80.6 146.0 149.6 600 68.0 116.0 116.0 82.2 146.0 150.1 610 68.0 116.0 116.0 83.7 146.0 150.6 620 68.0 116.0 116.0 85.1 146.0 151.0 630 68.0 116.0 116.0 86.5 146.0 151.4 640 68.0 116.0 116.0 87.9 146.0 151.8 650 68.0 116.0 116.0 89.2 146.0 152.2 660 68.0 116.0 116.0 90.5 146.0 152.7 670 68.0 116.0 116.0 91.8 146.0 153.1 680 68.0 116.0 116.0 93.1 146.0 153.5 690 68.0 116.0 116.0 94.3 146.0 153.9 700 69.2 116.0 116.0 95.4 146.0 154.3 710 70.7 116.0 116.0 96.6 146.0 154.7 720 72.1 116.0 116.0 97.7 146.0 155.1 730 73.5 116.0 116.0 98.8 146.0 155.5 740 74.8 116.0 116.0 99.9 146.0 155.9 750 76.1 116.0 116.0 101.0 146.0 156.2 760 77.4 116.0 116.8 102.0 146.0 156.6 B-4

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTIINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (*F) (OF) (-F) (OF) (*F) (*F) 770 78.6 116.0 117.6 103.0 146.0 157.0 780 79.8 116.0 118.3 104.0 146.0 157.4 790 81.0 116.0 119.1 105.0 146.0 157.8 800 82.2 116.0 119.9 105.9 146.1 158.1 810 83.3 116.0 120.6 106.9 146.5 158.5 820 84.4 116.0 121.4 107.8 146.9 158.9 830 85.5 116.0 122.1 108.7 147.2 159.2 840 86.5 116.0 122.8 109.6 147.6 159.6 850 87.6 116.0 123.5 110.4 147.9 159.9 860 88.6 116.0 124.2 111.3 148.3 160.3 870 89.6 116.0 124.9 112.1 148.6 160.6 880 90.5 116.0 125.6 113.0 149.0 161.0 890 91.5 116.0 126.3 113.8 149.3 161.3 900 92.4 116.0 126.9 114.6 149.7 161.7 910 93.4 116.0 127.6 115.4 150.0 162.0 920 94.3 116.2 128.2 116.1 150.4 162.4 930 95.1 116.9 128.9 116.9 150.7 162.7 940 96.0 117.5 129.5 117.7 151.0 163.0 950 96.9 118.1 130.1 118.4 151.4 163.4 960 97.7 118.7 130.7 119.1 151.7 163.7 970 98.6 119.3 131.3 119.9 152.0 164.0 980 99.4 119.9 131.9 120.6 152.4 164.4 990 100.2 120.5 132.5 121.3 152.7 164.7 1000 101.0 121.1 133.1 122.0 153.0 165.0 1010 101.7 121.7 133.7 122.6 153.3 165.3 1020 102.5 122.2 134.2 123.3 153.6 165.6 1030 103.3 122.8 134.8 124.0 154.0 166.0 B-5

GE Nuclear Energy GE-N E-0000-0003-5526-01 a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (°F) (F) (F) (OF) (OF) (F) 1040 104.0 123.4 135.4 124.6 154.3 166.3 1050 104.7 123.9 135.9 125.3 154.6 166.6 1060 105.4 124.5 136.5 125.9 154.9 166.9 1070 106.2 125.0 137.0 126.5 155.2 167.2 1080 106.9 125.5 137.5 127.2 155.5 167.5 1090 107.6 126.1 138.1 127.8 155.8 167.8 1100 108.2 126.6 138.6 128.4 156.1 168.1 1105 108.6 126.8 138.8 128.7 156.3 168.3 1110 108.9 127.1 139.1 129.0 156.4 168.4 1120 109.6 127.6 139.6 129.6 '156.7 168.7 1130 110.2 128.1 140.1 130.2 157.0 169.0 1140 110.9 128.6 140.6 130.7 157.3 169.3 1150 111.5 129.1 141.1 131.3 157.6 169.6 1160 112.1 129.6 141.6 131.9 157.9 169.9 1170 112.8 130.1 142.1 132.4 158.2 170.2 1180 113.4 130.6 142.6 133.0 158.5 170.5 1190 114.0 131.1 143.1 133.5 158.7 170.7 1200 114.6 131.5 143.5 134.1 159.0 171.0 1210 115.2 132.0 144.0 134.6 159.3 171.3 1220 115.8 132.5 144.5 135.2 159.6 171.6 1230 116.3 132.9 144.9 135.7 159.9 171.9 1240 116.9 133.4 145.4 136.2 160.2 172.2 1250 117.5 133.8 145.8 136.7 160.4 172.4 1260 118.0 134.3 146.3 137.2 160.7 172.7 1270 118.6 134.7 146.7 137.7 161.0 173.0 1280 119.1 135.2 147.2 138.2 161.2 173.2 1290 119.7 135.6 147.6 138.7 161.5 173.5 B-6

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-1. LaSalle Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM SUPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 1300 120.2 136.0 148.0 139.2 161.8 173.8 1310 120.7 136.5 148.5 139.7 162.1 174.1 1320 121.3 136.9 148.9 140.2 162.3 174.3 1330 121.8 137.3 149.3 140.6 162.6 174.6 1340 122.3 137.7 149.7 141.1 162.8 174.8 1350 122.8 138.1 150.1 141.6 163.1 175.1 1360 123.3 138.6 150.6 142.0 163.4 175.4 1370 123.8 139.0 151.0 142.5 163.6 175.6 1380 124.3 139.4 151.4 142.9 163.9 175.9 1390 124.8 139.8 151.8 143.4 164.1 176.1 1400 125.3 140.2 152.2 143.8 164.4 176.4 B-7

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (°F) (OF) (OF) (OF) (OF) 0 68.0 86.0 68.0 86.0 86.0 10 68.0 86.0 68.0 86.0 86.0 20 68.0 86.0 68.0 86.0 86.0 30 68.0 86.0 68.0 86.0 86.0 40 68.0 86.0 68.0 86.0 86.0 50 68.0 86.0 68.0 86.0 86.0 60 68.0 86.0 68.0 86.0 92.0 70 68.0 86.0 68.0 86.0 99.2 80 68.0 86.0 68.0 86.0 105.2 90 68.0 86.0 68.0 86.0 110.3 100 68.0 86.0 68.0 86.0 114.8 110 68.0 86.0 68.0 86.0 118.9 120 68.0 86.0 68.0 86.0 122.7 130 68.0 86.0 68.0 86.2 126.2 140 68.0 86.0 68.0 89.4 129.4 150 68.0 86.0 68.0 92.2 132.2 160 68.0 86.0 68.0 94.9 134.9 170 68.0 86.0 68.0 97.5 137.5 180 68.0 86.0 68.0 99.9 139.9 B-8

GE Nuclear Energy GE-N E-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (°F) (OF) (.F) (OF) (OF) 190 68.0 86.0 68.0 102.2 142.2 200 68.0 86.0 68.0 104.3 144.3 210 68.0 86.0 68.0 106.3 146.3 220 68.0 86.0 68.0 108.3 148.3 230 68.0 86.0 68.0 110.1 150.1 240 68.0 86.0 68.0 111.9 151.9 250 68.0 86.0 68.0 113.6 153.6 260 68.0 86.0 68.0 115.2 155.2 270 68.0 86.0 68.0 116.8 156.8 280 68.0 86.0 68.0 118.3 158.3 290 68.0 86.0 68.0 119.8 159.8 300 68.0 86.0 68.0 121.2 161.2 310 68.0 86.0 68.0 122.5 162.5 312.5 68.0 86.0 68.0 122.9 162.9 312.5 68.0 116.0 68.0 146.0 186.0 320 68.0 116.0 68.0 146.0 186.0 330 68.0 116.0 68.0 146.0 186.0 340 68.0 116.0 68.0 146.0 186.0 350 68.0 116.0 68.0 146.0 186.0 360 68.0 116.0 68.0 146.0 186.0 B-9

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) (OF) (OF) (OF) (OF) 370 68.0 116.0 68.0 146.0 186.0 380 68.0 116.0 68.0 146.0 186.0 390 68.0 116.0 68.0 146.0 186.0 400 68.0 116.0 68.0 146.0 186.0 410 68.0 116.0 68.0 146.0 186.0 420 68.0 116.0 68.0 146.0 186.0 430 68.0 116.0 68.0 146.0 186.0 440 68.0 116.0 68.0 146.0 186.0 450 68.0 116.0 68.0 146.0 186.0 460 68.0 116.0 68.0 146.0 186.0 470 68.0 116.0 68.0 146.0 186.0 480 68.0 116.0 68.0 146.0 186.0 490 68.0 116.0 68.0 146.0 186.0 500 68.0 116.0 68.0 146.0 186.0 510 68.0 116.0 68.0 146.0 186.0 520 68.0 116.0 68.2 146.0 186.0 530 68.0 116.0 70.2 146.0 186.0 540 68.0 116.0 72.1 146.0 186.0 550 68.0 116.0 73.9 146.6 186.6 560 68.0 116.0 75.7 147.4 187.4 B-10

GE Nuclear Energy GE-NE-O000-0003-5526-01a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) (°F) (OF) (°F) (°F) 570 68.0 116.0 77.4 148.1 188.1 580 68.0 116.0 79.0 148.9 188.9 590 68.0 116.0 80.6 149.6 189.6 600 68.0 116.0 82.2 150.1 190.1 610 68.0 116.0 83.7 150.6 190.6 620 68.0 116.0 85.1 151.0 191.0 630 68.0 116.0 86.5 151.4 191.4 640 68.0 116.0 87.9 151.8 191.8 650 68.0 116.0 89.2 152.2 192.2 660 68.0 116.0 90.5 152.7 192.7 670 68.0 116.0 91.8 153.1 193.1 680 68.0 116.0 93.1 153.5 193.5 690 68.0 116.0 94.3 153.9 193.9 700 69.2 116.0 95.4 154.3 194.3 710 70.7 116.0 96.6 154.7 194.7 720 72.1 116.0 97.7 155.1 195.1 730 73.5 116.0 98.8 155.5 195.5 740 74.8 116.0 99.9 155.9 195.9 750 76.1 116.0 101.0 156.2 196.2 760 77.4 116.8 102.0 156.6 196.6 B-11

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTL1NE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (°F) (°F) (7F) (OF) (OF) 770 78.6 117.6 103.0 157.0 197.0 780 79.8 118.3 104.0 157.4 197.4 790 81.0 119.1 105.0 157.8 197.8 800 82.2 119.9 105.9 158.1 198.1 810 83.3 120.6 106.9 158.5 198.5 820 84.4 121.4 107.8 158.9 198.9 830 85.5 122.1 108.7 159.2 199.2 840 86.5 122.8 109.6 159.6 199.6 850 87.6 123.5 110.4 159.9 199.9 860 88.6 124.2 111.3 160.3 200.3 870 89.6 124.9 112.1 160.6 200.6 880 90.5 125.6 113.0 161.0 201.0 890 91.5 126.3 113.8 161.3 201.3 900 92.4 126.9 114.6 161.7 201.7 910 93.4 127.6 115.4 162.0 202.0 920 94.3 128.2 116.1 162.4 202.4 930 95.1 128.9 116.9 162.7 202.7 940 96.0 129.5 117.7 163.0 203.0 950 96.9 130.1 118.4 163.4 203.4 960 97.7 130.7 119.1 163.7 203.7 B-12

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (°F) (OF) (OF) 970 98.6 131.3 119.9 164.0 204.0 980 99.4 131.9 120.6 164.4 204.4 990 100.2 132.5 121.3 164.7 204.7 1000 101.0 133.1 122.0 165.0 205.0 1010 101.7 133.7 122.6 165.3 205.3 1020 102.5 134.2 123.3 165.6 205.6 1030 103.3 134.8 124.0 166.0 206.0 1040 104.0 135.4 124.6 166.3 206.3 1050 104.7 135.9 125.3 166.6 206.6 1060 105.4 136.5 125.9 166.9 206.9 1070 106.2 137.0 126.5 167.2 207.2 1080 106.9 137.5 127.2 167.5 207.5 1090 107.6 138.1 127.8 167.8 207.8 1100 108.2 138.6 128.4 168.1 208.1 1105 108.6 138.8 128.7 168.3 208.3 1110 108.9 139.1 129.0 168.4 208.4 1120 109.6 139.6 129.6 168.7 208.7 1130 110.2 140.1 130.2 169.0 209.0 1140 110.9 140.6 130.7 169.3 209.3 1150 111.5 141.1 131.3 169.6 209.6 B-1 3

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (°F) 1160 112.1 141.6 131.9 169.9 209.9 1170 112.8 142.1 132.4 170.2 210.2 1180 113.4 142.6 133.0 170.5 210.5 1190 114.0 143.1 133.5 170.7 210.7 1200 114.6 143.5 134.1 171.0 211.0 1210 115.2 144.0 134.6 171.3 211.3 1220 115.8 144.5 135.2 171.6 211.6 1230 116.3 144.9 135.7 171.9 211.9 1240 116.9 145.4 136.2 172.2 212.2 1250 117.5 145.8 136.7 172.4 212.4 1260 118.0 146.3 137.2 172.7 212.7 1270 118.6 146.7 137.7 173.0 213.0 1280 119.1 147.2 138.2 173.2 213.2 1290 119.7 147.6 138.7 173.5 213.5 1300 120.2 148.0 139.2 173.8 213.8 1310 120.7 148.5 139.7 174.1 214.1 1320 121.3 148.9 140.2 174.3 214.3 1330 121.8 149.3 140.6 174.6 214.6 1340 122.3 149.7 141.1 174.8 214.8 1350 122.8 150.1 141.6 175.1 215.1 B-14

GE Nuclear Energy G E-N E-0000-0003-5526-01 a TABLE B-2. LaSalle Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-10, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT & BELTLINE 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (*F) (°F) (°F) (*F) (F) 1360 123.3 150.6 142.0 175.4 215.4 1370 123.8 151.0 142.5 175.6 215.6 1380 124.3 151.4 142.9 175.9 215.9 1390 124.8 151.8 143.4 176.1 216.1 1400 125.3 152.2 143.8 176.4 216.4 B-15

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELT.INE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (F) (OF) (OF) (OF) (OF) (.F) 0 68.0 86.0 86.0 68.0 86.0 86.0 10 68.0 86.0 86.0 68.0 86.0 86.0 20 68.0 86.0 86.0 68.0 86.0 86.0 30 68.0 86.0 86.0 68.0 86.0 86.0 40 68.0 86.0 86.0 68.0 86.0 86.0 50 68.0 86.0 86.0 68.0 86.0 86.0 60 68.0 86.0 86.0 68.0 86.0 86.0 70 68.0 86.0 86.0 68.0 86.0 86.0 80 68.0 86.0 86.0 68.0 86.0 86.0 90 68.0 86.0 86.0 68.0 86.0 86.0 100 68.0 86.0 86.0 68.0 86.0 86.0 110 68.0 86.0 86.0 68.0 86.0 86.0 120 68.0 86.0 86.0 68.0 86.0 86.0 130 68.0 86.0 86.0 68.0 86.0 86.0 140 68.0 86.0 86.0 68.0 86.0 86.0 150 68.0 86.0 86.0 68.0 86.0 86.0 160 68.0 86.0 86.0 68.0 86.0 86.0 170 68.0 86.0 86.0 68.0 86.0 86.0 180 68.0 86.0 86.0 68.0 87.9 86.0 190 68.0 86.0 86.0 68.0 90.2 87.2 200 68.0 86.0 86.0 68.0 92.3 89.3 210 68.0 86.0 86.0 68.0 94.3 91.3 220 68.0 86.0 86.0 68.0 96.3 93.3 230 68.0 86.0 86.0 68.0 98.1 95.1 240 68.0 86.0 86.0 68.0 99.9 96.9 B-16

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °Flhr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (F) ((F)

(F) (OF) (OF) 250 68.0 86.0 86.0 68.0 101.6 98.6 260 68.0 86.0 86.0 68.0 103.2 100.2 270 68.0 86.0 86.0 68.0 104.8 101.8 280 68.0 86.0 86.0 68.0 106.3 103.3 290 68.0 86.0 86.0 68.0 107.8 104.8 300 68.0 86.0 86.0 68.0 109.2 106.2 310 68.0 86.0 86.0 68.0 110.5 107.5 312.5 68.0 86.0 86.0 68.0 110.9 107.9 312.5 68.0 116.0 116.0 68.0 146.0 146.0 320 68.0 116.0 116.0 68.0 146.0 146.0 330 68.0 116.0 116.0 68.0 146.0 146.0 340 68.0 116.0 116.0 68.0 146.0 146.0 350 68.0 116.0 116.0 68.0 146.0 146.0 360 68.0 116.0 116.0 68.0 146.0 146.0 370 68.0 116.0 116.0 68.0 146.0 146.0 380 68.0 116.0 116.0 68.0 146.0 146.0 390 68.0 116.0 116.0 68.0 146.0 146.0 400 68.0 116.0 116.0 68.0 146.0 146.0 410 68.0 116.0 116.0 68.0 146.0 146.0 420 68.0 116.0 116.0 68.0 146.0 146.0 430 68.0 116.0 116.0 68.0 146.0 146.0 440 68.0 116.0 116.0 68.0 146.0 146.0 450 68.0 116.0 116.0 68.0 146.0 146.0 460 68.0 116.0 116.0 68.0 146.0 146.0 470 68.0 116.0 116.0 68.0 146.0 146.0 480 68.0 116.0 116.0 68.0 146.0 146.0 490 68.0 116.0 116.0 68.0 146.0 146.0 B-1 7

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (F) (OF) (O) (OF) (OF) f(F) 500 68.0 116.0 116.0 68.0 146.0 146.0 510 68.0 116.0 116.0 68.0 146.0 146.0 520 68.0 116.0 116.0 68.2 146.0 146.0 530 68.0 116.0 116.0 70.2 146.0 146.0 540 68.0 116.0 116.0 72.1 146.0 146.0 550 68.0 116.0 116.0 73.9 146.0 146.0 560 68.0 116.0 116.0 75.7 146.0 146.0 570 68.0 116.0 116.0 77.4 146.0 146.0 580 68.0 116.0 116.0 79.0 146.0 146.0 590 68.0 116.0 116.0 80.6 146.0 146.0 600 68.0 116.0 116.0 82.2 146.0 146.0 610 68.0 116.0 116.0 83.7 146.0 146.0 620 68.0 116.0 116.0 85.1 146.0 146.0 630 68.0 116.0 116.0 86.5 146.0 146.0 640 68.0 116.0 116.0 87.9 146.0 146.0 650 68.0 116.0 116.0 89.2 146.0 146.0 660 68.0 116.0 116.0 90.5 146.0 146.0 670 68.0 116.0 116.0 91.8 146.0 146.0 680 68.0 116.0 116.0 93.1 146.0 146.0 690 68.0 116.0 116.0 94.3 146.0 146.0 700 69.2 116.0 116.0 95.4 146.0 146.0 710 70.7 116.0 116.0 96.6 146.0 146.0 720 72.1 116.0 116.0 97.7 146.0 146.0 730 73.5 116.0 116.0 98.8 146.0 146.0 740 74.8 116.0 116.0 99.9 146.0 146.0 750 76.1 116.0 116.0 101.0 146.0 146.0 760 77.4 116.0 116.0 102.0 146.0 146.0 B-1 8

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (F) (F) (F) (F) (OF) (OF) 770 78.6 116.0 116.0 103.0 146.0 146.0 780 79.8 116.0 116.0 104.0 146.0 146.0 790 81.0 116.0 116.0 105.0 146.0 146.0 800 82.2 116.0 116.0 105.9 146.1 146.0 810 83.3 116.0 116.0 106.9 146.5 146.0 820 84.4 116.0 116.0 107.8 146.9 146.0 830 85.5 116.0 116.0 108.7 147.2 146.0 840 86.5 116.0 116.0 109.6 147.6 146.0 850 87.6 116.0 116.0 110.4 147.9 146.0 860 88.6 116.0 116.0 111.3 148.3 146.0 870 89.6 116.0 116.0 112.1 148.6 146.0 880 90.5 116.0 116.0 113.0 149.0 146.0 890 91.5 116.0 116.0 113.8 149.3 146.3 900 92.4 116.0 116.0 114.6 149.7 146.7 910 93.4 116.0 116.0 115.4 150.0 147.0 920 94.3 116.2 116.0 116.1 150.4 147.4 930 95.1 116.9 116.0 116.9 150.7 147.7 940 96.0 117.5 116.0 117.7 151.0 148.0 950 96.9 118.1 116.0 118.4 151.4 148.4 960 97.7 118.7 116.0 119.1 151.7 148.7 970 98.6 119.3 116.3 119.9 152.0 149.0 980 99.4 119.9 116.9 120.6 152.4 149.4 990 100.2 120.5 117.5 121.3 152.7 149.7 1000 101.0 121.1 118.1 122.0 153.0 150.0 1010 101.7 121.7 118.7 122.6 153.3 150.3 1020 102.5 122.2 119.2 123.3 153.6 150.6 1030 103.3 122.8 119.8 124.0 154.0 151.0 B-1 9

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (F) (F) ("F) (F) (F) (OF) 1040 104.0 123.4 120.4 124.6 154.3 151.3 1050 104.7 123.9 120.9 125.3 154.6 151.6 1060 105.4 124.5 121.5 125.9 154.9 151.9 1070 106.2 125.0 122.0 126.5 155.2 152.2 1080 106.9 125.5 122.5 127.2 155.5 152.5 1090 107.6 126.1 123.1 127.8 155.8 152.8 1100 108.2 126.6 123.6 128.4 156.1 153.1 1105 108.6 126.8 123.8 128.7 156.3 153.3 1110 108.9 127.1 124.1 129.0 156.4 153.4 1120 109.6 127.6 124.6 129.6 156.7 153.7 1130 110.2 128.1 125.1 130.2 157.0 154.0 1140 110.9 128.6 125.6 130.7 157.3 154.3 1150 111.5 129.1 126.1 131.3 157.6 154.6 1160 112.1 129.6 126.6 131.9 157.9 154.9 1170 112.8 130.1 127.1 132.4 158.2 155.2 1180 113.4 130.6 127.6 133.0 158.5 155.5 1190 114.0 131.1 128.1 133.5 158.7 155.7 1200 114.6 131.5 128.5 134.1 159.0 156.0 1210 115.2 132.0 129.0 134.6 159.3 156.3 1220 115.8 132.5 129.5 135.2 159.6 156.6 1230 116.3 132.9 129.9 135.7 159.9 156.9 1240 116.9 133.4 130.4 136.2 160.2 157.2 1250 117.5 133.8 130.8 136.7 160.4 157.4 1260 118.0 134.3 131.3 137.2 160.7 157.7 1270 118.6 134.7 131.7 137.7 161.0 158.0 1280 119.1 135.2 132.2 138.2 161.2 158.2 1290 119.7 135.6 132.6 138.7 161.5 158.5 B-20

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-3. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6, & 5-7 BOTTOM UPPER 20 EFPY BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTIINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVEB CURVE B (PSIG) (F) ("F) (F) (F) (F) (F) 1300 120.2 136.0 133.0 139.2 161.8 158.8 1310 120.7 136.5 133.5 139.7 162.1 159.1 1320 121.3 136.9 133.9 140.2 162.3 159.3 1330 121.8 137.3 134.3 140.6 162.6 159.6 1340 122.3 137.7 134.7 141.1 162.8 159.8 1350 122.8 138.1 135.2 141.6 163.1 160.1 1360 123.3 138.6 135.7 142.0 163.4 160.4 1370 123.8 139.0 136.2 142.5 163.6 160.8 1380 124.3 139.4 136.8 142.9 163.9 161.3 1390 124.8 139.8 137.3 143.4 164.1 161.7 1400 125.3 140.2 137.8 143.8 164.4 162.1 B-21

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER BOTTOM 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PSIG) (@F) (OF) (OF) 0 86.0 68.0 86.0 10 86.0 68.0 86.0 20 86.0 68.0 86.0 30 86.0 68.0 86.0 40 86.0 68.0 86.0 50 86.0 68.0 86.0 60 86.0 68.0 86.0 70 87.2 68.0 86.0 80 93.2 68.0 86.0 90 98.3 68.0 86.0 100 102.8 68.0 86.0 110 106.9 68.0 86.0 120 110.7 68.0 86.0 130 114.2 68.0 86.0 140 117.4 68.0 86.0 150 120.2 68.0 86.0 160 122.9 68.0 86.0 170 125.5 68.0 86.0 180 127.9 68.0 86.0 190 130.2 68.0 86.0 200 132.3 68.0 86.0 210 134.3 68.0 86.0 220 136.3 68.0 86.0 230 138.1 68.0 86.0 240 139.9 68.0 86.0 250 141.6 68.0 86.0 B-22

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °Flhr for Curve C For Figure 5-9 UPPER BOTTOM .20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PSIG) (OF) (O) (*F) 260 143.2 68.0 86.0 270 144.8 68.0 86.0 280 146.3 68.0 86.0 290 147.8 68.0 86.0 300 149.2 68.0 86.0 310 150.5 68.0 86.0 312.5 150.9 68.0 86.6 312.5 186.0 68.0 186.0 320 186.0 68.0 186.0 330 186.0 68.0 186.0 340 186.0 68.0 186.0 350 186.0 68.0 186.0 360 186.0 68.0 186.0 370 186.0 68.0 186.0 380 186.0 68.0 186.0 390 186.0 71.3 186.0 400 186.0 75.3 186.0 410 186.0 79.0 186.0 420 186.0 82.5 186.0 430 186.0 85.8 186.0 440 186.0 88.8 186.0 450 186.0 91.7 186.0 460 186.0 94.4 186.0 470 186.0 97.0 186.0 480 186.0 99.5 186.0 490 186.0 101.8 186.0 500 186.0 104.0 186.0 510 186.0 106.2 186.0 B-23

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER BOTTOM 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PSIG) ('F) (OF) (F) 520 186.0 108.2 186.0 530 186.0 110.2 186.0 540 186.0 112.1 186.0 550 186.0 113.9 186.0 560 186.0 115.7 186.0 570 186.0 117.4 186.0 580 186.0 119.0 186.0 590 186.0 120.6 186.0 600 186.0 122.2 186.0 610 186.0 123.7 186.0 620 186.0 125.1 186.0 630 186.0 126.5 186.0 640 186.0 127.9 186.0 650 186.0 129.2 186.0 660 186.0 130.5 186.0 670 186.0 131.8 186.0 680 186.0 133.1 186.0 690 186.0 134.3 186.0 700 186.0 135.4 186.0 710 186.0 136.6 186.0 720 186.0 137.7 186.0 730 186.0 138.8 186.0 740 186.0 139.9 186.0 750 186.0 141.0 186.0 760 186.0 142.0 186.0 770 186.0 143.0 186.0 780 186.0 144.0 186.0 790 186.0 145.0 186.0 B-24

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER BOTTOM 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PSIG) (°F) ("F) (F) 800 186.1 145.9 186.0 810 186.5 146.9 186.0 820 186.9 147.8 186.0 830 187.2 148.7 186.0 840 187.6 149.6 186.0 850 187.9 150.4 186.0 860 188.3 151.3 186.0 870 188.6 152.1 186.0 880 189.0 153.0 186.0 890 189.3 153.8 186.0 900 189.7 154.6 186.0 910 190.0 155.4 186.0 920 190.4 156.1 186.0 930 190.7 156.9 186.0 940 191.0 157.7 186.0 950 191.4 158.4 186.0 960 191.7 159.1 186.0 970 192.0 159.9 186.0 980 192.4 160.6 186.0 990 192.7 161.3 186.0 1000 193.0 162.0 186.0 1010 193.3 162.6 186.0 1020 193.6 163.3 186.0 1030 194.0 164.0 186.0 1040 194.3 164.6 186.0 1050 194.6 165.3 186.0 1060 194.9 165.9 186.0 1070 195.2 166.5 186.1 B-25

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER BOTTOM 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PSIG) (OF) (OF) (OF) 1080 195.5 167.2 186.7 1090 195.8 167.8 187.3 1100 196.1 168.4 187.8 1105 196.3 168.7 188.1 1110 196.4 169.0 188.4 1120 196.7 169.6 188.9 1130 197.0 170.2 189.4 1140 197.3 170.7 190.0 1150 197.6 171.3 190.5 1160 197.9 171.9 191.0 1170 198.2 172.4 191.5 1180 198.5 173.0 192.0 1190 198.7 173.5 192.5 1200 199.0 174.1 193.0 1210 199.3 174.6 193.5 1220 199.6 175.2 194.0 1230 199.9 175.7 194.5 1240 200.2 176.2 195.0 1250 200.4 176.7 195.5 1260 200.7 177.2 195.9 1270 201.0 177.7 196.4 1280 201.2 178.2 196.9 1290 201.5 178.7 197.3 1300 201.8 179.2 197.8 1310 202.1 179.7 198.2 1320 202.3 180.2 198.7 1330 202.6 180.6 199.1 1340 202.8 181.1 199.5 B-26

GE Nuclear Energy GE-NE-0000-0003-5526-01a TABLE B-4. LaSalle Unit 2 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER BOTTOM 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PSIG) (F) (F) (*F) 1350 203.1 181.6 200.0 1360 203.4 182.0 200.4 1370 203.6 182.5 200.8 1380 203.9 182.9 201.3 1390 204.1 183.4 201.7 1400 204.4 183.8 202.1 B-27

GE Nuclear Energy GE-NE-0000-0003-5526-01 a APPENDIX C Operating And Temperature Monitoring Requirements C-1

GE Nuclear Energy GE-NE-0000-0003-5526-01 a C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

One condition on monitoring the bottom head separately is that it must be demonstrated that the vessel beltline temperature can be accurately monitored during pressure testing. An experiment has been conducted at a BWR-4 that showed that thermocouples on the vessel near the feedwater nozzles, or temperature measurements of water in the recirculation loops provide good estimates of the beltline temperature during pressure testing. Thermocouples on the RPV flange to shell junction outside surface should be used to monitor compliance with upper vessel curve. Thermocouples on the bottom head outside surface should be used to monitor compliance with bottom head curves. A description of these measurements is given in GE SIL 430, attached in Appendix D. First, however, it should be determined whether there are significant temperature differences between the beltline region and the bottom head region.

C.2 DETERMINING WHICH CURVE TO FOLLOW The following subsections outline the criteria needed for determining which curve is governing during different situations. The application of the P-T curves and some of the assumptions inherent in the curves to plant operation is dependent on the proper monitoring of vessel temperatures.

C-2

GE Nuclear Energy GE-NE-0000-0003-5526-O1a C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by .- 0°F per hour. If the coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.

C.2.2 Curve B: Non-Nuclear Heatup/Cooldown Curve B should be used whenever Curve A or Curve C do not apply. In other words, the operator must follow this curve during times when the coolant is heating or cooling faster than 20°F 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 boltup, leakage testing and startup/shutdown operations, where operator actions can directly influence vessel pressures and temperatures.

The most severe unplanned transients relative to the P-T curves are those that result from SCRAMs, which sometimes include recirculation pump trips. Depending on operator responses following pump trip, there can be cases where stratification of colder water in the bottom head occurs while the vessel pressure is still relatively high.

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

C-3

GE Nuclear Energy GE-NE-0000-0003-5526-01a In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:

"* Head flange boltup

"* Leakage test (Curve A compliance) .,

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

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

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

C-4

GE Nuclear Energy GE-NE-0000-0003-5526-01a APPENDIX D GE SIL 430 D-1

GE Nuclear Energy GE-NE-0000-0003-5526-01 a September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions -. A 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°F for Tech steam pressure to from main steam instrument Spec 1OOOF/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 212°F for Tech Must comply with SEL 251 Spec 100°F/hr heatup to avoid vessel stratification.

and cooldown rate.

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

D-2

GE Nuclear Energy GE-NE-0000-0003-5526-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

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

RHR heat exchanger Alternate measurement Must have previously inlet coolant for Tech Spec IOOOF/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 100°F 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 GE-NE-0000-0003-5526-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Closure head flanges Primary measurement for Use for metal (not coolant) outside surface T/Cs BWR/6s to comply with temperature. Install Tech Spec brittle fracture temporary T/Cs for metal temperature limit alternate measurement, if for head boltup. 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 boltup.

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 GE-NE-O000-0003-5526-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

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

metal temperature (see SIL No. 251).

limit for hydro test.

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

during heatup.

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

D-5

GE Nuclear Energy GE-NE-0000-0003-5526-01a Product

Reference:

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

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

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

D-6

GE Nuclear Energy GE-NE-0000-0003-5526-01 a APPENDIX E Determination of Beitline Region and Impact on Fracture Toughness E-1

GE Nuclear Energy GE-NE-0000-0003-5526-01 a 10CFR50, Appendix G defines the beltline region of the reactor vessel as follows:

"T he 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.Oe17 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/cm2 , then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.

The following dimensions are obtained from the referenced drawings:

Shell # 2 - Top of Active Fuel (TAF): 366.31" (from vessel 0) (Reference 1)

Shell # 1 - Bottom of Active Fuel (BAF): 216.31" (from vessel 0) (Reference 1)

Bottom of LPCI Nozzle in Shell # 2: 355.06" (from vessel 0) (Reference 2)

Center line of LPCI Nozzle in Shell # 2: 372" (from vessel 0) (Reference 3)

Top of Recirculation Outlet Nozzle in Shell # 1:197.91" (from vessel 0) (Reference 4)

Center line of Recirculation Outlet Nozzle in Shell # 1: 172.5" (from vessel 0) (Reference 3)

Top of Recirculation Inlet Nozzle in Shell # 1: 198.56" (from vessel 0) (Reference 5)

Center line of Recirculation Inlet Nozzle in Shell # 1: 181" (from vessel 0) (Reference 3)

As shown above, the LPCI nozzle is within the core beitline region. This nozzle is bounded by the feedwater pressure-temperature curve as stated in Appendix A.

From [3], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -18" from BAF and the top of the recirculation outlet nozzle is -18" from BAF), and no other nozzles are within the BAF-TAF region of the reactor vessel. Therefore, if it can be shown that the peak E-2

GE Nuclear Energy GE-NE-0000-0003-5526-01a fluence at this location is less than I.0E17 n/cm 2, it can be safely concluded that all nozzles are outside the beltline region of the reactor vessel.

Based on the axial flux profile [6], the RPV flux level at -10" below the BAF dropped to less than 0.1 of the peak flux level at the same radius. Likewise, the RPV flux level at

-10" above the TAF dropped to less than 0.1 of the peak flux at the same radius.

Therefore, if the RPV fluence is 1.09E18 n/cm2 [6], fluence at -10" below BAF and -10" above TAF are expected to be less than 1.OE17 n/cm 2 at 32 EFPY. The beltline region considered in the development of the P-T curves is adjusted to include the additional 10" above and below the active fuel region. The adjusted beltline region extends from 206.31" to 376.31" above reactor vessel "0".

Based on the above, it is concluded that none of the LaSalle Unit 2 reactor vessel nozzles, other than the LPCI nozzle, which is considered in the P-T curve evaluation, are in the beltline region.

E-3

GE Nuclear Energy GE-NE-0000-0003-5526-01 a Appendix E

References:

1. Source of Bottom of Beitline Elevation: Figure Q121.7-2, uWelds in Beltline Region of Reactor Vessel - Unit 2", page Q121.7-12.

2. CBIN Drawing #58, Revision 5, "RHR/LPCI Nozzle N6",

(GE VPF #3073-58-5).

3. CBIN Drawing #R13, Revision 3, "Vessel, Nozzle, & Outside Bracket As Built Dimensions" (GE VPF #3073-104-8).

4. CBIN Drawing #46, Revision 5, "Recirculation Outlet Nozzle NI",

(GE VPF #3073-46-5).

5. CBIN Drawing #48, Revision 6, "Recirculation Inlet Nozzle N2",

(GE VPF #3073-48-6).

6. Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation", GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Rev. 0)(GE Proprietary Information).

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GE Nuclear Energy GE-NE-0000-0003-5526-01a APPENDIX F EVALUATION FOR UPPER SHELF ENERGY (USE)

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GE Nuclear Energy GE-NE-0000-0003-5526-01a Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy (USE) of the beltline materials. The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be up to 32 EFPY. Calculations of 32 EFPY USE, using Reg.

Guide 1.99, Rev. 2 [2] methods, are summarized in Table F-I.

The USE decrease prediction values from Reg. Guide 1.99, Rev. 2 [2] were used for the beltline plates and welds in Table F-I. These calculations are based on the peak 1/4T fluence for all materials other than the LPCI nozzle, for conservatism. Because the Charpy data available for the LPCI nozzle consists of shear energy of 60%, this conservatism is not applied to the 32 EFPY USE calculation for this component; the 1/4T fluence for the LPCI nozzle as provided in Table 4-4 is used. Based on these results, the beltline materials will have USE values above 50 ft-lb at 32 EFPY, as required in 10CFR50 Appendix G [1]. The lowest USE predicted for 32 EFPY is 53 ft-lb (for Lower Shell plate heat C9434-2).

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GE Nuclear Energy GE-N E-0000-0003-5526-01 a Table F-I: Upper Shelf Energy Evaluation for LaSalle Unit 2 Beltline Materials Initial Initial 32 EFPY Test Longitudinal Transverse I11T  % Decrease 32 EFPY Location Heat Temperature USE USE* %Cu Fluence USEb Transverse USE6 (OF) (ft4b) (ft4b) (n/cm2 ) (ft4b)

Plates:

Lower C9425-1 d 102 66.3 0.12 7.,E+17 12 58 C9425-2 d 94 61.1 0.12 7.5E+17 12 54 C9434-2 40 91 59.2 0.09 7.5E+17 10 53 Lower Intermediate C9481-10 40 n/a 95.5 0.11 7.5E+17 11 85 C9404-2 d 116 75.4 0.07 7.5E+17 8.5 69 eC960-2 40 107 69.6 0.12 7.5E+17 12 61 Welds:

Vertical:

Intermediate 3P4000 10 n/a 99 0.02 7.5E+17 8 91 Lower 3P4966 10 n/a 84 0.026 7.5E+17 8.5 77 Girth:

Lower to I IIIII Lower-I Intermediate 5P6771 10 We 61 0.04 7.5E+17 10 55 Nozzles:

LPCI 02a36W' -10 66 0.22 1.8E+17 12.5 58 a Values obtained from [3]

b Values obtained from Figure 2 of [2] for 32 EFPY 1/4T fluence c 32 EFPY Transverse USE - Initial Transverse USE °[1 - (%Decrease USE /100)]

d USE values estimated from statistical evaluation inAppendix B of [3]

e Initial Transverse USE value obtained from baseline transverse data set [4]

f Average of Charpy V-Notch data for %Shear = 60 F-3

GE Nuclear Energy GE-NE-0000-0003-5526-01a 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. T.A. Caine, "Upper Shelf Energy Evaluation for LaSalle Units 1 and 2", GENE, San Jose, CA, June 1990 (GE Report SASR 90-07).

4. Letter, dated 3/16/94, G.W. Contreras (GE San Jose) to R. Willems (Oak Brook),

"LaSalle RPV Archive Material Records Search".

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