ML101130371

GE-NE-0000-0003-5526-02R1a, Rev, 1, Pressure-Temperature Curves for Exelon LaSalle Unit 1.

ML101130371
Person / Time
Site:	LaSalle
Issue date:	05/31/2004
From:	Tilly L General Electric Co
To:	Office of Nuclear Reactor Regulation
References
DRF 0000-0028-1044, Rev 1, RS-10-080 GE-NE-0000-0003-5526-02R1a
Download: ML101130371 (164)

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ATTACHMENT 6 GE-NE-0000-0003-5526-02R1a, "Pressure-Temperature Curves For Exelon LaSalle Unit 1," dated May 2004 (Non-Proprietary)

GE Nuclear Energy Engineering and Technology GE-N E-0000-0003-5526-02R 1 a General Electric Company DRF 0000-0028-1044 175 Curtner Avenue Revision 1 San Jose, CA 95125 Class I May 2004 Pressure-Temperature Curves For Exelon LaSalle Unit I Prepared by: L7 Ti I~y L.J. Tilly, Senior Engineer Structural Analysis & Hardware Design Verified by: (B)DFrew B.D. Frew, Principal Engineer Structural Analysis & Hardware Design Approved by: B7 lBranfund B.J. Branlund, Principal Engineer Structural Analysis & Hardware Design

GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version REPORT REVISION STATUS Revision I Purpose 0 1Initial Issue 1 Proprietary notations have been updated to meet current requirements.

• Revision bars have been provided in the right margin of each paragraph denoting change from the previous report.

• Sections 1.0 and 2.0 have been updated to include mention of Appendix H.

• Section 4.3.2.1 has been revised for clarification of the transients evaluated for the P-T curves.

• Section 4.3.2.1.2 has been revised to reflect a new analysis defining the CRD Penetration (Bottom Head) Core Not Critical P-T Curve; Appendix H has been added to provide a detailed discussion of the subject analysis and conclusions.

• A clarifying statement has been added to Section 4.3.2.2.4 regarding the use of K1t in the Beltline Core Not Critical P-T curves.

• Section 5.0 Figures 5-5 and 5-11, and Appendix B Tables B-I, B-2, and B-3 have been revised to incorporate changes to the CRD Penetration (Bottom Head) Core Not Critical P-T curve, as defined in Section 4.3.2.1.2 and Appendix H.

• Section 5.0 Figures 5-13 and 5-14 have been added to present composite pressure test and core not critical curves for 20 EFPY. Table B-5 has been added to present the tabulated values representinca these fiaures.

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GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version IMPORTANT NOTICE This is a non-proprietary version of the document GE-NE-0000-0003-5526-02R1, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( )).

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 11114101, 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-0000-0003-5526-02Rla Non-Proprietary Version EXECUTIVE

SUMMARY

This report provides the pressure-temperature curves (P-T curves) developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curve methodology includes the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the M, 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 Kjc 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 a 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 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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 beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature.

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

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GE Nuclear Energy GE-NE-0000-0003-5526-02RIa Non-Proprietary Version TABLE OF CONTENTS

1.0 INTRODUCTION

1 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITIAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 14 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 19

5.0 CONCLUSION

S AND RECOMMENDATIONS 50

6.0 REFERENCES

67

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GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 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)

APPENDIX G THICKNESS TRANSITION DISCONTINUITY EVALUATION APPENDIX H CORE NOT CRITICAL CALCULATION FOR BOTTOM HEAD (CRD PENETRATION)

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE LASALLE UNIT 1 RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2. CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 31 FIGURE 4-3. FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 37 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 53 0

FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [2 '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/IHR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100°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

[100F/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

[1000 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 FIGURE 5-13: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 20 EFPY [201F/HR OR LESS COOLANT HEATUP/COOLDOWN] 65 FIGURE 5-14: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 20 EFPY

[100lF/HR OR LESS COOLANT HEATUP/COOLDOWN] 66

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GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR LASALLE UNIT I VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR LASALLE UNIT 1 NOZZLE MATERIALS 12 TABLE 4-3: RTNDT VALUES FOR LASALLE UNIT 1 WELD MATERIALS 13 TABLE 4-4: LASALLE UNIT 1 BELTLINE ART VALUES (20 EFPY) 17 TABLE 4-5: LASALLE UNIT 1 BELTLINE ART VALUES (32 EFPY) 18 TABLE 4-6:

SUMMARY

OF THE IOCFR50 APPENDIX G REQUIREMENTS 21 TABLE 4-7: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 23 TABLE 4-8: APPLICABLE BWR/5 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 23 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 52

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version

1.0 INTRODUCTION

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

Complete P-T curves were developed for 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 K1c 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-0000-0003-5526-02R1 a Non-Proprietary Version 1.02 x 1018 n/cm 2 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. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Appendix G contains an evaluation of the vessel wall thickness discontinuity in the beltline region. Finally, Appendix H provides the core not critical calculation for the bottom head (CRD Penetration).

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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 K0c 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 1 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-NE-0000-0003-5526-02Rla Non-Proprietary Version 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. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Appendix G contains an evaluation of the vessel wall thickness discontinuity in the beltline region. Finally, Appendix H provides the core not critical calculation for the bottom head (CRD Penetration).

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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 1 Technical Specification, Section 1.1.

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 and are summarized as follows:

a. Test specimens shall be longitudinally oriented CVN specimens.

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

c. Pressure tests shall be conducted at a temperature at least 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.

10CFR50 Appendix G [8] states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented in an approved manner. GE developed methods for analytically GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data in WRC Bulletin 217 [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 1 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 1 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 20 F per ft-lb energy difference from 50 ft-lb.

For example, for the LaSalle Unit 1 beltline plate heat C5978-2 in the lower shell course, the lowest Charpy energy and test temperature from the CMTRs is 41 ft-lb at 40 0 F. The estimated 50 ft-lb longitudinal test temperature is:

T5OL = 40°F + [(50 - 41) ft-lb , 2 0F/ft-lb] = 58 0 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, T50T = 58 0F + 301F = 881F The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5oT- 60°F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version the case above is -10°F. Thus, the initial RTNDT for plate heat C5978-2 would be 28 0 F; however, a semi curve-fit approach using CMTR data was performed [5] that resulted i6*

an RTNDT for plate heat 05978-2 of 230F.

For the LaSalle Unit 1 beltline weld heat 1P3571 with flux lot 3958 (contained in the middle 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 1P3571 has a lowest Charpy energy of 40 ft-lb at 10°F as recorded in weld qualification records.

Therefore, T5 0T = 10°F + [(50 - 40) ft-lb , 2 0F/ft-lb] = 30°F The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T50T- 600 F). For LaSalle Unit 1, the dropweight testing to establish NDT was -30 0 F.

The value of (T50T - 60 0F) in this example is -30°F; therefore, the initial RTNDT was -300F.

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 1 (Heat Q2Q14VW-174W-1/6), the NDT is 40°F and the lowest CVN data is 48 ft-lb at 10°F. The corresponding value of (T5oT- 600 F) is:

(ToT - 600F) = {[10 + (50 - 48) ft-lb

• 2 0F/ft-lb] + 30 0F} - 60°F = -160F.

Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T50T- 60 0 F), which is 40 0 F.

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version In the bottom head region of the vessel, the full Charpy longitudinal test data was fit using a hyperbolic tangent fit to determine the 50 ft-lb transition temperature. For the bottom dome plate of LaSalle Unit 1 (Heat C6003-3), the NDT is 40°F and the 50 ft-lb longitudinal transition temperature is 77 0 F. The corresponding value of (TS0T - 600 F) was:

(TsoT - 60 0 F) = {77 0F + 30 0 1F} - 60°F = 470 F.

Therefore, the initial RTNDT was 47 0 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 60OF (as discussed in Section 4.3.2.3) is the LST for the bolting materials. Charpy data for the LaSalle Unit 1 closure studs do not meet the 45 ft-lb, 25 MLE requirement at 100 F. Therefore, the LST for the bolting material is 70 0 F. The highest RTNDT in the closure flange region is 120 F, for the vessel shell flange materials. Thus, the higher of the LST and the RTNDT +600 F is 72 0 F, the boltup limit in the closure flange region.

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

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version TOP~ HEAD

-TOP HEAD FLANGE SHELL #

,-SHELL 44 SHELL 93 VEI.OS 3-308 SHELL #2,

'GIRT WELD 8OTTOMOFACTIVE ~ SHELL #1 FUEL~BF 2"3 31" BOTTOM OF ELITUN E REO'DON~2-631 AXL LOD S 2-307 0 T MH A SUPPORT SKIRT Notes: (1) Refer to Tables 4-1, 4-2, and 4-3 for reactor vessel components and their heat identifications.

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

Figure 4-1: Schematic of the LaSalle Unit 1 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Table 4-1: RTNDT Values for LaSalle Unit 1 Vessel Materials COMPONENT HEAT TEST CHARPY TEMP. I C TET P

ENERGY N Y (TSOT-60) IDROPRTT DROPGHT WEIGHT RTF)

N I_ I NDT I

_ _ _(_F) _

PLATES & FORGINGS:

Top Head & Flange:

Vessel Flange, 308-02 2V-659 ATF-1 12 10 70 68 97 -20 10 10 Closure Flange, 319-02 ACT-USS-4P-1 997 Ser. 118 10 92 110 91 -20 10 10 Dome, 319-05 C7434-1 10 65 76 67 -20 -10 -10 Jpper Torus, 319-04 C7434-1 10 65 76 67 -20 -10 -10 Lower Torus, 319-03 C7376-2 10 65 74 73 -20 -10 -10 Shell Courses:

Jpper Shell C5987-1 10 63 55 35 10 -10 10 305-04 C5987-2 10 76 79 51 -20 -10 -10 C6003-2 40 65 49 50 12 10 12 Jpper Int. Shell C5996-2 10 62 71 66 -20 -10 -10 305-04 C5979-2 10 64 63 49 -18 -10 -10 C5996-1 10 65 60 77 -20 -10 -10 Middle Shell A5333-1 10 56 67 53 -20 -10 -10 305-03 B0078-1 10 73 49 70 -18 -10 -10 C6123-2 10 77 60 73 -20 -10 -10 Low-Int. Shell C6345-1 10 109 88 77 -20 -40 -20 305-02 C6318-1 10 80 66 72 -20 -20 -20 C6345-2 10 93 94 67 -20 -40 -20 Lower Shell C5978-1 40 53 48 48 14 10 14 305-01 C5978-2 40 62 60 41 28 -10 23*

C5979-1 40 73 92 65 10 -10 10 Bottom Head:

Bottom Head Dome, 306-17 C6003-3 40 36 39 40 38 40 47**

_ower Torus C5540-1 10 54 78 82 -20 -10 -10 306-18 C5328-1 40 64 51 51 10 -10 10 C5328-2 40 55 62 59 10 -10 10 Jpper Torus C5505-2 10 63 96 73 -20 -10 -10 306-19 C5445-3 10 70 67 70 -20 -10 -10 Support Skirt:

309-08 5P2003 Ser.201 10 81 74 103 -20 40 40 309-06 B1042-3 10 70 61 68 -20 10 10 309-04 C7159-4 40 28 25 34 60 60 60 STUDS: LST Closure Head Studs, 32-01 14716 10 45 1 43 1 43 70 Closure Nut/Washers, 326-02/03 24632 10 38 36 39 70 Value of RTNOT was obtained from semi curve-fit calculation using CMTR data.

Value of RTýDT is obtained from curve-fit of CMTR data.

GE Nuclear Energy GE-NE-0000-0003-5526-02R 1a Non-Proprietary Version Table 4-2: RTNDT Values for LaSalle Unit 1 Nozzle Materials TEST CHARPY ENERGY (TsoT-60) DROP RTNDT COMPONENT HEAT TEMP. CHRYEEG (T ) WEIGHT RTNO 0

(*F) (FT-LB) (°F) NOT ( F)

NOZZLES:

Recirc. Outlet Nozzle AV5840-OK9380 10 73 84 65 -20 0 0 314-02 AV5840-OK9381 10 56 84 80 -20 10 10 Recirc. Inlet Nozzle 02Q14VVW-175W 10 30 30 43 20 40 40 314-07 Q2Q6VW-175W 10 34 36 39 12 40 40 Steam Outlet Nozzles AV4276-919074 10 44 62 42 -4 30 30 316-07 AV4279-919236 10 84 55 80 -20 30 30 AV4442-9J9176 10 93 97 82 -20 30 30 AV4274-9H9176 10 69 100 71 -20 30 30 Feedwater Nozzle, 316-02 Q2Q14VW-174W-1/6 10 48 72 60 -16 40 40 Core Spray Nozzle AV4067-9H9168 10 79 70 71 -20 30 30 316-12 AV4068-9H9169 10 45 35 76 10 30 30 RHR/LPC1 Nozzles, 316-17 Q2022W-569F-1/3 10 44 44 37 6 10 10 CDR Hydro Return Nozzle, 315-10 AV3142-9G9640 10 34 30 44 20 30 30 Jet Pump Nozzles, 314-12 AV3138-9F-9231 B/C 10 116 90 96 -20 30 30 Closure Head Inst. Nozzle, 318-07 Q2023W-346J-1A 10 35 47 31 18 30 30 Vent Nozzle, 318-02 Q2Q24W-345J 10 78 109 122 -20 10 10 Drain Nozzle, 315-14 Q10lVW-738T 10 39 25 32 30 30 30 Stabilizer Bracket, 324-19 C4943-3 10 36 35 36 10 10 10 GE Nuclear Energy GE-NE-0000-0003-5526-02R 1 a Non-Proprietary Version Table 4-3: RTNDT Values for LaSalle Unit 1 Weld Materials T TETD ENERGY ROPRTT TES CHARPY (TSOT- 60

) WEIGHT NOT C(F) I (FT-LB) ('F) NDT (*F)

WELDS:

Vertical Welds:

2-307 Bottom Shell Long Seams 21935-1092-3889 10 97 90 83 -50 -50 -50 1-308 Upper Shell Long Seams 2-308 Upper Inter. Shell Long Seams 1-308 Upper Shell Long Seams 12008-1092-3889 10 97 90 83 -50 -50 -50 2-307 Bottom Shell Long Seams 1-308 Upper Shell Long Seams 305424-1092-3889 10 82 87 92 -50 -50 -50 3-308 Middle Shell Long Seams 3-308 Middle Shell Long Seams IP3571-1092-3958 10 40 46 46 -30 -50 -30 4-308 Lower Inter. Shell Long Seams 305414-1092-3947 10 82 66 80 -50 -50 -50 4-308 Lower Inter. Shell Long Seams 12008-1092-3947 10 92 91 92 -50 -50 -50 1-319 Closure Head Seg. Lower Torus FOAA 10 125 124 130 -50 -50 -50 2-319 Closure Head Seg. Upper Torus 1-319 Closure Head Seg. Lower Torus EAIB 10 118 1 129 1 107 -50 -50 -50 Girth Welds:

3-306 Bottom Hd. Build up for sup. Skirt 305414-1092-3951 10 66 1 61 1 62 -50 -50 -50 5-306.Bottom Hd. Dome to Side Seg, 6-306 Bottom Hd. Low. To Up Side Seg.

6-306 Bottom Hd. Low. To Up Side Seg. 305424-1092-3889 10 82 1 87 1 92 -50 -50 -50 4-307 Inlay in Bot. Sd for Core Sup Attch.

9-307 Bottom Head to Lower Shell 10120-0091-3458 10 124 1 130 1 122 -50 -50 -50 3-319 Close. Hd. Torus to Close. Hd. Fig.

9-307 Bottom Head to Lower Shell 51874-0091-3458 10 89 1 64 1 87 -50 -50 -50 3-319 Close. Hd. Torus to Close. Hd. Fig.

6-308 Upper Vessel Shell Girth Seam 9-307 Bottom Head to Lower Shell 51912-0091-3490 10 93 84 1 92 -50 -50 -50 6-308 Upper Vessel Shell Girth Seam 10137-0091-3999 10 101 108 107 -50 -50 -50 15-308 Flange to Upper Shell 6-308 Upper Vessel Shell Girth Seam 5P5622-0091-0831 -20 95 87 86 -80 -80 -80 6-308 Upper Vessel Shell Girth Seam 2P5755-0091-0831 -10 81 80 82 -70 -70 -70 6-308 Upper Vessel Shell Girth Seam 6329637-0091-3458 10 103 65 88 -50 -50 -50 6-308 Upper Vessel Shell Girth Seam 6329637-0091-3999 10 101 108 103 -50 -50 -50 15-308 Flange to Upper Shell 5-319 Closure Hd. Upper Torus to Dome 4-309 Support Skirt Forging to Bot. Hd. 90099-0091-3977 10 96 97 89 -50 -50 -50 4-309 Support Skirt Forging to Bot. Hd. 90136-0091-3998 10 110 109 107 -50 -50 -50 1-313 Up. Assy to Lower Closing Seams 4P6519-0091-0145 0 98 101 102 -60 -60 -60 1-313 Up. Assy to Lower Closing Seams 4P6519-0091-0842 0 46 59 48 -52 -80 -52 1-313 Up. Assy to Lower Closing Seams 4P6519-0091-0653 -40 57 63 73 -100 -60 -60 4-319 Close. Hd. UDDer Torus to Lower 6061L40-0091-3489 10 96 95 77 -50 -50 -50 GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (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 beltline plates and welds was made and summarized in Table 4-4 for 20 EFPY and Table 4-5 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- 0.10 log 0 2 + G2)0.

Margin = 2(7 CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = 1/4Tfluence / 1019 2

Margin = 2(a, + A2)0.5 c= = standard deviation on initial RTNDT, which is taken to be 0°F.

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

ART = Initial RTNOT+ SHIFT The margin term 0 A has constant values in Rev 2 of 170 F for plate and 28 0 F for weld.

However, orA need not be greater than 0.5 , ARTNDT. Since the GEBWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value of al is taken to be 0°F for the vessel plate and weld materials.

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 4.2.1.1 Chemistry The vessel beltline chemistries were obtained from the LaSalle Unit 1 NRC RAI submittal [13]. 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 1 flux for the vessel ID wall [14a] is 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 [14a] 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 1.01e9 n/cm 2-s. The peak fast flux for the RPV inner surface determined from surveillance capsule flux wires removed during the outage in Spring 1994 after Fuel Cycle 6 at a full power of 3323 MWt is 4.41e8 n/cm 2-s [5]. Linearly scaling the Reference 5 flux by 1.05 to the currently licensed power of 3489 MWt results in an estimated flux of 4.63e8 n/cm 2-s. Therefore, the Reference 14 flux bounds the flux determined from the surveillance capsule flux wire results by 218%.

The time period 32 EFPY is 1.01e9 sec, therefore the RPV ID surface fluence is as follows: RPV ID surface fluence = 1.01e9 n/cm 2 -s*l.01e9 s = 1.02e18 n/cm 2 . This fluence applies to the lower-intermediate and middle shells, the vertical welds for these shells, and the girth welds. The fluence is adjusted for the lower shell and the vertical welds for the lower shell based upon a peak / lower shell location ratio of 0.44 (at an elevation of approximately 230" above vessel "0"); hence the peak ID surface fluence used for these components is 4.49e17 n/cm 2. Similarly, the fluence is adjusted for the LPCI nozzle based upon a peak / LPCI nozzle location ratio of 0.244 (at an elevation of approximately 372" and at 450, 1350, and 2250 azimuths); hence the peak ID surface fluence used for this component is 2.49e17 n/cm 2.

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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. Table 4-4 lists values of beltline ART for 20 EFPY and Table 4-5 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-0000-0003-5526-02R1a Non-Proprietary Version Table 4-4: LaSalle Unit 1 Beltline ART Values (20 EFPY)

Middle & Loweer-lnlermedlte Plate. and Weld. 3-309, 4-308, 6-308 & 1-313 Thickness in Inches = 6.13 Ratio Peak/ Location = 1.00 32 EFPY Peak I.D. fluence = 1.02E+18 n/cmn2 32 EFPY Peak 1/4 T fluence = 7.1E+17 n/cm^2 20 EFPY Peak 1/4 T luence = 4.4E+17 n/cm^2 Leowe Plate and Welds 2-307 Thickness in inches= 7.13 Ratio Peak/ Location = 0.44 32 EFPY Peak I.D. fluence = 4.49E+17 nrcm^2 Location = 229 7,8"Eleo'eion 32 EFPY Peak 1/4 T fluence = 2.9E+17 n/cm^2 20 EFPY Peak 114 T flience = 1.8E+17 rVcm^2 LPCINozzle Thickness in inches= 6.13 Ratio Peak/ Location = 0.244 32 EFPY Peak I.D. fluence = 2.49E+17 nrcm'n2 Location =-355" Elevation 32 EFPY Peak 1/4 T fluence = 1.7E+17 n/cm^2 20 EFPY Peak 1/4T fluence = 1.1E+17 n/cm^2 Initial 1/4 T 20 EFPY 20 EFPY 20 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF RTT Fluence A RTNlT a, cr Margin Shift ART PF ntcm^2 'F PF °F °F PLATES:

Lower Shell Assy 307-04 G-5603-1 C5978-1 0.110 0.580 74 14 1.8E+17 12 0 6 12 24 38 G-5603-2 C5978-2 0,110 0.590 74 23 1.8E+17 12 0 6 12 24 47 G-5603-3 C5979-1 0.120 0660 84 10 1.8E+17 14 0 7 14 27 37 Lower-intermedlate Shell Assy 308-06 G5604-1 C6345-1 0.150 0.490 104 -20 4.4E+17 28 0 14 28 57 37 G5604-2 C6318-1 0.120 0.510 81 -20 4.4E+17 22 0 11 22 44 24 G5604-3 C6345-2 0.150 0.510 105 -20 4.46+17 29 0 14 29 57 37 Middle Shell Aasy 308-05 G5605-1 A5333-1 0.120 0.540 82 -10 4.4E+17 22 0 11 22 45 35 G5605-2 B0078-1 0.150 0500 105 -10 4.4E+17 29 0 14 29 57 47 G5605-3 C6123-2 0.130 0.680 93 -10 4.4E+17 25 0 13 25 51 41 WELDS:

Middle 3-308 A,B,C 305424/3889 0.273 0.629 189.5 -50 4.46+17 52 0 26 52 104 54 1P3571/3958 0.283 0.755 212 -30 4.4E+17 58 0 28 56 114 84 Lower-Inlermedlale 4-308 A,B,C 305414/3947 0.337 0.609 209 -50 4.4E+17 57 0 28 56 113 63 12008/3947 0.235 0.975 233 -50 4.4E+17 64 0 28 56 120 70 305414&12008 Tandem 0.286 0.792 219 -50 4.4E+17 60 0 28 56 116 66 Lower 2-307 A,B,C 21935/3889 0.183 0.704 172 -50 1.86+17 28 0 14 28 56 6 12008/3889 0.235 0.975 233 -50 1.8E+17 38 0 19 38 76 26 21935&12008 tandem 0.213 0.867 209 -50 1.8E+17 34 0 17 34 68 18 Girth 6-308 6329637 0.205 0.105 98 -50 4.4E+17 27 0 13 27 54 4 1-313 4P6519 0.131 0.060 64 -52 4.4E+17 18 0 9 18 35 -17 FORGINGS:

LPCI Nozzle Q2Q22W 0.100 0.820 67 10 1.1E+17 8 0 4 8 15 25 GE Nuclear Energy GE-NE-0000-0003-5526-02R 1a Non-Proprietary Version Table 4-5: LaSalle Unit 1 Beltline ART Values (32 EFPY)

Middle & Lower-nltermedlate Plates and Weida 3-308, 4-308,6-308 & 1-313 Thickness in Inches = 6,13 Ratio Peak/ Location = 1.00 32 EFPY Peak ID. fluence = 1.02E+18 n/cm^2 32 EFPY Peak 1/4 T fluence = 7.1E+17 n/cm^2 32 EFPY Peak 1/4 T fluence = 7.1E+17 n/cm^2 Low- Plate and Welds 2-307 Thickness In inches= 7.13 Ratio Peak/ Location = 0.44 32 EFPY Peak I.D. fluence = 4,49E+17 n/cm^2 Location = 229 7/8" Elevation 32 EFPY Peak 114 T fluence = 2.9E+17 n/cm^2 32 EFPY Peak 1/4 T fluence = 2.9E+17 nlcm^2 LPCI Nozz.e Thickness in inches= 6.13 Ratio Peak/ Location = 0.244 32 EFPY Peak I.D. fluence = 2.49E+17 n/cm^2 Location =-355" Elevation 32 EFPY Peak 1/4 T fluence = 1.7E+17 n/cm^2 32 EFPY Peak 1/4 T fluence = 1.7E+17 n/cm^2 Initial 1/4T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF RTNDT Fluence A RTrDT o" o', Margin Shift ART P

°F ncm^2 °F °F °F 'F PLATES:

Lower Shell Assy 307-04 G-5603-1 C5978-1 0.110 0.580 74 14 2.9E+17 16 0 8 16 32 46 G-5603-2 C5978-2 0.110 0.590 74 23 2.9E+17 16 0 8 16 32 55 G-5603-3 C5979-1 0.120 0.660 84 10 2.9E+17 18 0 9 18 36 46 Lower-Intermediate Shell Assy 308-06 G5604-1 C6345-1 0.150 0.490 104 -20 7.1E+17 36 0 17 34 70 50 G5604-2 C6318-1 0.120 0.510 81 -20 7.1E+17 28 0 14 28 57 37 G5604-3 C6345-2 0.150 0.510 105 -20 7.1E+17 37 0 17 34 71 51 Middle Shell Assy 308-05 G5605-1 A5333-1 0.120 0.540 82 -10 7.1E+17 29 0 14 29 58 48 G5605-2 B0078-1 0.150 0500 105 -10 7.1E+17 37 0 17 34 71 61 G5605-3 C6123-2 0.130 0.680 93 -10 7.1E+17 33 0 16 33 65 55 WELDS:

Middle 3-308 A,BC 305424/3889 0.273 0.629 189.5 -50 7.1E+17 67 0 28 56 123 73 1P3571/3958 0.283 0.755 212 -30 7.16+17 74 0 28 56 130 100 Lower-intermediate 4-308 A,B,C 305414/3947 0.337 0.609 209 -50 7.1E+17 73 0 28 56 129 79 12008/3947 0.235 0.975 233 -50 7.1E+17 82 0 28 56 138 88 305414&12008 Tandem 0.286 0.792 219 -50 7.1E+17 77 0 28 56 133 83 Lower 2-307 A,6,C 21935/3889 0.183 0.704 172 -50 2.9E+17 37 0 19 37 74 24 12008/3889 0.235 0.975 233 -50 2.9E+17 50 0 25 50 101 51 21935&12008 tandem 0.213 0.867 209 -50 2.9E+17 45 0 23 45 90 40 Girth 6-308 6329637 0.205 0.105 98 -50 7.1E+17 34 0 17 34 69 19 1-313 4P6519 0.131 0,060 64 -52 7.1E+17 22

• 0 11 22 45 -7 FORGINGS:

LPCI Nozzle Q2022W 0.100 0820 67 10 1.7E+17 10 0 5 10 21 31 GE Nuclear Energy GE-NE-0000-0003-5526-02Rl a Non-Proprietary Version 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 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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, Kir, 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 4-6:

GE Nuclear Energy GE-NE-0000-0003-5526-02R 1 a Non-Proprietary Version Table 4-6: Summary of the 10CFR50 Appendix G Requirements

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 0 F*

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 0 F*

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

b. Core critical - Curve C

1. At < 20% of preservice hydrotest Larger of ASM E 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 10CFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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/cm 2) to cause any significant shift of RTNDT. Non-beltline components include nozzles (see Appendix E),

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

Detailed stress analyses of the non-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The 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, and loss of recirculation pump flow. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWR/6 components: the feedwater nozzle (FW) and the CRD penetration (bottom head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-7 and 4-8.

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version Table 4-7: Applicable BWR/5 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B 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 (B only)

Top Head Nozzles (B only)

Recirculation Outlet Nozzle (B only)

Table 4-8: Applicable BWR/5 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity Identificatlon CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Shell**

Support Skirt**

Shroud Support Attachments" Core AP and Liquid Control Nozzle**

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

The P-T curves for the non-beltline region were conservatively developed for a large BWR/6 (nominal inside diameter of 251 inches). The analysis is considered appropriate for LaSalle Unit 1 as the plant specific geometric values are bounded by the generic

- 23 -

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version 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 1 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.

[1]

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

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

The (( )) evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section Xl Appendix G [6] and shown below. The results of that computation were K = 143.6 ksi-in11 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 0 F.

The limit for the coolant temperature change rate is 2 0 °F/hr or less.

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version The value of M, for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

was based on a thickness of 8.0 inches; hence, t112 = 2.83. The resulting value obtained was:

Mm = 1.85 for -Vi<2 Mm = 0.926 It for 2< It <3.464 = 2.6206 Mm = 3.21 for -j >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Kib is calculated from the equation in Paragraph G-2214.2 [6]:

1 Kim M Mm pm = (( )) ksi-in /2 112 KIb = (2/3) Mm "apb = )) ksi-in The total K, is therefore:

K, = 1.5 (Kim+ Kib) + Mm (a-srm + (2/3)" Csb) = 143.6 ksi-in112 This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific K, is based on the K, equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) = In [(1 - 33.2) / 20.734] / 0.02

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GE Nuclear Energy GE-NE-0000-0003-5526-02R 1a Non-Proprietary Version (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 84'F The generic curve was generated by scaling 143.6 ksi-in11 2 by the nominal pressures and calculating the associated (T - RTNDT):

((I The highest RTNDT for the bottom head plates and welds is 47°F, as shown in Tables 4-1, 4-2, and 4-3. ((

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version Second, the P-T curve is dependent on the calculated K4 value, and the K4 value is proportional to the stress and the crack depth as shown below:

K,cc 7a ) / (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, K is proportional to R/(t) 1 2. The generic curve value of R/(t) 1/2, based on the generic BWR/6 bottom head dimensions, is:

Generic: R/ (t)112 = 138/(8)1/2 = 49 inch 1/ 2 (4-2)

The LaSalle Unit 1 specific bottom head dimensions are R = 127.4 inches and t =7.38 inches minimum [19], resulting in:

LaSalle Unit 1 specific: R / (t)"/ 2 = 127.4 / (7.38)1/2 = 46.9 inch1 /2 (4-3)

Since the generic value of R/(t)1/ 2 is larger, the generic P-T curve is conservative when applied to the LaSalle Unit 1 bottom head.

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version 4.3.2.1.2 Core Not CriticalHeatup/Cooldown - Non-Beitline Curve B (Using Bottom Head)

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

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

The calculated value of K4 for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KIR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical. Therefore, the K,value for the core not critical condition is (143.6 / 1.5) *2.0 = 191.5 ksi-in 112 GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the K10 equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:

(T - RTNDT) = In [(K - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] /0.02 (T - RTNDT) = 102'F The generic curve was generated by scaling 192 ksi-in'12 by the nominal pressures and calculating the associated (T - RTNDT):

Core Not Critical CRD Penetration K,and (T - RTNDT) as a Function of Pressure 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 47 0 F, as shown in Tables 4-1, 4-2, and 4-3. ((

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 4-7, 4-8, and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 4.3.2.1.3 PressureTest - Non-Beltline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, K1, for the feedwater nozzle was computed using the methods from WRC 175 [15] together with the nozzle dimension for a generic 251-inch BWR/6 feedwater nozzle. The result of that computation was K4 = 200 ksi-in 112 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 K, is shown below using the BWR/6, 251-inch dimensions:

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

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

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

a= 1/4(t 2 + t 2)1/2 =2.36 inches tn = thickness of nozzle = 7.125 inches tv = thickness of vessel = 6.1875 inches rn = apparent radius of nozzle = ri + 0.29 r,=7.09 inches ri = actual inner radius of nozzle = 6.0 inches

r. = nozzle radius (nozzle corner radius) = 3.75 inches Thus, a/rn = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, is 1.4. Including the safety factor of 1.5, the stress 1/2 intensity factor, KI, is 1.5 c (7ca) - F(a/rn):

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GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Nominal K4 = 1.5 34.97* (7c.2.36)1/2 . 1.4 = 200 ksi-in 112 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 pressure test condition:

(T - RTNDT) = In [(K4 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 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-in1/2 by the nominal pressures and calculating the associated (T - RTND-), ((

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version The highest RTNDT for the feedwater nozzle materials is 40°F as shown in Tables 4-1, 4-2, and 4-3. The generic pressure test P-T curve is applied to the LaSalle Unit 1 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 400 F.

))

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

Vessel Radius, R, 127 inches Vessel Thickness, t. 6.69 inches Vessel Pressure, P,, 1563 psig GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version Pressure stress: a = PR / t = 1563 psig - 127 inches / (6.69 inches) = 29,671 psi. The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding ay = 32.64 ksi. The factor F (a/rn) from Figure A5-1 of WRC-175 is determined where:

a . 1/4( 2 + tv 2)1/2 =2.31 inches tn= thickness of nozzle = 6.38 inches tv= thickness of vessel = 6.69 inches rn= apparent radius of nozzle = ri + 0.29 r,=7.29 inches ri= actual inner radius of nozzle = 6.13 inches rc nozzle radius (nozzle corner radius) = 4.0 inches Thus, a/rn = 2.31 / 7.29 = 0.32. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.32, is 1.5. Including the safety factor of 1.5, the stress intensity factor, KI, is 1.5 a (ra) 112

• F(a/rn):

1/2 Nominal KI= 1.5 *32.64 *(7 *2.31)1/2. 1.5 = 197.9 ksi-in 4.3.2.1.4 Core Not CriticalHeatup/Cooldown - Non-Beltline 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 GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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 .o (7a)Y2 F(a/rn) (4-4) where SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/rn) is the shape correction factor.

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version Finite element analysis of a nozzle corner 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, apm, and primary bending, Cpb. Secondary membrane, asm, and secondary bending, asb, stresses are included in the total K, by using ASME Appendix G [6] methods for secondary portion, Kisý:

S= Mm (cysm + (2/3) 37 -b) (4-5)

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [15].

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

The method to solve for (T - RTNDT) for a specific K, is based on the K1c 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 [(K4 - 33.2) / 20.734] 0.02 (4-6)

Example Core Not Critical Heatup/Cooldown Calculation for Feedwater Nozzle/Upper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the ((

feedwater nozzle (( )) analysis, where feedwater injection of 40°F into the vessel while at operating conditions (551.4 0 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 (arpm) was adjusted for the actual (( ))

vessel thickness of 6.1875 inches (i.e., cprm = 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 7y, = 45.0 ksi t= 6.1875 inches

• pb = 0.22 ksi asb = 19.04 ksi a = 2.36 inches r, = 7.09 inches t, = 7.125 inches In this case the total stress, 60.29 ksi, exceeds the yield stress, ays, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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 = [cays - apm + ((atotaI - a*y) / 30)] / (c'total - Opm) (4-7)

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

crpm = 24.84 ksi asm = 9.44 ksi apb = 0.13 ksi asb = 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, t1/2 = 3.072. The resulting value obtained was:

Mm = 1.85 for Vft<_2 Im = 0.926 ft for 2<1it <3.464 = 2.845 Mm = 3.21 for Vft'>3.464 The value F(a/rn), taken from Figure A5-1 of WRC'Bulletin 175 for an a/rn of 0.33, is therefore, F (a / rn) = 1.4 Kip is calculated from Equation 4-4:

Kip = 2.0 * (24.84 + 0.13). (* .2.36)1/2* 1.4 112 Kip = 190.4 ksi-in Kis is calculated from Equation 4-5:

KI, = 2.845 * (9.44 + 2/3

• 11.10)

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rl a Non-Proprietary Version 2

K1r = 47.9 ksi-in 1 The total K,is, therefore, 238.3 ksi-in1 /2.

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

(T - RTNDT) = In [(238.3- 33.2) / 20.734] / 0.02 (T - RTNDT) = 115*F The (( )) curve was generated by scaling the stresses used to determine the K4; 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 2 , the pressure is 1050 psig and the hot reactor vessel temperature is 551.4°F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tsaturation - 40) / (551.4 - 40). From K1 the associated (T - RTNDT) can be calculated:

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GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure Noiq7_nalIP essure~ -Saturation Tem~p. R K1j> (T - RTNDT) 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 K1 .

The highest non-beltline RTNDT for the feedwater nozzle at LaSalle Unit 1 is 40°F as shown in Tables 4-1, 4-2, and 4-3. The generic curve is applied to the LaSalle Unit 1 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-0000-0003-5526-02R1a Non-Proprietary Version 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.

An evaluation was performed [22] for the vessel wall thickness transition discontinuity located between the lower and lower-intermediate shells in the beltline region.

Appendix G of this report contains an update of the evaluation.

4.3.2.2.1 Beltline 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 (tin) ratio of 15, is treated as a thin-walled cylinder. The maximum. stress is the hoop stress, given as:

am = PR / tmin (4-8)

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

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

The relationship between K1c and temperature relative to reference temperature (T - RTNDT) is based on the K1, equation of Paragraph A-4200 in ASME Appendix A [17]

for the pressure test condition:

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Kim SF = Kic = 20.734 exp[0.02 (T - RTNDT )] + 33.2 (4-9)

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

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

4.3.2.2.2 Calculationsfor the Belfine 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 = -30 + 130 = 100°F (Based on ART values in Section 4.2)

Vessel Height H = 863.3 inches Bottom of Active Fuel Height B = 216 inches Vessel Radius (to inside of clad) R = 126.7 inches Minimum Vessel Thickness (without clad) t = 6.13 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 + (863.3 - 216) 0.0361 = 1128 psig Pressure stress:

o= PR/t (4-11)

= 1.128 126.7/6.13=23.3 ksi GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

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

hence, t112 = 2.48. The resulting value obtained was:

Mm = 1.85 for -J<2 Mrm = 0.926 ft for 2< ft<3.464 = 2.29 Mm = 3.21 for Ft >3.464 The stress intensity factor for the pressure stress is Kim = Mm - C. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20°F/hr instead of 100 0 F/hr.

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

Using the KIo equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 53.4, and Kit= 3.01 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 - 53.4 + 3.01 - 33.2) / 20.734] / 0.02

= 43.9°F T can be calculated by adding the adjusted RTNDT:

T = 43.9 + 100 = 143.9°F for P = 1105 psig 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 +Kit (4-13)

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version where Kim is primary membrane K due 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 ATw is based on one-dimensional heat conduction through an insulated flat plate:

C 2T(x,t) / 0 x2 = 1 /f3(ff(x,t) / Ot) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and P is the thermal diffusivity.

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

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 / 2P - GCx / P3+ To (4-15)

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version This equation is normalized to plot (T - To) / AT, versus x / C.

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

4.3.2.2.4 Calculationsfor the Beitline 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. In addition, there is a Kit term for the thermal stress.

The additional inputs used to calculate Kjt are:

Coolant heatup/cooldown rate, normally 100°F/hr G = 100 °F/hr Minimum vessel thickness, including clad thickness C = 0.588 ft (7.06 inches)

(the maximum vessel thickness is conservatively used)

Thermal diffusivity at 550°F (most conservative value) P = 0.354 ft/ 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 /2p (4-16)

= 100 (0.588)2/ (2 . 0.354) = 49°F GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.308) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kt = Mt - AT = 15.1, can be calculated. The conservative value for thermal diffusivity at 550OF is used for all calculations; therefore, K1t is constant for all pressures. Kim has the same value as that calculated in Section 4.3.2.2.2.

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

(T - RTNDT) = ln[((2

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

= ln[(2

• 53.4 + 15.1 - 33.2) / 20.734] /0.02

= 72.7 *F T can be calculated by adding the adjusted RTNDT:

T = 72.7 + 100 = 172.7 '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 1 at low pressures.

The approach used for LaSalle Unit 1 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 60OF 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°F requirement above the RTNDT

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version provides the additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Tables 4-1 and 4-3, the limiting initial RTNDT for the closure flange region is represented by both the top head and vessel shell flange materials at 120 F, and the LST of the closure studs is 70°F; therefore, the bolt-up temperature value used is 72 0 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 + 90 0F) and Curve B temperature no less than (RTNDT + 120°F).

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

However, temperatures should not be permitted to be lower than 68 0 F for the reason discussed below.

The shutdown margin, provided in the LaSalle Unit 1 Technical Specification, is calculated for a water temperature of 680F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68°F limit, further extensive calculations would be required to justify a lower temperature. The 72 0 F limit for the upper vessel and beltline region and the 68 0 F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel. When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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 1 of 10CFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60 0 F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 72 0 F, based on an RTNDT of 120 F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160°F or the temperature required for the hydrostatic pressure test (Curve A at 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-0000-0003-5526-02Rla Non-Proprietary Version

5.0 CONCLUSION

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

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

• Closure flange region (Region A)

• Core beltline region (Region B)

• Upper vessel (Regions A & B)

" Lower vessel (Regions B & C)

For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 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 beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature.

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version The following P-T curves were generated for LaSalle Unit 1.

" Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 20 and 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 beltline limited above 1040 psig for curve A and 1090 psig for curve B for 20 EFPY. The P-T curves are beltline limited above 710 psig for curve A and 660 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-0000-0003-5526-02R1 a Non-Proprietary Version Table 5-1: Composite and Individual Curves Used To Construct Composite 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 Bottom Head Limits (CRD Nozzle) Figure 5-5 B-1 & B-3 Upper Vessel Limits (FW Nozzle) Figure 5-6 B-1 & B-3 Beltline 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 A &B Composite Curves for 20 EFPY Bottom Head and Composite Curve A Figure 5-13 B-5 for 20 EFPY*

Bottom Head and Composite Curve B Figure 5-14 B-5 for 20 EFPY*

• The Composite Curve A & B curve is the more limiting of three limits: 10CFR50 Bolt-up Limits, Upper Vessel Limits (FW Nozzle), and Beltline Limits. A separate Bottom Head Limits (CRD Nozzle) curve is individually included on this figure.

• • The Composite Curve C curve is the more limiting of four limits: 10CFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Limits (FW Nozzle), and Beltline Limits.

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version 1400 1300 1200 1100 0

1000

_j 900 SINITIAL 47°F FORRTndt BOTTOMVALUE IS HEAD I-LU

-J 800 LU LU 700 HEATUP/COOLDOWN RATE OF COOLANT I- < 20 °F/HR C-)

U) 600 w

I-J 500 LU U/) 400 UJ a-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]

- 53 -

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version 1400 1300 1200 1100 0 1000 Lu

0. 900 0 INITIAL RTndt VALUE IS

_j , 40'F FOR UPPER VESSELI 800 Lu o 700 HEATUP/COOLDOWN RATE OF COOLANT 600 < 2 0 oF/HR z

3 500 w

400 Lu F312 P-SI1G1 300 200

-UPPER VESSEL LIMITS (Including 100 FLANGE REGION 72°F *Flange and FW Nozzle Limits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)

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

[200F/hr or less coolant heatup/cooldown]

- 54 -

GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version 1400 1300 1200 1100 S ""1000 900 0

I-

-j BELTLINE CURVE w 800 ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 20 114 0 700 F-HEATUP/COOLDOWN 600 RATE OF COOLANT z

< 20 °F/HR 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 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 20 EFPY

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

GE Nuclear Energy GE-NE-0000-0003-5526-02RIa Non-Proprietary Version 1400 1300 1200 1100 a.1000 0

0. 900 I--

,-J BELTLINE CURVE 800 ADJUSTED AS SHOWN:

I',L EFPY SHIFT ('F)

L) 32 130 o 700 i-Z w I HEATUP/COOLDOWN 3600 RATE OF COOLANT 2 < 20°F/HR

_7 500 ca400 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-N E-0000-0003-5526-02R 1a Non-Proprietary Version 1400 1300 1200 1100 1000 900 INITIAL RTndt VALUE IS 0 61.6°F FOR BOTTOM HEADJ

-J 800 Co w

o 700 HEATUP/COOLDOWN RATE OF COOLANT 600 < 100°F/HR z

500 w

) 400 w

0~

300 200 100 0

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

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

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

- 57 -

GE Nuclear Energy GE-NE-0000-0003-5526-02RIa Non-Proprietary Version 1400 1300 1200 1100 a._

" 1000 0

900 SINITIAL RTndt VALUE IS 40'F FOR UPPER VESSEL M

HEATUP/COOLDOWN I-RATE OF COOLANT 600 < 100°F/HR C',

500 3 00 CU 400 w

200

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

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

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

- 58 -

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 1400 1300 1200 1100 U) a.

1000' BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT ('F)

0. 900 20 114 I-0 800 I-

.- I w 700 HEATUP/COOLDOWN uLJ RATE OF COOLANT I-U) 600 < 100°F/HR a.M n,, 500 Z

0m 400 300 /

200 IOCFR50 - BELTLINE LIMITS BOLTUP 100 72°F 0 -

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

('F)

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 -

GE Nuclear Energy GE-NE-000-0003-5526-02R1a Non-Proprietary Version 1400 1300 1200 1100

'a CL 1000 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 0~ 900 32 130 0 800 700 HEATUPICOOLDOWN RATE OF COOLANT 600 <_100°F/HR 0

cf)

Lu 500 400 300 r 200___

10CFR50 - BELTLINE LIMITS BOLTUP 100 72°F 0 4- -

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

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

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 -30°F FOR BELTLINE, 40°F FOR UPPER VESSEL, 1200 AND 47°F FOR BOTTOM HEAD 1100 C. BELTLINE CURVE 1000 ADJUSTED AS SHOWN:

EFPY SHIFT ('F)

(. 900 20 114 0

i-

.-J (0 800 Uo HEATUP/COOLDOWN RATE OF COOLANT

< 100OF/HR 0I-. 700 M

C-,

600 z

-3 500 Lu n 400 312 PSIG 0.

300 200

- BELTLINE AND NON-BELTLINE 100 0Minimum Criticality LIMITS Temperature 72°F 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°F/hr or less coolant heatup/cooldown]

-61 -

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 -30'F FOR BELTLINE, 40'F FOR UPPER VESSEL, AND 1100 47°F FOR BOTTOM HEAD

- 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT ('F) 1L 900 32 130 0 800 LU W

o 700 Fo HEATUP/COOLDOWN 600 RATE OF COOLANT

< 20°F/HR Z

LU 500 z

n 400 LU 300

-- UPPER VESSEL 200 AND BELTLINE LIMITS BOTTOM HEAD -

100 CURVE 0

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

(°F)

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

[200F/hr or less coolant heatup/cooldown]

- 62 -

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 1400 1300 I INITIAL RTndt VALUES ARE 1200 I -30°F FOR BELTLINE, 40°F FOR UPPER VESSEL, 1100 AND 61.6°F FOR BOTTOM HEAD 1000

'C w

BELTLINE CURVES ADJUSTED AS SHOWN:

900 EFPY SHIFT ('F) 0 32 130

-J 800 w

o 700 U

HEATUP/COOLDOWN 600 RATE OF COOLANT z < 100°F/HR 500 w

400 W

300

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

CURVE 0

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

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

- 63 -

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 -30°F FOR BELTLINE, 40'F FOR UPPER VESSEL, 1200 AND 47°F FOR BOTTOM HEAD 1100 CL BELTLINE CURVE 1000 ADJUSTED AS SHOWN:

LU, EFPY SHIFT ('F)

a. 900 32 130 0

I-

,-J U)

(I, 800 u.'

HEATUP/COOLDOWN RATE OF COOLANT

< 100°F/HR o

I- 700 ul

, 600 z

1-

500 m 400 300 200 BELTLINE AND NON-BELTLINE 100 LIMITS 0

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

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

[100°F/hr or less coolant heatup/cool down]

- 64 -

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200

-30 F FOR BELTLINE, 40'F FOR UPPER VESSEL, AND 1100 47°F FOR BOTTOM HEAD 1000 BELTLINE CURVES 4

w ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 900 0 20 114 Cn 800 0 700 F-HEATUP/COOLDOWN RATE OF COOLANT z 600 < 20°F/HR

• 500 w

400 w

300

-UPPER VESSEL 200 AND BELTLINE LIMITS

---

BOTTOM HEAD 100 CURVE 0

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

(*F)

Figure 5-13: Composite Pressure Test P-T Curves [Curve A] up to 20 EFPY

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

- 65 -

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE

-30°F FOR BELTLINE, 40'F FOR UPPER VESSEL, 1100 AND 61.6'F FOR BOTTOM HEAD 1000 BELTLINE CURVES w ADJUSTED AS SHOWN:

CL 900 EFPY SHIFT ('F) 0 F- 20 114.

800 w

o 700 C-)

HEATUP/COOLDOWN 600 RATE OF COOLANT Z < 100°F/HR w 500 O

400 W

0:

300 UPPER VESSEL 200 AND BELTLINE LIMITS 100 -BOTTOM HEAD CURVE 0

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

(°F)

Figure 5-14: Composite Core Not Critical P-T Curves [Curve B] up to 20 EFPY

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

- 66 -

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version

6.0 REFERENCES

1. B.J. Branlund, "Pressure-Temperature Curves for CornEd LaSalle Unit 1", GE-NE, San Jose, CA, May 2000, (GE-NE-B13-02057-00-06R1, Revision 1).

2. GE Drawing Number 731E776, "Reactor Vessel Thermal Cycles," GE-NED, San Jose, CA, Revision 3 (GE Proprietary).

3. GE Drawing Number 158B8136, "Reactor Vessel Nozzle Thermal Cycles",

GE-NED, San Jose, CA, Revision 7 (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. 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).

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

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

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

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

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

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

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version

11. Letter from B. Sheron to R.A. Pinelli, "Safety Assessment of Report NEDC-32399-P, Basis for GE RTNDT Estimation Method, September 1994", USNRC, December 16,1994.

12. QA Records & RPV CMTR's:

LaSalle Unit 1 -QA Records & RPV CMTR's LaSalle Unit 1 GE PO# 205-AK104, Manufactured by CE.

13. 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-11 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.

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.

GE Nuclear Energy GE-NE-0000-0003-5526-02RIa Non-Proprietary Version

18. ((

19. Bottom Head and Feedwater Nozzle Dimensions:

a) CE Drawing # E232-842, Rev. 2, "Bottom Head Machining and Welding for 251" ID BWR", (GE VPF # 2029-107, Rev. 4).

b) CE Drawing # E-232-863, Rev. 4, "Nozzle Details for 251" ID BWR",

(GE VPF 2029-099, Rev. 7).

20. ((

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

22. B.J. Branlund, "Plant LaSalle Units 1 and 2 RPV Shell Thickness Transition and Other Geometric Discontinuities", (GE-NE-B1301869-01), June 1998.

GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version 1]

A-2

GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Per ASME Code Appendix G, Section G2223 (c), fracture toughness analysis to demonstrate protection against non-ductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5" or less provided the lowest service temperature is not lower than RTNDT plus 60'F. 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 317-01 Core Differential Pressure SB 166 1.5.12 & Thickness is < 2.5" and made

& Liquid Poison - 1.6 of Alloy 600; therefore, no Penetration < 2.5" further fracture toughness Bottom Head evaluation is required.

315-14 Drain- Penetration < 2.5" SA-508 Cl. 1 1.5.1, The discontinuity of the CRD

- Bottom Head 1.5.15 & nozzle listed in Table A-i 1.6 bounds this discontinuity; therefore, no further fracture toughness evaluation is required.

321-05 Seal Leak Detection* - 1.5.1 Not a pressure boundary Penetration -1" component; therefore, Flange requires no fracture toughness evaluation.

319-06 Top Head Lifting Lugs SA-533 GR. B 1.5.1 & Not a pressure boundary Attachment to Top Head CL. 1 1.5.13, 1.6 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-0000-0003-5526-02Rla Non-Proprietary Version APPENDIX A

REFERENCES:

1.5. RPV Drawings 1.5.1. CE Drawing # 232-788, Rev. 3, "General Arrangement Elevation for 251" I.D. BWR," (GE VPF #2029-117, Rev. 4).

1.5.2. CE Drawing # 232-790, Rev. 8, "Lower Vessel Shell Assembly Machining & Welding for 251" I.D. BWR", (GE VPF#2029-036, Rev. 8).

1.5.3. CE Drawing # 232-791, Rev. 15, "Upper Vessel Shell Assembly Machining & Welding for 251" I.D. BWR", (GE VPF #2029-037, Rev. 14).

1.5.4. CE Drawing # 232-792, Rev. 7, "Vessel Machining for 251" I.D.

BWR," (GE VPF #2029-054, Rev. 8).

1.5.5. CE Drawing # 232-796, Rev. 9", Vessel External Attachments for 251" I.D. BWR," (GE VPF #2029-085, Rev. 10).

1.5.6. CE Drawing # 232-801, Rev. 0, "Closure Head Final Machining for 251" I.D. BWR", (GE VPF #2029-114, Rev. 2).

1.5.7. CE Drawing # 232-839, Rev. 4, "Closure Head Nozzle Details for 251" I.D. BWR," (GE VPF #2029-108, Rev. 6).

1.5.8. CE Drawing # 232-842, Rev. 2, "Bottom Head Machining & Welding for 251" I.D. BWR," (GE VPF #2029-107, Rev. 4).

1.5.9. CE Drawing # 232-861, Rev. 0, "Vessel Support Skirt Assembly and Details for 251" I.D. BWR," (GE VPF #2029-121, Rev. 2).

1.5.10. CE Drawing # 232-862, Rev. 0, "Bottom Head Penetrations for 251" I.D. BWR," (GE VPF #2029-120, Rev. 2).

1.5.11. CE Drawing # 232-863, Rev. 4, "Nozzle Details for 251" I.D. BWR,"

(GE VPF #2029-099, Rev. 7).

1.5.12. CE Drawing # 232-880, Rev. 1, "Nozzle Details for 251" I.D. BWR,"

(GE VPF #2029-115, Rev. 3).

1.5.13. CE Drawing # 232-911, Rev. 4, "Closure Head Machining & Welding for 251" 1.D. BWR," (GE VPF #2029-083, Rev. 6).

1.5.14. CE Drawing # 232-937, Rev. 3, "Shroud Support Details and Assembly for 251" I.D. BWR," (GE VPF #2029-082, Rev. 5).

1.5.15. CE Drawing # 232-938, Rev. 6, "Nozzle Details for 251" I.D. BWR,"

(GE VPF #2029-084, Rev. 8) 1.6. CE Stress Report, "Analytical Report for LaSalle County Station Unit 1 for Commonwealth Edison Company," CE Power Systems, Combustion Engineering, Inc, Chattanooga, TN, (Report No CENC-1250.)

A-4

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 1.7. Wu, Tang, "LaSalle 1&2 Neutron Flux Evaluation", GE-NE, San Jose, CA, May 2002, (GE-NE-0000-0002-5244-01, Revision 0)(GE Proprietary).

A-5

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-I. LaSalle Unit 1 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, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32EFPY BOTTOM UPPER 32 EFPY~

HEAD VESSEL BELTLINE7 HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A3/4 CURVE B CURVE B CURVE B (PSIG) ('F) 0 68.0 72.0 72.0 68.0 72.0 72.0 68.0 72.0 72.0 10 72.0 68.0 72.0 68.0 72.0 72.0 20 72.0 68.0 72.0 68.0 72.0 72.0 30 72.0 68.0 72.0 68.0 72.0 72.0 40 72.0 68.0 72.0 68.0 72.0 72.0 50 72.0 68.0 72.0 68.0 72.0 72.0 60 72.0 68.0 72.0 68.0 72.0 72.0 70 72.0 68.0 72.0 68.0 72.0 72.0 68.0 72.0 80 72.0 68.0 72.0 72.0 90 72.0 68.0 72.0 68.0 72.0 72.0 100 72.0 68.0 72.0 68.0 72.0 72.0 110 72.0 68.0 72.0 68.0 72.0 72.0 120 72.0 68.0 72.0 68.0 72.0 72.0 130 72.0 68.0 74.2 68.0 72.0 72.0 140 72.0 68.0 77.4 68.0 72.0 72.0 150 72.0 68.0 80.2 68.0 72.0 160 72.0 68.0 82.9 73.9 68.0 72.0 170 72.0 68.0 85.5 76.5 68.0 72.0 180 72.0 68.0 87.9 78.9 68.0 72.0 190 72.0 68.0 90.2 81.2 68.0 72.0 200 72.0 68.0 92.3 83.3 68.0 72.0 210 72.0 68.0 94.3 85.3 68.0 72.0 220 72.0 68.0 96.3 87.3 68.0 72.0 230 72.0 68.0 98.1 89.1 B-2

GE Nuclear Energy GE-N E-0000-0003-5526-02Rl1a Non-Proprietary Version TABLE B-1. LaSalle Unit 1 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, 5-3, 5-4, 5-5, 5-6, & 5-8

~BOTTOM UAPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD, VIESSEL BELTLINE 'HEAD VESSEL BELTLlNE~

PRESSURE CURVEA CURVENA CURVEA CURVEB CURVE B CURVE B Q(PSIG) ('F)< ('F) ý"~NF) ('F) (-F) ('F) 240 68.0 72.0 72.0 68.0 99.9 90.9 250 68.0 72.0 72.0 68.0 101.6 92.6 260 68.0 72.0 72.0 68.0 103.2 94.2 270 68.0 72.0 72.0 68.0 104.8 95.8 280 68.0 72.0 72.0 68.0 106.3 97.3 290 68.0 72.0 72.0 68.0 107.8 98.8 300 68.0 72.0 72.0 68.0 109.2 100.2 310 68.0 72.0 72.0 68.0 110.5 101.5 312.5 68.0 72.0 72.0 68.0 110.9 101.9 312.5 68.0 102.0 102.0 68.0 132.0 132.0 320 68.0 102.0 102.0 68.0 132.0 132.0 330 68.0 102.0 102.0 68.0 132.0 132.0 340 68.0 102.0 102.0 68.0 132.0 132.0 350 68.0 102.0 102.0 68.0 132.0 132.0 360 68.0 102.0 102.0 68.0 132.0 132.0 370 68.0 102.0 102.0 68.0 132.0 132.0 380 68.0 102.0 102.0 68.0 132.0 132.0 390 68.0 102.0 102.0 68.0 132.0 132.0 400 68.0 102.0 102.0 68.0 132.0 132.0 410 68.0 102.0 102.0 68.0 132.0 132.0 420 68.0 102.0 102.0 68.0 132.0 132.0 430 68.0 102.0 102.0 68.0 132.0 132.0 440 68.0 102.0 102.0 68.0 132.0 132.0 450 68.0 102.0 102.0 68.0 132.0 132.0 460 68.0 102.0 102.0 68.0 132.0 132.0 470 68.0 102.0 102.0 69.6 132.0 132.0 480 68.0 102.0 102.0 72.1 132.0 132.0 B-3

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE B-I. LaSalle Unit 1 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, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD~ VESSEL BELTLINE~ IHEAD~ VESSEL~ BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) ~('F). ('F)7 ('0 F) (3F) 490 68.0 102.0 102.0 74.4 132.0 132.0 500 68.0 102.0 102.0 76.6 132.0 132.0 510 68.0 102.0 102.0 78.8 132.0 132.0 520 68.0 102.0 102.0 80.8 132.2 132.0 530 68.0 102.0 102.0 82.8 133.0 132.0 540 68.0 102.0 102.0 84.7 133.8 132.0 550 68.0 102.0 102.0 86.5 134.6 132.0 560 68.0 102.0 102.0 88.3 135.4 132.0 570 68.0 102.0 102.0 90.0 136.1 132.0 580 68.0 102.0 102.0 91.6 136.9 132.0 590 68.0 102.0 102.0 93.2 137.6 132.9 600 68.0 102.0 102.0 94.8 138.1 134.0 68.0 610 68.0 102.0 102.0 96.3 138.6 135.2 620 68.0 102.0 102.0 97.7 139.0 136.3 630 68.0 102.0 102.0 99.1 139.4 137.4 640 68.0 102.0 102.0 100.5 139.8 138.5 650 68.0 102.0 102.0 101.8 140.2 139.5 660 68.0 102.0 102.0 103.1 140.7 140.5 670 68.0 102.0 102.0 104.4 141.1 141.5 680 68.0 102.0 102.0 105.7 141.5 142.5 690 68.0 102.0 102.0 106.9 141.9 143.5 700 102.0 102.0 108.0 142.3 144.4 68.7 710 102.0 102.8 109.2 142.7 145.4 70.1 720 102.0 104.4 110.3 143.1 146.3 71.5 730 102.3 106.0 111.4 143.5 147.2 72.8 740 103.1 107.5 112.5 143.9 148.1 74.1 750 104.0 108.9 113.6 144.2 148.9 B-4

GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version Enrg Nucea TABLE B-I. LaSalle Unit 1 P-T Curve Values for 32 EFPY GE~~i Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-4, 5-5, 5-6, & 5-8 (OftTOM UPPERý 32 EFPY BOTTOM UPPER ~32 EFPY E HEAD~ VESSELP BELTLINE HEAD VESSEL> BELTLINE PRESSURE URVEA CURVES CURVEA CURVE Bi. CURVE B~ CURVE B

~<(PSIG) ('F) (*F)~ ('F) 760 75.4 104.8 110.3 114.6 144.6 149.8 770 76.6 105.6 111.7 115.6 145.0 150.6 780 77.8 106.3 113.0 116.6 145.4 151.4 790 79.0 107.1 114.4 117.6 145.8 152.2 800 80.2 107.9 115.6 118.5 146.1 153.0 810 81.3 108.6 116.9 119.5 146.5 153.8 820 82.4 109.4 118.1 120.4 146.9 154.6 830 83.5 110.1 119.3 121.3 147.2 155.4 840 84.5 110.8 120.4 122.2 147.6 156.1 850 85.6 111.5 121.5 123.0 147.9 156.9 860 86.6 112.2 122.6 123.9 148.3 157.6 870 87.6 112.9 123.7 124.7 148.6 158.3 880 88.5 113.6 124.8 125.6 149.0 159.0 890 89.5 114.3 125.8 126.4 149.3 159.7 900 90.4 114.9 126.8 127.2 149.7 160.4 910 91.4 115.6 127.8 128.0 150.0 161.1 920 92.3 116.2 128.8 128.7 150.4 161.7 930 93.1 116.9 129.7 129.5 150.7 162.4 940 94.0 117.5 130.7 130.3 151.0 163.0 950 94.9 118.1 131.6 131.0 151.4 163.7 960 95.7 118.7 132.5 131.7 151.7 164.3 970 96.6 119.3 133.4 132.5 152.0 165.0 980 97.4 119.9 134.3 133.2 152.4 165.6 990 98.2 120.5 135.1 133.9 152.7 166.2 1000 99.0 121.1 136.0 134.6 153.0 166.8 1010 99.7 121.7 136.8 135.2 153.3 167.4 1020 100.5 122.2 137.6 135.9 153.6 168.0 B-5

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version TABLE B-1. LaSalle Unit 1 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, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM ,UPPER S32 EF*Y BOTTOM UPPERP~ 3B2 EFPY

>HEAD VESSEL BELTLINE HEAD~ VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVEB~ CURVE B~ CURVE B (PSIG). ('F) ('F)2 ('F)j 1030 101.3 122.8 138.4 136.6 154.0 168.6 1040 102.0 123.4 139.2 137.2 154.3 169.1 1050 102.7 123.9 140.0 137.9 154.6 169.7 1060 103.4 124.5 140.7 138.5 154.9 170.3 1070 104.2 125.0 141.5 139.1 155.2 170.8 1080 104.9 125.5 142.2 139.8 155.5 171.4 1090 105.6 126.1 143.0 140.4 155.8 171.9 1100 106.2 126.6 143.7 141.0 156.1 172.5 1105 106.6 126.8 144.0 141.3 156.3 172.7 1110 106.9 127.1 144.4 141.6 156.4 173.0 1120 107.6 127.6 145.1 142.2 156.7 173.5 1130 108.2 128.1 145.8 142.8 157.0 174.1 1140 108.9 128.6 146.5 143.3 157.3 174.6 1150 109.5 129.1 147.1 143.9 157.6 175.1 1160 110.1 129.6 147.8 144.5 157.9 175.6 1170 110.8 130.1 148.4 145.0 158.2 176.1 1180 111.4 130.6 149.1 145.6 158.5 176.6 1190 112.0 131.1 149.7 146.1 158.7 177.1 1200 112.6 131.5 150.4 146.7 159.0 177.6 1210 113.2 132.0 151.0 147.2 159.3 178.0 1220 113.8 132.5 151.6 147.8 159.6 178.5 1230 114.3 132.9 152.2 148.3 159.9 179.0 1240 114.9 133.4 152.8 148.8 160.2 179.5 1250 115.5 133.8 153.4 149.3 160.4 179.9 1260 116.0 134.3 154.0 149.8 160.7 180.4 1270 116.6 134.7 154.6 150.3 161.0 180.8 1280 117.1 135.2 155.1 150.8 161.2 181.3 B-6

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-1. LaSalle Unit 1 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, 5-3, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPP~ER EýFP5Y> BOTTOM~ UPPER -32 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURES CURVE A CUR~VEA CURVE A~ CURVE B CURVE B CURVE Bi (P, I9:

.(IF)

(IF) (IF) 1290 117.7 135.6 155.7 151.3 161.5 181.7 1300 118.2 136.0 156.3 151.8 161.8 182.2 1310 118.7 136.5 156.8 152.3 162.1 182.6 1320 119.3 136.9 157.4 152.8 162.3 183.1 1330 119.8 137.3 157.9 153.2 162.6 183.5 1340 120.3 137.7 158.4 153.7 162.8 183.9 1350 120.8 138.1 159.0 154.2 163.1 184.3 1360 121.3 138.6 159.5 154.6 163.4 184.8 1370 121.8 139.0 160.0 155.1 163.6 185.2 1380 122.3 139.4 160.5 155.5 163.9 185.6 1390 122.8 139.8 161.0 156.0 164.1 186.0 1400 123.3 140.2 161.5 156.4 164.4 186.4 B-7

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-2. LaSalle Unit 1 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 & 5-12 BQOM*LLJPPEffiR RPV & ~BOTTOM UPPER RPV &: NONbELTLINE HEAD BELTLINE AT HEAD BELTLINE AT AND BELTLINE 32 EFPY~ 32 EFPY AT 32 EFPY PRESSUR CURVE A CURVEA CURVE B CURVEDB CURVE C (PSIG) ('F) (F ('F) 0 68.0 72.0 72.0 10 68.0 72.0 68.0 72.0 72.0 20 68.0 72.0 68.0 72.0 72.0 30 68.0 72.0 68.0 72.0 72.0 68.0 72.0 40 68.0 72.0 72.0 50 68.0 72.0 68.0 72.0 72.0 60 68.0 72.0 68.0 72.0 80.0 70 68.0 72.0 68.0 72.0 87.2 80 68.0 72.0 68.0 72.0 93.2 90 68.0 72.0 68.0 72.0 98.3 100 68.0 72.0 68.0 72.0 102.8 110 68.0 68.0 72.0 72.0 68.0 72.0 106.9 120 68.0 68.0 72.0 72.0 68.0 72.0 110.7 130 68.0 72.0 68.0 74.2 114.2 140 68.0 72.0 68.0 77.4 117.4 150 68.0 72.0 68.0 80.2 120.2 160 68.0 72.0 68.0 82.9 122.9 170 68.0 72.0 68.0 85.5 125.5 180 68.0 72.0 68.0 87.9 127.9 190 68.0 72.0 68.0 90.2 130.2 200 68.0 68.0 72.0 72.0 68.0 92.3 132.3 210 68.0 94.3 134.3 220 68.0 96.3 136.3 B-8

GE Nuclear Energy GE-N E-0000-0003-5526-02 Rl1a Non-Proprietary Version TABLE B-2. LaSalle Unit 1 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 & 5-12

$<BOTTOM UPPER RPW& BOTTOM UPPEffR RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELILINE AT AND BELTLINE 32 EFPY 32 EFPY AT 32 EFPY PRESSURE -CURVE A CURVE A ~CURVE B CURVE B ~. CURVE C (P I)('F) ('F) (' F) F ('F) 230 68.0 72.0 68.0 98.1 138.1 240 68.0 72.0 68.0 99.9 139.9 250 68.0 72.0 68.0 101.6 141.6 260 68.0 72.0 68.0 103.2 143.2 270 68.0 72.0 68.0 104.8 144.8 280 68.0 72.0 68.0 106.3 146.3 290 68.0 72.0 68.0 107.8 147.8 300 68.0 72.0 68.0 109.2 149.2 310 68.0 72.0 68.0 110.5 150.5 312.5 68.0 72.0 68.0 110.9 150.9 312.5 68.0 102.0 68.0 132.0 172.0 320 68.0 102.0 68.0 132.0 172.0 330 68.0 102.0 68.0 132.0 172.0 340 68.0 102.0 68.0 132.0 172.0 350 68.0 102.0 68.0 132.0 172.0 360 68.0 102.0 68.0 132.0 172.0 370 68.0 102.0 68.0 132.0 172.0 380 68.0 102.0 68.0 132.0 172.0 390 68.0 102.0 68.0 132.0 172.0 400 68.0 102.0 68.0 132.0 172.0 410 68.0 102.0 68.0 132.0 172.0 420 68.0 102.0 68.0 132.0 172.0 430 68.0 102.0 68.0 132.0 172.0 440 68.0 102.0 68.0 132.0 172.0 450 68.0 102.0 68.0 132.0 172.0 B-9

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version TABLE B-2. LaSalle Unit 1 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 & 5-12 BOTTOM ~UPPER FRPV & BOTTOM UPPER RPV & NONBELTLINE HEAD BELTLINE AT HEAD BELTLINE AT AND BELTLINE 32 EFPY S32 EFPY AT 32 EFPY PRESSURE CURVE A ~~CURVE A CURVE B CURVE B CURVE C (PS IG) <(OF) ('F) (:F). (°F) ('F) 460 68.0 102.0 68.0 132.0 172.0 470 68.0 102.0 69.6 132.0 172.0 480 68.0 102.0 72.1 132.0 172.0 490 68.0 102.0 74.4 132.0 172.0 500 68.0 102.0 76.6 132.0 172.0 510 68.0 102.0 78.8 132.0 172.0 520 68.0 102.0 80.8 132.2 172.2 530 68.0 102.0 82.8 133.0 173.0 540 68.0 102.0 84.7 133.8 173.8 550 68.0 102.0 86.5 134.6 174.6 560 68.0 102.0 88.3 135.4 175.4 570 68.0 102.0 90.0 136.1 176.1 580 68.0 102.0 91.6 136.9 176.9 590 68.0 102.0 93.2 137.6 177.6 600 68.0 102.0 94.8 138.1 178.1 610 68.0 102.0 96.3 138.6 178.6 620 68.0 102.0 97.7 139.0 179.0 630 68.0 102.0 99.1 139.4 179.4 640 68.0 102.0 100.5 139.8 179.8 650 68.0 102.0 101.8 140.2 180.2 660 68.0 102.0 103.1 140.7 180.7 670 68.0 102.0 104.4 141.5 181.5 680 68.0 102.0 105.7 142.5 182.5 690 68.0 102.0 106.9 143.5 183.5 700 68.0 102.0 108.0 144.4 184.4 8-10

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE B-2. LaSalle Unit I 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 & 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & *NONBELTLINEE H EAD BELTLINE AT HEAD BELTLINE AT AND BELTLINE 32 EFPY~ 32 EFPY AT 32 EFPY PRESSURE CURVEA ~CURVE A CURVE B CURVE B CURVE C K~(PSIG) ('F) ('F) ('F) ('F) 710 68.7 102.8 109.2 145.4 185.4 720 70.1 104.4 110.3 146.3 186.3 730 71.5 106.0 111.4 147.2 187.2 740 72.8 107.5 112.5 148.1 188.1 750 74.1 108.9 113.6 148.9 188.9 760 75.4 110.3 114.6 149.8 189.8 770 76.6 111.7 115.6 150.6 190.6 780 77.8 113.0 116.6 151.4 191.4 790 79.0 114.4 117.6 152.2 192.2 800 80.2 115.6 118.5 153.0 193.0 810 81.3 116.9 119.5 153.8 193.8 820 82.4 118.1 120.4 154.6 194.6 830 83.5 119.3 121.3 155.4 195.4 840 84.5 120.4 122.2 156.1 196.1 850 85.6 121.5 123.0 156.9 196.9 860 86.6 122.6 123.9 157.6 197.6 870 87.6 123.7- 124.7 158.3 198.3 880 88.5 124.8 125.6 159.0 199.0 890 89.5 125.8 126.4 159.7 199.7 900 90.4 126.8 127.2 160.4 200.4 910 91.4 127.8 128.0 161.1 201.1 920 92.3 128.8 128.7 161.7 201.7 930. 93.1 129.7 129.5 162.4 202.4 940 94.0 130.7 130.3 163.0 203.0 950 94.9 .131.6 131.0 163.7 203.7 B-11

GE Nuclear Energy GE-N E-0000-0003-5526-02Rl1a Non-Proprietary Version TABLE B-2. LaSalle Unit 1 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 & 5-12 IBOTTOM UPPER RPV & BOTTOM UPPER RPV & NONBELTLINE HEAD9K BELTLINE AT HEAD BELTLINE AT AND BELTLINE 32 EFPY 32 EFPY AT 32 EFPY.

DRESSURE CURVE A CURVE A CURVE B CURVE B CURVE &

(PSIG) ~ ( F) ( F) ~ (,F) (IF) ('F) 960 95.7 132.5 131.7 164.3. 204.3 970 96.6 133.4 132.5 165.0 205.0 980 97.4 134.3 133.2 165.6 205.6 990 98.2 135.1 133.9 166.2 206.2 1000 99.0 136.0 134.6 166.8 206.8 1010 99.7 136.8 135.2 167.4 207.4 1020 100.5 137.6 135.9 168.0 208.0 1030 101.3 138.4 136.6 168.6 208.6 1040 102.0 139.2 137.2 169.1 209.1 1050 102.7 140.0 137.9 169.7 209.7 1060 103.4 140.7 138.5 170.3 210.3 1070 104.2 141.5 139.1 170.8 210.8 1080 104.9 142.2 139.8 171.4 211.4 1090 105.6 143.0 140.4 171.9 211.9 1100 106.2 143.7 141.0 172.5 212.5 1105 106.6 144.0 141.3 172.7 212.7 1110 106.9 144.4 141.6 173.0 213.0 1120 107.6 145.1 142.2 173.5 213.5 1130 108.2 145.8 142.8 174.1 214.1 1140 108.9 146.5 143.3 174.6 214.6 1150 109.5 147.1 143.9 175.1 215.1 1160 110.1 147.8 144.5 175.6 215.6 1170 110.8 148.4 145.0 176.1 216.1 1180 111.4 149.1 145.6 176.6 216.6 1190 112.0 149.7 146.1 177.1 217.1 B- 12

GE Nuclear Energy GE-N E-OO00-OO03-5526-02R 1 a Non-Proprietary Version TABLE B-2. LaSalle Unit 1 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 & 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RP V & NONBELtTLINE HEAD BELTILINE AT HEAD BELTLINE AT AND BELTLINE.

32 EFPY~ - 32 EFPY AT32 EFPY PRESSURE CURVEA CURVE A CURVE B <CURVE B ~CURVE C (PSIG) S ('F) 1200 112.6 150.4 146.7 177.6 217.6 1210 113.2 151.0 147.2 178.0 218.0 1220 113.8 151.6 147.8 178.5 218.5 1230 114.3 152.2 148.3 179.0 219.0 1240 114.9 152.8 148.8 179.5 219.5 1250 115.5 153.4 149.3 179.9 219.9 1260 116.0 154.0 149.8 180.4 220.4 1270 116.6 154.6 150.3 180.8 220.8 1280 117.1 155.1 150.8 181.3 221.3 1290 117.7 155.7 151.3 181.7 221.7 1300 118.2 156.3 151.8 182.2 222.2 1310 118.7 156.8 152.3 182.6 222.6 1320 119.3 157.4 152.8 183.1 223.1 1330 119.8 157.9 153.2 183.5 223.5 1340 120.3 158.4 153.7 183.9 223.9 1350 120.8 159.0 154.2 184.3 224.3 1360 121.3 159.5 154.6 184.8 224.8 1370 121.8 160.0 155.1 185.2 225.2 1380 122.3 160.5 155.5 185.6 225.6 1390 122.8 161.0 156.0 186.0 226.0 1400 123.3 161.5 156.4 186.4 226.4 B-13

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-3. LaSalle Unit 1 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~ 720 E6FP. BOTTOM UPPER 20 EýFPY VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A~ CURVE B CURVE B CURVE B HEAD 0 0 (PSIG) ( F) ('F) (2.0 ( 0 7.(

F) ('F) 0 68.0 72.0 72.0 68.0 72.0 72.0 10 68.0 72.0 72.0 68.0 72.0 72.0 20 68.0 72.0 72.0 68.0 72.0 72.0 30 68.0 72.0 72.0 68.0 72.0 72.0 40 68.0 72.0 72.0 68.0 72.0 72.0 50 68.0 72.0 72.0 68.0 72.0 72.0 60 68.0 72.0 72.0 68.0 72.0 72.0 70 68.0 72.0 72.0 68.0 72.0 72.0 80 68.0 72.0 72.0 68.0 72.0 72.0 90 68.0 72.0 72.0 68.0 72.0 72.0 100 68.0 72.0 72.0 68.0 72.0 72.0 110 68.0 72.0 72.0 68.0 72.0 72.0 120 68.0 72.0 72.0 68.0 72.0 72.0 130 68.0 72.0 72.0 68.0 74.2 72.0 140 68.0 72.0 72.0 68.0 77.4 72.0 150 68.0 72.0 72.0 68.0 80.2 72.0 160 68.0 72.0 72.0 68.0 82.9 72.0 170 68.0 72.0 72.0 68.0 85.5 72.0 180 68.0 72.0 72.0 68.0 87.9 72.9 190 68.0 72.0 72.0 68.0 90.2 75.2 200 68.0 72.0 72.0 68.0 92.3 77.3 210 68.0 72.0 72.0 68.0 94.3 79.3 220 68.0 72.0 72.0 68.0 96.3 81.3 230 68.0 72.0 72.0 68.0 98.1 83.1 B- 14

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-3. LaSalle Unit 1 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~ 20iEFPY BELTLINE HEAD ~VESSEL HEAD VESSEL BELTLINE.

PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CUIRVE'B

('F). ('F) (6F) 240 68.0 72.0 72.0 68.0 99.9 84.9 250 68.0 72.0 72.0 68.0 101.6 86.6 260 68.0 72.0 72.0 68.0 103.2 88.2 270 68.0 72.0 72.0 68.0 104.8 89.8 280 68.0 72.0 72.0 68.0 106.3 91.3 290 68.0 72.0 72.0 68.0 107.8 92.8 300 68.0 72.0 72.0 68.0 109.2 94.2 310 68.0 72.0 72.0 68.0 110.5 95.5 68.0 312.5 72.0 72.0 68.0 110.9 95.9 312.5 68.0 102.0 102.0 68.0 132.0 132.0 320 68.0 102.0 102.0 68.0 132.0 132.0 330 68.0 102.0 102.0 68.0 132.0 132.0 340 68.0 102.0 102.0 68.0 132.0 132.0 350 68.0 102.0 102.0 68.0 132.0 132.0 360 68.0 102.0 102.0 68.0 132.0 132.0 370 68.0 102.0 102.0 68.0 132.0 132.0 380 68.0 102.0 102.0 68.0 132.0 132.0 390 68.0 102.0 102.0 68.0 132.0 132.0 400 68.0 102.0 102.0 68.0 132.0 132.0 410 68.0 102.0 102.0 68.0 132.0 132.0 420 68.0 102.0 102.0 68.0 132.0 132.0 430 68.0 102.0 102.0 132.0 132.0 68.0 440 68.0 102.0 102.0 68.0 132.0 132.0 450 68.0 102.0 102.0 68.0 132.0 132.0 68.0 460 102.0 102.0 68.0 132.0 132.0 68.0 470 102.0 102.0 69.6 132.0 132.0 68.0 480 102.0 102.0 72.1 132.0 132.0 B-15

GE Nuclear Energy GE-NE-0000-0003-5526-02R 1 a Non-Proprietary Version TABLE B-3. LaSalle Unit 1 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, BELTLI NE.

PRESSURE CURVEA CL!RVEA~ CURVE A CURVE B CURVE B CURVE B' 0

(3F) 490 68.0 102.0 102.0 74.4 132.0 132.0 102.0 500 68.0 102.0 76.6 132.0 132.0 102.0 510 68.0 102.0 78.8 132.0 132.0 102.0 520 68.0 102.0 80.8 132.2 132.0 102.0 530 68.0 102.0 82.8 133.0 132.0 102.0 540 68.0 102.0 84.7 133.8 132.0 102.0 550 68.0 102.0 86.5 134.6 132.0 102.0 560 68.0 102.0 88.3 135.4 132.0 102.0 570 68.0 102.0 90.0 136.1 132.0 102.0 580 68.0 102.0 91.6 136.9 132.0 102.0 590 68.0 102.0 93.2 137.6 132.0 102.0 600 68.0 102.0 94.8 138.1 132.0 102.0 610 68.0 102.0 96.3 138.6 *132.0 102.0 620 68.0 102.0 97.7 139.0 132.0 102.0 630 68.0 102.0 99.1 139.4 132.0 102.0 640 68.0 102.0 100.5 139.8 132.9 132.0 102.0 650 68.0 102.0 101.8 140.2 132.0 102.0 660 68.0 102.0 103.1 140.7 132.0 102.0 670 68.0 102.0 104.4 141.1 132.0 102.0 680 68.0 102.0 105.7 141.5 132.0 102.0 690 68.0 102.0 106.9 141.9 132.0 102.0 700 68.0 102.0 108.0 142.3 132.0 102.0 710 68.7 102.0 109.2 142.7 132.0 102.0 720 70.1 102.0 110.3 143.1 132.0 102.0 730 71.5 102.3 111.4 143.5 132.0 102.0 740 72.8 103.1 112.5 143.9 132.1 102.0 750 74.1 104.0 113.6 144.2 B-16

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE B-3. LaSalle Unit 1 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 CURVEA CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) ('F) (F)( ) ("F) ("F) 760 75.4 104.8 102.0 114.6 144.6 133.8 770 76.6 105.6 102.0 115.6 145.0 134.6 780 77.8 106.3 102.0 116.6 145.4 135.4 790 79.0 107.1 102.0 117.6 145.8 136.2 800 80.2 107.9 102.0 118.5 146.1 137.0 810 81.3 108.6 102.0 119.5 146.5 137.8 820 82.4 109.4 102.1 120.4 146.9 138.6 830 83.5 110.1 103.3 121.3 147.2 139.4 840 84.5 110.8 104.4 122.2 147.6 140.1 850 85.6 111.5 105.5 123.0 147.9 140.9 860 86.6 112.2 106.6 123.9 148.3 141.6 870 87.6 112.9 107.7 124.7 148.6 142.3 880 88.5 113.6 108.8 125.6 149.0 .143.0 890 89.5 114.3 109.8 126.4 149.3 143.7 900 90.4 114.9 110.8 127.2 149.7 144.4 910 91.4 115.6 111.8 128.0 150.0 145.1 920 92.3 116.2 112.8 128.7 150.4 145.7 930 93.1 116.9 113.7 129.5 150.7 146.4 940 94.0 117.5 114.7 130.3 151.0 147.0 950 94.9 118.1 115.6 131.0 151.4 147.7 960 95.7 118.7 116.5 131.7 151.7 148.3 970 96.6 119.3 117.4 132.5 152.0 149.0 980 97.4 119.9 118.3 133.2 152.4 149.6 990 98.2 120.5 119.1 133.9 152.7 150.2 1000 99.0 121.1 120.0 134.6 153.0 150.8 1010 99.7 121.7 120.8 135.2 153.3 151.4 1020 100.5 122.2 121.6 135.9 153.6 152.0 B-17

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-3. LaSalle Unit 1 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

~KBOTTOM UPPER 20FP BOTTOM UPPER 20 EFPY HEAD VESSEL BELTLINE HEAD VESSEL1 BELTLINE

(' F)UP E !i PRESSURE~ CURVE A CURVE A CURVEA CURVE B CURVEB- ~CURVE B.

(PSIG). (F) ('F) ('F)~ ('F) 1030 101.3 122.8 122.4 136.6 154.0 152.6 1040 102.0 123.4 123.2 137.2 154.3 153.1 1050 102.7 123.9 124.0 137.9 154.6 153.7 1060 103.4 124.5 124.7 138.5 154.9 154.3 1070 104.2 125.0 125.5 139.1 155.2 154.8 1080 104.9 125.5 126.2 139.8 155.5 155.4 1090 105.6 126.1 127.0 140.4 155.8 155.9 1100 106.2 126.6 127.7 141.0 156.1 156.5 1105 106.6 126.8 128.0 141.3 156.3 156.7 1110 106.9 127.1 128.4 141.6 156.4 157.0 1120 107.6 127.6 129.1 142.2 156.7 157.5 1130 108.2 128.1 129.8 142.8 157.0 158.1 1140 108.9 128.6 130.5 143.3 157.3 158.6 1150 109.5 129.1 131.1 143.9 157.6 159.1 1160 110.1 129.6 131.8 144.5 157.9 159.6 1170 110.8 130.1 132.4 145.0 158.2 160.1 1180 111.4 130.6 133.1 145.6 158.5 160.6 1190 112.0 131.1 133.7 146.1 158.7 161.1 1200 112.6 131.5 134.4 146.7 159.0 161.6 1210 113.2 132.0 135.0 147.2 159.3 162.0 1220 113.8 132.5 135.6 147.8 159.6 162.5 1230 114.3 132.9 136.2 148.3 159.9 163.0 1240 114.9 133.4 136.8 148.8 160.2 163.5 1250 115.5 133.8 137.4 149.3 160.4 163.9 1260 116.0 134.3 138.0 149.8 160.7 164.4 1270 116.6 134.7 138.6 150.3 161.0 164.8 1280 117.1 135.2 139.1 150.8 161.2 165.3 B-18

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-3. LaSalle Unit 1 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 BOUOMK ~UPPER~ BOTTOM UPPER 20 EFPYA H EAD VESSEL, BELTLINE HEAD VESSEL, BELTLINE PRESSURE CURVEA CURVE A ~CURVE A~ CURVE B CURVE B CURVE B~

(PSIG) ('F) ('F) j(0 F) 1290 117.7 135.6 139.7 151.3 161.5 165.7 1300 118.2 136.0 140.3 151.8 161.8 166.2 1310 118.7 136.5 140.8 152.3 162.1 166.6 152.8 1320 119.3 136.9 141.4 162.3 167.1 1330 119.8 137.3 141.9 153.2 162.6 167.5 1340 120.3 137.7 142.4 153.7 162.8 167.9 1350 120.8 138.1 143.0 154.2 163.1 168.3 1360 121.3 138.6 143.5 154.6 163.4 168.8 1370 121.8 139.0 144.0 155.1 163.6 169.2 1380 122.3 139.4 144.5 155.5 163.9 169.6 1390 122.8 139.8 145.0 156.0 164.1 170.0 1400 123.3 140.2 145.5 156.4 164.4 170.4 B-19

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C

. For Figure 5-9 UPPER BOTTOM 20OEFPYi PRESSURE VESSEL CURVE C H~EAD CURVE C BELTLINE CURVE C (iPSIG) ('F) ('F). ('F) 0 72.0 68.0 72.0 10 72.0 68.0 72.0 20 72.0 68.0 72.0 30 72.0 68.0 72.0 40 72.0 68.0 72.0 50 72.0 68.0 72.0 60 80.0 68.0 72.0 70 87.2 68.0 72.0 80 93.2 68.0 72.0 90 98.3 68.0 72.0 100 102.8 68.0 72.0 110 106.9 68.0 72.0 120 110.7 68.0 72.0 130 114.2 68.0 72.0 140 117.4 68.0 72.0 150 120.2 68.0 72.0 160 122.9 68.0 72.0 170 125.5 68.0 72.0 180 127.9 68.0 72.0 190 130.2 68.0 72.0 200 132.3 68.0 72.0 210 134.3 68.0 72.0 220 136.3 68.0 72.0 230 138.1 68.0 72.0 240 139.9 68.0 72.0 250 141.6 68.0 75.0 B-20

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPERS BOTbM~ 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C BELTLINE CURVE C (PS IG) ('F) 260 143.2 68.0 80.7 68.0 270 144.8 85.9 68.0 280 146.3 90.5 68.0 290 147.8 94.8 68.0 300 149.2 98.7 68.0 310 150.5 102.4 68.0 312.5 150.9 103.3 68.0 312.5 172.0 172.0 68.0 320 172.0 172.0 68.0 330 172.0 172.0 68.0 340 172.0 172.0 68.0 350 172.0 172.0 68.0 360 172.0 172.0 68.0 370 172.0 172.0 68.0 380 172.0 172.0 390 172.0 69.3 172.0 400 172.0 73.3 172.0 410 172.0 77.0 172.0 420 172.0 80.5 172.0 430 172.0 83.8 172.0 440 172.0 86.8 172.0 450 172.0 89.7 172.0 460 172.0 92.4 172.0 470 172.0 95.0 172.0 480 172.0 97.5 172.0 490 172.0 99.8 172.0 500 172.0 102.0 172.0 B-21

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version TABLE B-4. LaSalle Unit 1 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) 510 172.0 104.2 172.0 520 172.2 106.2 172.0 530 173.0 108.2 172.0 540 173.8 110.1 172.0 550 174.6 111.9 172.0 560 175.4 113.7 172.0 570 176.1 115.4 172.0 580 176.9 117.0 172.0 590 177.6 118.6 172.0 600 178.1 120.2 172.0 610 178.6 121.7 172.0 620 179.0 123.1 172.0 630 179.4 124.5 172.0 640 179.8 125.9 172.0 650 180.2 127.2 172.0 660 180.7 128.5 172.0 670 181.1 129.8 172.0 680 181.5 131.1 172.0 690 181.9 132.3 172.0 700 182.3 133.4 172.0 710 182.7 134.6 172.0 720 183.1 135.7 172.0 730 183.5 136.8 172.0 740 183.9 137.9 172.1 750 184.2 139.0 172.9 760 184.6 140.0 173.8 770 185.0 141.0 174.6 B-22

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9 UPPER BOTTOMa

• 20 EFPY PRESSURE VESSEL CURVE C HEAD CURVE C .~BELTLINE CURVE C (PSIG)j ('F) ('F) (QF) 780 185.4 142.0 175.4 790 185.8 143.0 176.2 800 186.1 143.9 177.0 810 186.5 144.9 177.8 820 186.9 145.8 178.6 830 187.2 146.7 179.4 840 187.6 147.6 180.1 850 187.9 148.4 180.9 860 188.3 149.3 181.6 870 188.6 150.1 182.3 880 189.0 151.0 183.0 890 189.3 151.8 183.7 900 189.7 152.6 184.4 910 190.0 153.14 185.1 920 190.4 154.1 185.7 930 190.7 154.9 186.4 940 191.0 155.7 187.0 950 191.4 156.4 187.7 960 191.7 157.1 188.3 970 192.0 157.9 189.0 980 192.4 158.6 189.6ý 990 192.7 159.3 190.2 1000 193.0 160.0 190.8 1010 193.3 160.6 191.4 1020 193.6 161.3 192.0 1030 194.0 162.0 192.6 1040 194.3 162.6 193.1 B-23

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-4. LaSalle Unit I 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 BELT~LINE ~CURVE C (PSIG) ('F) 1050 194.6 163.3 193.7 1060 194.9 163.9 194.3 1070 195.2 164.5 194.8 1080 195.5 165.2 195.4 1090 195.8 165.8 195.9 1100 196.1 166.4 196.5 1105 196.3 166.7 196.7 1110 196.4 167.0 197.0 1120 196.7 167.6 197.5 1130 197.0 168.2 198.1 1140 197.3 168.7 198.6 1150 197.6 169.3 199.1 1160 197.9 169.9 199.6 1170 198.2 170.4 200.1 1180 198.5 171.0 200.6 1190 198.7 171.5 201.1 1200 199.0 172.1 201.6 1210 199.3 172.6 202.0 1220 199.6 173.2 202.5 1230 199.9 173.7 203.0 1240 200.2 174.2 203.5 1250 200.4 174.7 203.9 1260 200.7 175.2 204.4 1270 201.0 175.7 204.8 1280 201.2 176.2 205.3 1290 201.5 176.7 205.7 1300 201.8 177.2 206.2 B-24

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-4. LaSalle Unit 1 P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curve C For Figure 5-9

<~UPPER ~BOTTOM 2EFPY 20~

PRESSURE <VESSEL CURVE C HEAD C URVE C BELTLINE CURVE C (PSIG) ('F) ('F) 1310 202.1 177.7 206.6 1320 202.3 178.2 207.1 1330 202.6 178.6 207.5 1340 202.8 179.1 207.9 1350 203.1 179.6 208.3 1360 203.4 180.0 208.8 1370 203.6 180.5 209.2 1380 203.9 180.9 209.6 1390 204.1 181.4 210.0 1400 204.4 181.8 210.4 B-25

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite 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 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER.RP.V.&

HEAD BELTLINE AT. 9HEAD. .BELTLINE ATh 20 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B (PSIG) 0

( F) ('F) ('F)  ; ('F) 0 68.0 72.0 68.0 72.0 10 68.0 72.0 68.0 72.0 20 68.0 72.0 68.0 72.0 30 68.0 72.0 68.0 72.0 40 68.0 72.0 68.0 72.0 50 68.0 72.0 68.0 72.0 60 68.0 72.0 68.0 72.0 70 68.0 72.0 68.0 72.0 80 68.0 72.0 68.0 72.0 90 68.0 72.0 68.0 72.0 100 68.0 72.0 68.0 72.0 110 68.0 72.0 68.0 72.0 120 68.0 72.0 68.0 72.0 130 68.0 72.0 68.0 74.2 140 68.0 72.0 68.0 77.4 150 68.0 72.0 68.0 80.2 160 68.0 72.0 68.0 82.9 170 68.0 72.0 68.0 85.5 180 68.0 72.0 68.0 87.9 B-26

GE Nuclear Energy . GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 oF/hr for Curve A FOR FIGURES 5-13 and 5-14 BOTTOM* UPPER RPV & BOTTOM SUPPER RPV &

HEAD BELTLINE AT HEAD BELTIIIN EAT 20 EFPY 20 EFPY iPRESSURE CURVE A CURVE A CURVE B CURVE B (PSIG. ('F) ('F) 190 68.0 72.0 (8.0 68.0 90.2 200 68.0 72.0 68.0 92.3 210 68.0 *72.0 68.0 94.3 220 68.0 72.0 68.0 96.3 230 68.0 72.0 68.0 98.1 240 68.0 72.0 68.0 99.9 250 68.0 72.0 68.0 101.6 260 68.0 72.0 68.0 103.2 270 68.0 72.0 68.0 104.8 280 68.0 72.0 68.0 106.3 290 68.0 72.0 68.0 107.8 72.0 300 68.0 68.0 109.2 72.0 310 68.0 68.0 110.5 72.0 312.5 68.0 68.0 110.9 102.0 312.5 68.0 68.0 132.0 102.0 320 68.0 68.0 132.0 102.0 330 68.0 68.0 132.0 102.0 340 68.0 68.0 132.0 102.0 350 68.0 68.0 132.0 102.0 360 68.0 68.0 132.0 102.0 370 68.0 68.0 132.0 B-27

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite 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 5-13 and 5-14 BOITTOM ~UPPER RPV & BOTTOM UPPER RI-'V&

HEAD BELTLINE AT HEAD BELTLIINE AT 20 EFPY~ 20 EFPY PRESSURE CURVE A CURVE A CURVE B. CURVE B (PSIG) ( 0F) >('F) 380 68.0 102.0 132.0 68.0 390 68.0 102.0 132.0 68.0 400 68.0 102.0 132.0 68.0 410 68.0 102.0 132.0 68.0 420 68.0 102.0 132.0 68.0 430 68.0 102.0 68.0 132.0 68.0 440 68.0 102.0 68.0 132.0 450 68.0 102.0 68.0 132.0 460 68.0 102.0 132.0 69.6 470 68.0 102.0 132.0 72.1 480 68.0 102.0 132.0 74.4 490 68.0 102.0 132.0 76.6 500 68.0 102.0 132.0 78.8 510 68.0 102.0 132.0 80.8 520 68.0 102.0 132.2 82.8 530 68.0 102.0 133.0 84.7 540 68.0 102.0 133.8 86.5 550 68.0 102.0 134.6 560 68.0 102.0 88.3 135.4 570 68.0 102.0 90.0 136.1 580 68.0 102.0 91.6 136.9 B-28

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite 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 5-13 and 5-14

~BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT~ HEAD BELTLINE AT 2OEFPY 20 EFPY PRESSURE CURVE A >CURVE A CURVE B JCURVE B (PSIG) 0

( F) 102.

590 68.0 93.2 137.6 102.0 600 68.0 94.8 138.1 102.0 610 68.0 96.3 138.6 102.0 620 68.0 97.7 139.0 102.0 630 68.0 99.1 139.4 102.0 640 68.0 100.5 139.8 102.0 650 68.0 101.8 140.2 102.0 660 68.0 103.1 140.7 102.0 10203 670 68.0 104.4 141.1 102.0 680 68.0 105.7 141.5 102.0 690 68.0 106.9 141.9 102.0 700 68.0 108.0 142.3 102.0 710 68.7 102.0 109.2 142.7 720 70.1 102.0 110.3 143.1 730 71.5 111.4 143.5 103.1 740 72.8 112.5 143.9 104.0 750 74.1 113.6 144.2 104.8 760 75.4 114.6 144.6 105.6 770 76.6 115.6 145.0 106.3 780 77.8 116.6 145.4 107.1 790 79.0 117.6 145.8 I B-29

GE Nuclear Energy GE-N E-0000-0003-5526-02Rl1a Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite 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 5-13 and 5-14

. BOTTOMJ UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD. ~BELTLINE AT 20 EFPY 20OEFPY PRESSURE CURVE A CURVE A CURVE B CURVE B (PSIG) (F o)( )(F 800 80.2 107.9 118.5 146.1 810 81.3 108.6 119.5 146.5 820 82.4 109.4 120.4 146.9 830 83.5 110.1 121.3 147.2 840 84.5 110.8 122.2 147.6 850 85.6 111.5 123.0 147.9 860 86.6 112.2 123.9 148.3 870 87.6 112.9 124.7 148.6 880 88.5 113.6 125.6 149.0 890 89.5 114.3 126.4 149.3 900 90.4 114.9 127.2 149.7 910 91.4 115.6 128.0 150.0 920 92.3 116.2 128.7 150.4 930 93.1 116.9 129.5 150.7 940 94.0 117.5 130.3 151.0 950 94.9 118.1 131.0 151.4 960 95.7 118.7 131.7 151.7 970 96.6 119.3 132.5 152.0 980 97.4 119.9 133.2 152.4 990 98.2 120.5 133.9 152.7 1000 99.0 121.1 134.6 153.0 B-30

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite P-T Curve Values for 20 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 0 F/hr for Curve A FOR FIGURES 5-13 and 5-14 BOTTOM U R RPV & BOTTOM UPPER RPV*&

HEA BELTLINE AT < HEAD ~BELTLINE AT LA~~2 ~,~EFPYL 2EFPY, 20 PRESSURE <CURVE AN N CURVE A CURVEB CURVEB (PSIG) (')('F) ('F)y ~ ('F) 1010 99.7 121.7 135.2 153.3 1020 100.5 122.2 135.9 153.6 1030 101.3 122.8 136.6 154.0 1040 102.0 123.4 137.2 154.3 1050 102.7 124.0 137.9 154.6 1060 103.4 124.7 138.5 154.9 1070 104.2 125.5 139.1 155.2 1080 104.9 126.2 139.8 155.5 1090 105.6 127.0 140.4 155.9 1100 106.2 127.7 141.0 156.5 1105 106.6 128.0 141.3 156.7 1110 106.9 128.4 141.6 157.0 1120 107.6 129.1 142.2 157.5 1130 108.2 129.8 142.8 158.1 1140 108.9 130.5 143.3 158.6 1150 109.5 131.1 143.9 159.1 1160 110.1 131.8 144.5 159.6 1170 110.8 132.4 145.0 160.1 1180 111.4 133.1 145.6 160.6 1190 112.0 133.7 146.1 161.1 1200 112.6 134.4 146.7 161.6 B-31

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version TABLE B-5. LaSalle Unit 1 Composite 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 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AlT HEAD: BELTLINE AT 20 EFPY 20 EFPY PRESSURE CURVE A CURVE A CURVE B CURVEB6 (PIG)> ('F) ('F) ('F) 1210 113.2 135.0 147.2 162.0 1220 113.8 135.6 147.8 162.5 1230 114.3 136.2 148.3 163.0 1240 114.9 136.8 148.8 163.5 1250 115.5 137.4 149.3 163.9 1260 116.0 138.0 149.8 164.4 1270 116.6 138.6 150.3 164.8 1280 117.1 139.1 150.8 165.3 1290 117.7 139.7 151.3 165.7 1300 118.2 140.3 151.8 166.2 1310 118.7 140.8 152.3 166.6 1320 119.3 141.4 152.8 167.1 1330 119.8 141.9 153.2 167.5 1340 120.3 142.4 153.7 167.9 1350 120.8 143.0 154.2 168.3 1360 121.3 143.5 154.6 168.8 1370 121.8 144.0 155.1 169.2 1380 122.3 144.5 155.5 169.6 1390 122.8 145.0 156.0 170.0 1400 123.3 145.5 156.4 170.4 B-32

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version APPENDIX C Operating And Temperature Monitoring Requirements C-1

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the-recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

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

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

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

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

C-2

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the cooiant temperature is changing by _<20°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-0000-0003-5526-02Rla Non-Proprietary Version 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-0000-0003-5526-02R1 a Non-Proprietary Version APPENDIX D GE SIL 430 D-1

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions concerning primary and alternate reactor pressure vessel (RPV) temperature monitoring measurements for complying with RPV brittle fracture and thermal stress requirements.

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

TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)

Measurement 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 I 00°F/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 212OF for Tech Must comply with SIL 251 Spec 1OOOF/hr heatup to avoid vessel stratification.

and cooldown rate.

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

D-2

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

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

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

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

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

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

outside metal surface Should have drain line flow.

temperatures.

D-3

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 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 BWR/6s.

D-4

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version 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-0000-0003-5526-02Rla Non-Proprietary Version 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, ifany, of information contained in SILs to any plant or facility other than GE BWRs as designed and furnished by GE. Determination of applicability of information contained in any SIL to a specific GE BWR and implementation of recommended action are responsibilities of the owner of that GE BWR.SILs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of its owner. Such operation involves activities of which GE has no knowledge and over which GE has no control. Therefore, GE makes no warranty or representation expressed or implied with respect to the accuracy, completeness or usefulness of information contained in SILs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SILs.

D-6

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version APPENDIX E Determination of Beltline Region and Impact on Fracture Toughness E-1

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 10CFR50, Appendix G defines the beltline region of the reactor vessel as follows:

"The region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage" To establish the value of peak fluence for identification of beltline materials (as discussed above), the 10CFR50 Appendix H fluence value used to determine the need for a surveillance program was used; the value specified is a peak fluence (E>1 MEV) of 1.Oel 7 n/cm 2 . Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm 2, then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.

The following dimensions are obtained from the referenced drawings:

Shell # 3 - Top of Active Fuel (TAF): 366.31" (from vessel 0) [1]

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

Bottom of LPCI Nozzle in Shell # 3: 355.6" (from vessel 0) [2]

Center line of LPCI Nozzle in Shell # 3: 372.5" (from vessel 0) [3]

Top of Recirculation Outlet Nozzle in Shell # 1: 197.188" (from vessel 0) [4]

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

Top of Recirculation Inlet Nozzle in Shell # 1: 197.688" (from vessel 0) [4]

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

As shown above, the LPCI nozzle is within the core beltline 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 -19" 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-0000-0003-5526-02Rla Non-Proprietary Version fluence at this location is less than 1.0e17 n/cm 2, it can be safely concluded that all nozzles (other than the LPCI nozzle) are outside the beltline region of the reactor vessel.

Based on the axial flux profile [5], the RPV flux level at -10" below the BAF dropped to less than 0.1 of the peak flux 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 1.02e18 n/cm 2 [5], fluence at -10" below BAF and -10" above TAF are expected to be less than 1.0e17 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 1 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-0000-0003-5526-02Rla Non-Proprietary Version APPENDIX E

REFERENCES:

1. CornEd Nuclear Design Information Transmittal (NDIT) No. LS-1169, "Pressure-Temperature Curves", 12/10/99.

2. CE Drawing #232-863, Revision 4, "Nozzle Details for 251" I.D. BWR", (GE VPF #2029-099, Revision 7).

3. CE Drawing #232-788, Revision 3, "General Arrangement Elevation for 251" I.D. BWR" (GE VPF #2029-117, Revision 4).

4. CE Drawing #232-879, Revision 3, "Nozzle Details for 251" I.D. BWR", (GE VPF #2029-092, Revision 6).

5. 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).

E-4

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version APPENDIX F EVALUATION FOR UPPER SHELF ENERGY (USE)

F-1

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 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-1.

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 70-80%, 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 60 ft-lb (for vertical weld heat 1P3571).

F-2

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version Table F-I: Upper Shelf Energy Evaluation for LaSalle Unit 1 Beltline Materials Initial Initial 32 EFPY Test Longitudinal Transverse 114T  % Decrease 32 EFPY Location Heat Temperature USE USEa %Cu Fluence USEb Transverse USEC (lF) (ft-lb) (ft-lb) (n/cm ) (ft-lb)

Plates:

Lower C5978-1 160 136 88.4 0.11 7.1E+17 11 79 C5978-2 160 120 78 0.11 7.1E+17 11 69 C5979-1 160 136 88.4 0.12 7.1E+17 11.5 78 Lower-Intermediate C6345-1 d 160 165 107.3 0.15 7.1E+17 13 93 C6318-1 160 140 91 0.12 7.1E+17 11.5 81 C6345-2 160 161 104.7 0.15 7.1E+17 13 91 Middle A5333-1 160 155 100.8 0.12 7.1E+17 11.5 89 B0078-1 160 151 98.2 0.15 7.1E+17 13 85 C6123-2 160 151 98.2 0.13 7.1E+17 12 86 Welds:

Vertical:

3-308 305424 10 92 0.273 7.1E+17 23 71 1P3571 10 79 0.283 7.1E+17 23.5 60 4-308 305414' 10 92 0.337 7.1E+17 26.5 68 305414' 10 92 0.286 7.1E+17 23.5 70 12008W 10 92 0.235 7.1E+17 21 73 1_2008 10 92 0.286 7.1E+17 23.5 70 2-307 21935e 10 97 0.183 7.1E+17 18 80 2110_93_57 97 0.213 7.1E+17 19.5 78 12008' 10 97 0.235 7.1E+17 21 77 12008! 10 97 0.213 7.1E+17 19.5 78 Girth:

6-308 6329637 10 103 0.205 7.1E+17 19 83 1-313 4P6519 10 116 0.131 7.1E+17 15 99 Forg ingls:

LPCI Nozzle Q2022W9I 10 1 73 0.10 1.7E+171 7.5 68 a Values obtained from [31 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 The initial transverse USE value is 65% of the highest 160°F data from CMTRs e Single Wire f Tandem Wire g Average of Charpy V-Notch data for %Shear >= 70 F-3

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 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).

F-4

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version APPENDIX G THICKNESS TRANSITION DISCONTINUITY EVALUATION G-1

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version Objectives:

The purpose of the following evaluations is to determine the hydrotest and the heat-up/cool-down temperature (T) for the shell thickness transition discontinuity and to demonstrate that the temperature is bounded by the beltline hydrotest and heat-up/cool-down temperature.

Methods and Assumptions:

An ANSYS finite element analysis was performed for the thickness discontinuity in the beltline region of LaSalle Unit 1. The purpose of this evaluation was to determine the RPV discontinuity stresses (hoop and axial) that result from a thickness transition discontinuity in the beltline region. The transition is modeled as a transition from 6 1/8" minimum thickness to 7 1/8" minimum thickness [1].

Three load cases were evaluated for the beltline shell discontinuity: 1) hydrostatic test pressure at 1563 psig, 2) a cool-down transient of 100°F/hr, starting at 550°F and decreasing to 70°F on the inside surface wall [2] and with an initial operating pressure of 1050 psig, and 3) a heat-up transient of 100°F/hr, starting at 70°F and increasing to 550°F on the inside surface wall [2] and with a final operating pressure of 1050 psig. For both transient cases it was assumed that the outside RPV wall surface is insulated with a heat transfer coefficient of 0.2 BTU/hr-ft2 OF [3] and that the ambient temperature is 100 0 F.

These are the bounding beltline transients of those described in Table 5.2-4 of the LaSalle Unit 1 and 2 UFSAR and Region B of the thermal cycle diagram [2] at temperatures for which brittle fracture could occur. Material properties were used from the Code of construction for the RPV Materials: Shell Plate Materials are ASME SA533, Grade B, Class 1, low alloy steel (LAS) [4].

Methods consistent with those described in Section 4.3 were used to calculate the T-RTNDT for the shell discontinuity for a hydrotest pressure of 1563 psig and the two transient cases.

The adjusted reference temperature values shown in Table 4-4 were added to the T-RTNDT to determine the temperature "T". The value of "T" was compared to that of the beltline G-2

GE Nuclear Energy GE-NE-0000-0003-5526-02R1 a Non-Proprietary Version region for the same condition as described in Sections 4.3.2.2.1 for the hydrotest pressure case and 4.3.2.2.4 for the transient cases.

As shown below the stresses that result from the transition discontinuity are not significantly greater than those remote from the discontinuity (the difference in stress is less than 1 ksi for the pressure case and less than 2 ksi for the thermal cases). Therefore, the shell transition discontinuity stresses are also bounded by the beltline shell calculation.

The methods of ASME Code Section Xl, Appendix G [5] are used to calculate the pressure test and thermal limits. The membrane and bending stress were determined from the finite element analysis and are shown below. The hoop stresses were more limiting than the axial stresses.

The stress intensity factors, Kim and Kib, are calculated using Code Case N-640 [6], and ASME Code Section XI Appendix A [7] and Appendix G [5]. Therefore, Kim= Mm*crm and Kib

= Mb*Cob. The values of Mm and Mb were determined from the ASME Code Appendix G [5].

The stress intensity is based on a 1/4 T radial flaw with a six-to-one aspect ratio (length of 1.5T). The flaw is oriented normal to the maximum stress direction, in this case a vertically oriented flaw since the hoop stress was limiting.

The calculated value of Kim + KIb is multiplied by a safety factor (SF) (1.5 for pressure test and 2.0 for the transient cases), per ASME Appendix G [5] for comparison with KIR, the material fracture toughness expressed as KIc.

The relationship between K~c and temperature relative to reference temperature (T - RTNDT) is provided in ASME Code Section XI Appendix A [7] Paragraph A-4200, represented by the relationship (K1 units ksi-in 0.5):

Ic = 33.2 + 20.734 exp[0.02 (T - RTNDT)]; therefore, T-RTNDT = ln[(Kjc-33.2)/20.734]/0.02, G-3

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version where K1c = SF* (KIm + Kqb) for pressure test and Kic = (SF

• Kip) + Kis for transient case.

This relationship is derived in the Welding Research Council (WRC) Bulletin 175 [8] as the lower bound of all dynamic fracture toughness data. This relationship provides values of pressure versus temperature (from KIR and (T - RTNDT), respectively).

The RTNDT is added to the (T-RTNDT) to determine the hydrotest, heat-up, and cool-down temperatures.

Analysis Information:

Thin Section Thickness tmin = 6.13 inch 5

4(t) = 2.47 inch Thick Section Thickness tmi, = 7.13 inch (t)= 2.67 inch. 5 G-4

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Analysis and Results for the Hvdrotest Pressure (Case 1):

Primary Primary membrane bending Km = Mb = Kb =

Pressure Pm Pb Mm Mm*Pm 2/3 Mm Mb*Pb T-RTNDT 1 12 2 (psig) (psi) (psi) (psi in ) (psi in"1 ) (OF)

Maximum Hoop Stress - Adjacent to the discontinuity in thin section (6.125")

1000 20920 475 I I 1000 17820 460 1563 27853 719 2.47 68870 1.65 1185 62.2 Note that the axial stress is approximately 1/2 of the hoop stress.

Results and

Conclusions:

The maximum LaSalle Unit 1 plant-specific T-RTNDT for the thickness discontinuity is 68 0 F as shown in the table above. The limiting beltline weld material RTNDT at the region of the discontinuity is 88°F (see Table 4-4), so T = 156 0 F. The limiting beltline plate RTNDT at the region of the discontinuity is 61OF (see Table 4-4), so T = 1290 F.

At 1563 psig, Curve A is limited by the beltline curve. The T- RTNDT for the beltline region Curve A is 81°F at 1563 psig, and T = 1697F.

Because the beltline region pressure test temperature "T" of 169°F bounds the limiting plant-specific thickness discontinuity for the case with the limiting ART value (T = 156°F for the weld material in the region of the discontinuity), the thickness discontinuity remains bounded by the beltline curve.

G-5

GE Nuclear Energy GE-N E-0000-0003-5526-02R la Non-Proprietary Version Analysis and Results for Cool-down (CD - Case 2) and Heat-up (HU - Case 3):

Hoop Stress G-6

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version Results and

Conclusions:

The maximum LaSalle Unit 1 plant-specific T-RTNDT for the thickness discontinuity is 72 0 F. The limiting beltline material RTNDT in the region of the discontinuity is 88 0 F (see Table 4-4), so T = 1'60 0 F. The limiting beltline plate RTNDT inR.the region of the discontinuity is 61OF (see Table 4-4), so T = 133 0 F.

At 1050 psig, Curve B is limited by the beltline curve. The T- RTNDT for the beltline region is 82 0 F at 1050 psig, and T = 170 0 F.

Because the beltline region pressure test temperature "T" of 170°F bounds the limiting plant-specific thickness discontinuity for the case with the limiting ART value (T = 160'F for the weld material in the region of the discontinuity), the thickness discontinuity remains bounded by the beltline curve.

G-7

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version Appendix G

References:

1. RPV Drawings a) CE Drawing # 232-788, Rev. 3, "General Arrangement Elevation for 251" I.D. BWR," (GE VPF #2029-117, Rev. 4).

b) CE Drawing # 232-790, Rev. 8, "Lower Vessel Shell Assembly Machining

& Welding for 251" I.D. BWR," (GE VPF #2029-036, Rev. 8).

c) CE Drawing # 232-791, Rev. 15, "Upper Vessel Shell Assembly Machining & Welding for 251" I.D. BWR," (GE VPF #2029-037, Rev. 14).

2. GE Drawing Number 731E776, "Reactor Vessel Thermal Cycles", GE-NED, San Jose, CA, Revision 3 (GE Proprietary).

3. "Reactor Vessel Purchase Specification, Reactor Pressure Vessel", (21A9242AF, Revision 9), December 1975.

4. T.A. Caine, "LaSalle Unit 1 RPV Surveillance Materials Testing and Analysis",

(GE-NE-523-A166-1294, Revision 1), June 1995.

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

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

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

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

Welding Research Council Bulletin 175, August 1972.

G-8

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version APPENDIX H CORE NOT CRITICAL CALCULATION FOR BOTTOM HEAD (CRD PENETRATION)

H-1

GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version TABLE OF CONTENTS The following outline describes the contents of this Appendix:

H.1 Executive Summary H.2 Scope H.3 Analysis Methods H.3.1 Applicability of the ASME Code Appendix G methods H.3.2 Finite Element Fracture Mechanics Evaluation H.3.3 ASME Code Appendix G Evaluation H.4 Results H.5 Conclusions H.6 References H.1 Executive Summary This Appendix describes the analytical methods used to determine the T-RTNDT value applicable for the Bottom Head Core Not Critical P-T curves. This evaluation uses new finite element fracture mechanics technology developed by the General Electric Company, which is used to augment the methods described in the ASME Boiler and Pressure Vessel Code [Reference 1]. ((

)) This method more accurately predicts the expected stress intensity ((

)) The peak stress intensities for the pressure and thermal load cases evaluated are used as inputs into the ASME Code Appendix G evaluation methodology to calculate a T-RTNDT. ((

H-2

GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version H.2 Scope This Appendix describes the analytical methods used to determine the T-RTNDT value applicable for the Bottom Head Core Not Critical P-T curves. This evaluation uses new finite element fracture mechanics technology developed by the General Electric Company which is used to augment the methods described in the ASME Boiler and Pressure Vessel Code [Reference 1]h This Appendix discusses the finite element .

analysis and the Appendix G [Reference 1] calculations separately below.

H.3 Analysis Methods This section contains technical descriptions of the analytical methods used to perform the BWR Bottom Head fracture mechanics evaluation. The applicability of the current ASME Code,Section XI, Appendix G methods [Reference 1] considering the specific bottom head geometry is discussed first followed by a detailed discussion of the finite element analysis and Appendix G evaluation [Reference 1].

H.3.1 Applicability of the ASME Code Appendix G Methods The methods described in the ASME Code Section X1, Appendix G [Reference 1] for demonstrating sufficient margin against brittle fracture in the RPV material are based upon flat plate solutions which consider uniform stress distributions along the crack tip.

The method also suggests that a 1/4 wall thickness semi-elliptical flaw with an aspect ratio of 6:1 (length to depth) be considered in the evaluation. When the bottom head specific geometry is considered in more detail the following items become evident:

Noting these items, the applicability of the methods suggested in Appendix G ((

)). The ASME Code does not preclude using other methods; therefore, a more detailed 3-D finite element fracture mechanics analysis ((

H-3

GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version I]

was performed. The stress intensity obtained from this analysis is used in place of that determined using the Appendix G methods [Reference 1].

H.3.2 Finite Element Fracture Mechanics Evaluation An advanced (( )) finite element analysis of a BWR bottom head geometry

[1 was performed to determine the mode I stress intensity at the tip of a %thickness postulated flaw. ((

Finite Elements ((

All Finite Element Analyses were done using ANSYS Version 6.1 [Reference 2]. ((

))

Structural Boundary Conditions The modeled geometry is one-fourth of the Bottom Head hemisphere so symmetry boundary conditions are used. ((,..

)) The mesh is shown in Figure 1.

H-4

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version IT Material Properties Two materials are used as per the ASME Code. Material 1 is SA533 which is used to model the vessel. Material 2 ((

)) The ANSYS listing of these materials in (pound-inch-second-°F) units are:

H-5

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version 1]

EX is the Young's Modulus, NUXY is the Poisson's Ratio, ALPX is the Thermal Expansion Coefficient, DENS is the Density, KXX is the Thermal Conductivity and C is the Heat Capacity.

H-6

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version Loads Two loads cases were independently analyzed.

1. Pressure Loading -

An internal pressure of 1250 PSI is applied to the interior of the vessel ((

)) In addition, the thin cylindrical shell stress due to this pressure is applied as a blowoff pressure (( )) at the upper extremity of the vertical wall of the BWR. Figure 2 shows these loads. ((

Figure 2. Pressure Loads

2. (( )) Thermal Transient -

I]

Thermal loads are applied to the model as time dependent convection coefficients and bulk temperatures. Referring to the regions identified in Figure 3, the corresponding values follow. Convection coefficients (h) are in units of BTU/(hr-ft-°F) and temperatures (T) are in OF.

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GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version h,

h4.

Figure 3. Regions to which thermal loads are applied

a. Region 1: h = 25, T = 60

b. Regions 2 and 3:

Time (min) h2 h3 T 0 496 413 (( ]

341 354 (( ))

R )) 496 413 [R ))

1] 496 413 (( 1]

[R 1]

Temperature Plot vs. Time (min.)

c. Region 4: Adiabatic (exaggerated in size in drawing)

d. Region5: h = 0.2, T = 100 The peak thermal gradients were used to compute the thermal stresses based on a uniform reference temperature of 70 OF.

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GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Crack Configurations The following four cracks were analyzed:

1. A part through crack, % of the vessel wall thickness deep, measured from inside the vessel, ((

))

2. Same as 1, but depth is measured from outside the vessel

3. Same as 1, []

4. Same as 2, [

The cracks considered for this analysis ((

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version 1))

))]

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GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version 1[

))

Er

))

1]

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version Stress Intensity Factor Computation

((1 H-12

GE Nuclear Energy GE-NE-0000-0003-5526-02R1a Non-Proprietary Version

((

((

1]

Benchmarkinq (( )) Methodology

((

)) The results of these benchmarking studies have demonstrated the accuracy of this method used for this evaluation.

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GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version Pressure LoadinQ Analysis Results I]

II I]

1]

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version Benchmarkinq of Pressure Loading Results Pressure Loading analyses ((

))

1]

1]

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version R]

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GE Nuclear Energy GE-NE-0000-0003-5526-02Rla Non-Proprietary Version

))]

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GE Nuclear Energy GE-N E-0000-0003-5526-02R 1 a Non-Proprietary Version Thermal Transients Analysis Results For the thermal transient considered, the inner diameter of the vessel is hotter than the outer diameter; hence, the I.D. cracks, (( )), close due to the thermal gradient and result in negative Stress Intensity Factors, which is not critical.

However, the O.D. cracks open [ )). All results for the thermal transient will consequently be shown for the O.D. (( ))

crack.

In order to identify the peak gradient, three locations were chosen. ((

)) Thermal Gradients (( ))

Figure 10a is a plot of these three gradients vs. time. Figure 10b. is zoomed in to the peaking region.

))

I]

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GE Nuclear Energy GE-N E-0000-0003-5526-02R i a Non-Proprietary Version 1]

It can be seen that the peak times and values based on each gradient are:

Gradient Peak Time (Min.) Peak Value (OF)

Stress analyses were performed using the temperature distributions obtained from the thermal analyses at each of these peak times and the Stress Intensity Factors are shown in Figure 11.

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GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version

((I H.3.3 ASME Code Appendix G Evaluation The peak stress intensities for the pressure and thermal load cases evaluated above are used as inputs into the ASME Code Appendix G evaluation methodology [Reference 1]

to calculate a T-RTNDT. The Core Not Critical Bottom Head P-T curve T-RTNDT is calculated using the formulas listed below:

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GE Nuclear Energy GE-N E-0000-0003-5526-02R1 a Non-Proprietary Version KI = SFp.K1 p + SFT-KIt SF =2.0 p

SFt= 1.0 T - RTN =In (K -33.2)§ 113-(...73 .02 1 Where: KI is the total mode I stress intensity, Kip is the pressure load stress intensity, Kit is the thermal load stress intensity, SFp is the pressure safety factor, SFt is the thermal safety factor, 2

Note that the stress intensity is defined in units of: ksi*inlI H.4 Results Review of the (( )) results above demonstrates that the OD (( ))

crack exhibits the highest stress intensity for the considered loading. The T-RTNDT to be used in the Core Not Critical Bottom Head P-T curves shall be calculated using the stress intensities obtained at this location. The calculations are shown below:

Note that the pressure stress intensity has been adjusted by the factor (( )) to account for the vessel pressure at which the maximum thermal stress occurred. The H-21

GE Nuclear Energy GE-N E-0000-0003-5526-02R 1a Non-Proprietary Version finite element results summarized above were calculated using a vessel pressure [

1]

Comparing the T-RTNDT calculated using the methods described above to that determined using the previous GE methodology, ((

1]

H.5 Conclusions For the (( )) transient, the appropriate T-RTNDT for use in determining the Bottom Head Core Not Critical P-T curves (( )). Existing Bottom Head Core Not Critical curves developed using the previous GE methodology ((

1]

H.6 References

1. American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PV Code), Section X1. 1998 Edition with Addenda to 2000.

I1. ANSYS User's Manual, Version 6.1.

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