ML043360217

GE-NE-0000-0002-9600-03R3a, Pressure-Temperature Curves for Exelon Quad Cities Unit 2, Revision 2

ML043360217
Person / Time
Site:	Quad Cities
Issue date:	05/31/2004
From:	Tilly L General Electric Co
To:	Office of Nuclear Reactor Regulation
References
DRF 0000-0028-4024, GE-NE-0000-0002-9600-03R3a
Download: ML043360217 (188)

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GE NuclearEnergy Engineering and Technology GE-NE-0000-0002-9600-03R2a General Electric Company DRF 0000-0028-4024 175 Curtner Avenue Revision 2 San Jose, CA 95125 Class I May 2004 Pressure-Temperature Curves For Exelon Quad Cities Unit 2 Prepared by: L qua L.J. Tilly, Senior Engineer Structural Analysis & Hardware Design Verified by: OD (Frew B.D. Frew, Principal Engineer Structural Analysis & Hardware Design Approved by: 037 (1Branfund B.J. Branlund, Principal Engineer Structural Analysis & Hardware Design

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version REPORT REVISION STATUS Revision Purpose 0 Initial Issue 1 The report was revised to eliminate conservatism in the upper vessel P-T curve contained in the Appendix G P-T curves. The required bounding shift for this curve was reduced from 550 F to 480 F. The bounding P-T curve figures and tables contained in Appendix G were revised. This includes Figures G-2, G-6, and G-9 through G-14.

2

• 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.

• Discussion regarding a comparison of calculated flux vs. dosimetry results has been deleted from Section 4.2.1.2.

• The description of the transients considered in Section 4.3.2.1 has been revised.

• 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 F provides 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 K, in the Beltline Core Not Critical P-T curves.

• Reference 14b has been deleted; all required information from this reference is contained in Reference 14a. Reference 14c has been renumbered to Reference 14b.

• Reference 22 has been deleted and replaced with the NRC Safety Evaluation for the Dresden P-T curve reports.

• Reference 6 of Appendix E has been deleted; all required information from this reference is contained in Reference 5.

• Appendix F has been deleted and replaced with a discussion regarding the CRD (Bottom Head) Core Not Critical evaluation.

• Section 5.0 Figures 5-5, 5-12 and 5-14, Appendix G Figures G-5, G-12, and G-14 and all Appendix B and Appendix G Tables have been revised to incorporate changes to the CRD Penetration (Bottom Head) Core Not Critical P-T curves, as defined in Section 4.3.2.1.2 and Appendix F.

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GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version IMPORTANT NOTICE This is a non-proprietary version of the document GE-NE-0000-0002-9600-03R2, 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, Interim Power Uprate for Dresden and Quad Cities, effective 6/30/00, 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, 2003

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GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 curves in this report represent 32 and 54 EFPY as determined for a 40- and 60-year life. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Case N-640, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kic 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. Additional detail regarding P-T curve methodology has been submitted in response to NRC questions, and was accepted in the Safety Evaluation for the Dresden Units 2 & 3 P-T curves 122].

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

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

• Closure flange region (Region A)

• Core beltline region (Region B)

• Upper vessel (Regions A & B)

• Lower vessel (Regions B & C)

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 100'F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20'F/hr or less must be maintained at all times.

The P-T curves apply for both heatup and cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the I/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 32 and 54 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 32 and 54 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

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GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 13 4.3 PRESSURE-TENMEPERATURE CURVE METHODOLOGY 19

5.0 CONCLUSION

S AND RECOMMENDATIONS 50

6.0 REFERENCES

67

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GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 CORE NOT CRITICAL CALCULATION FOR BOTTOM HEAD (CRD PENETRATION)

APPENDIX G BOUNDING P-T CURVES FOR QUAD CITIES UNITS 1&2

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GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF QUAD CITIES UNIT 2 RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2: CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 31 FIGURE 4-3: FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 37 FIGURE 5-I: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20F/IHR OR LESS COOLANT HEATUP/COOLDOWN] 53 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [20 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 54 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 32 EFPY [201F/HR OR LESS COOLANT HEATUP/COOLDOWN] 55 FIGURE 5-4: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 54 EFPY 120 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100F/HR OR LESS COOLANT HEATUP/COOLDOWN] 57 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] 1100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 58 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 32 EFPY

[I00F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-8: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 54 EFPY

[100F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 FIGURE 5-9: CORE CRITICAL P-T CURVES [CURVE C] UP TO 32 EFPY [100 0 F/HR OR LESS COOLANTHEATUP/COOLDOWN] 61 FIGURE 5-10: CORE CRITICAL P-T CURVES [CURVE C] UP TO 54 EFPY [l00F/HR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-1 1: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 32 EFPY [20 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 63 FIGURE 5-12: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 32 EFPY

[1 000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64 FIGURE 5-13: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 54 EFPY [20TF/HR OR LESS COOLANT HEATUP/COOLDOWN] 65 FIGURE 5-14: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 54 EFPY I 100F/HR OR LESS COOLANT HEATUP/COOLDOWN] 66

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GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR QUAD CITIES UNIT 2 VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR QUAD CITIES UNIT 2 NOZZLE AND WELD MATERIALS 12 TABLE 4-3: QUAD CITIES UNIT 2 BELTLINE ART VALUES (32 EFPY) 17 TABLE 4-4: QUAD CITIES UNIT 2 BELTLINE ART VALUES (54 EFPY) 18 TABLE 4-5:

SUMMARY

OF THE IOCFR50 APPENDIX G REQUIREMENTS 21 TABLE 4-6: APPLICABLE BWR/3 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 23 TABLE 4-7: APPLICABLE BWR/3 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-0000-0002-9600-03R2a 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 beitline.

Complete P-T curves were developed for 32 and 54 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. 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.

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], and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of 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 beftline materials.

Additional detail regarding P-T curve methodology has been submitted in response to NRC questions, and was accepted in the Safety Evaluation for the Dresden Units 2 & 3 P-T curves [22].

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 GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version beltline material chemistry. The ART calculation, methodology, and ART tables for 32 and 54 EFPY are included in Section 4.2. The peak ID fluence. values of 3.3 x 10" n/cm2 (32 EFPY) and 5.7 x 1017n/cm2 (54 EFPY) used in this report are discussed in Section 4.2.1.2. Beltline chemistry values are discussed in Section 4.2.1.1.

Comprehensive documentation of the RPV discontinuities that are considered in this report is included in Appendix A. This appendix also includes a table that documents which non-beltline discontinuity curves are used to protect 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 are outside the beltline region. Appendix F provides the core not critical calculation for the bottom head (CRD Penetration). Finally, Appendix G provides a set of P-T curves that bound all requirements for both Quad Cities Unit 1 and Quad Cities Unit 2.

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a 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]. 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, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kc 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 Quad Cities Unit 2 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].

Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize heating requirements prior to pressure testing. Operating and temperature monitoring requirements are found in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles are outside the GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version beltline region. Appendix F provides the core not critical calculation for the bottom head (CRD Penetration). Finally, Appendix G provides a set of P-T curves that bound all requirements for both Quad Cities Unit 1 and Quad Cities Unit 2.

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:

For end-of-license (54 EFPY) fluence, a mixed capacity factor is used to determine the EFPY for a 60-year plant life. An 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).

For a 60-year license an 80% capacity factor is assumed for up to 21 EFPY, and to consider recent improvements in plant operation, a 97.5% capacity factor is used beginning at 21 EFPY; hence, 54 EFPY is assumed to represent 60 years of operation.

The hydrostatic pressure test will be conducted at or below 1105.5 psig.

The shutdown margin, provided in the Quad Cities Unit 2 Technical Specification, is calculated for a water temperature of 680F.

The flux is calculated using a pre-EPU and a post-EPU flux [14], both calculated in accordance with Regulatory Guide 1.190. The pre-EPU flux is applied for 21 EFPY and the post-EPU flux is applied for 11 EFPY and 33 EFPY for 32 and 54 EFPY, respectively.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 1I1,Subsection NB-2300 and are summarized as follows:

a. Test specimens shall be longitudinally oriented CVN specimens.

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

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

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

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

b. RTNDT is defined as the higher of the dropweight NDT or 600 F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion 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-N E-0000-0002-9600-03R2a 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 Quad Cities Unit 2 vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDT calculations for vessel plate, weld, HAZ, and forging, and for bolting material LST are summarized in the remainder of this section.

The RTNDT values for the vessel weld materials (except for the Quad Cities Unit 2 lower shell to lower-intermediate shell weld, shown below) were not calculated; these values were obtained from other sources (see Section 4.2, Tables 4-3 and 4-4).

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 Quad Cities Unit 2 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy tests. The lowest energy Charpy value is adjusted by adding 20F per ft-lb energy difference from 50 ft-lb.

For example, for the Quad Cities Unit 2 beltline plate heat C1722-2 in the lower shell course, the lowest Charpy energy and test temperature from the CMTRs is 35 ft-lb at 100F. The estimated 50 ft-lb longitudinal test temperature is:

T50L = 10F + [(50 - 35) ft-lb

• 20F/ft-lb] = 40 0F The transition from longitudinal data to transverse data is made by adding 300 F to the 50 ft-lb longitudinal test temperature; thus, for this case above, TS0T = 400F + 301F = 700 F GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5o- 600F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is -10 0F. Thus, the initial RTNDT for plate heat C1722-2 is 100F.

For the Quad Cities Unit 2 lower shell to lower-intermediate shell weld, the CVN results were used to calculate the RTNDT. The 50 ft-lb test temperature is applicable to the weld material, but the 300F adjustment to convert longitudinal data to transverse data is not applicable to weld material. Heat S3986 with Flux Lot 3870 has a lowest Charpy energy of 41 ft-lb at 100F as recorded in weld qualification records. Therefore, TsoT = 10F + [(50-41) ft-lb

• 20F/ft-lb] = 280F For Quad Cities Unit 2, the dropweight testing to establish NDT was not recorded for the weld materials, therefore, GE procedure requires that, when no NDT is available for the weld, the resulting RTNDT should be -500 F or higher. The value of (TfoT - 600F) in this example was -320 F; therefore the initial RTNDT was -32 0F.

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

For vessel forging material, such as nozzles and closure flanges, the method for establishing RTNDT is the same as for vessel plate material. For the feedwater nozzle at Quad Cities Unit 2 (Heat ZT2885), the NDT is 300F and the lowest CVN data is 32 ft-lb at 400 F. The corresponding value of (T50T- 600F) is:

(T5OT - 601F) = {[40 + (50 - 32) ft-lb 20F/ft-lb] + 300 F) - 60WF = 460F.

Therefore, the initial RTNOT is the greater of nil-ductility transition temperature (NDT) or (To-r 60 0F), which is 460F.

In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNDT. For the lower torus heat of Quad Cities Unit 2 (Heat A1899-1), the NDT is 400 F and the lowest CVN data was 32 ft-lb at 400F. The corresponding value of (T50T - 600F) was:

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version (TSOT - 60'F) {[40 + (50 - 32) ft-lb

• 20F/ft-lb] + 300F} - 600 F = 46 0F.

Therefore, the initial RTNDT was 460F.

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section 11I, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 600 F is the LST for the bolting materials. Charpy data for the Quad Cities Unit 2 closure studs did not all meet the 45 ft-lb, 25 MLE requirements at 100 F. Therefore, the LST for the bolting material is 700F. The highest RTNDT in the closure flange region is 23.1 0F, for the vertical electroslag weld material.

Thus, the higher of the LST and the RTNDT +600 F is 83.10F, the boltup limit in the closure flange region.

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

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version TOP HEAD TOP HEAD FLANGE SHELL FLANGE

- SHELL#4

, SHELL*3 TOP OF SHELL*2 ACTVE FUEL (TAF) 360.3' BOTTOM OF SHELL#1 ACTIVE FUEL (BAF) 21&3 BOTMOM HEAD SUPPORT SKIRT Notes: (1) Refer to Tables 4-1 and 4-2 for reactor vessel components and their heat identifications.

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

Figure 4-1: Schematic of the Quad Cities Unit 2 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version Table 4-1: RTNDT Values for Quad Cities Unit 2 Vessel Materials COMPONENT HEAT TEST CHARPY ENERGY. (TsoT-60) DROP RTNDT TEMP. (FT-LB) (OF) WEIGHT (F)

_ __ _ _ __ _ _ _ _ _ _ _ (OF) __ _ __ _ _ _ N DT _ _ _

PLATES & FORGINGS:

TOP HEAD and FLANGE Dollar Plate B5845-1 40 45 47 48 20 40 40 Mk201 Torus B5853-1 10 75 82 77 -20 10 10 MK202 C2748-1 10 45 35 37 10 10 10 C2748-3 10 45 39 60 2 10 10 A0313-1 10 34 58 57 12 10 12 Flange Mk209 3P1131 10 92 168 115 -20 10 10 Mk48 3P1 118 10 96 60 104 -20 10 10 SHELL COURSES Upper Shell A0985-1 10 55 67 69 -20 10 10 Mk6O A0942-1 10 64 58 50 -20 10 10 A0998-1 10 43 60 49 -6 IC 10 Upper-Int Shell C1717-2 10 50 46 62 -12 10 10 Mk59 C1717-1 10 49 55 65 -18 10 10 C1510-2 10 34 36 44 12 10 12 Lower-Int Shell C2753-2 10 68 64 50 -20 10 10 Mk58 C2868-1 10 72 51 44 -8 10 10 C3307-2 10 60 87 77 -20 IC 10 Lower Shell C1516-2 10 40 39 37 6 -20 6 Mk57 C1501-2 10 66 54 60 -20 -10 -10 C1722-2 10 41 47 35 10 -10 10 BOTTOM HEAD Dollar plate C2393-1 40 50 53 34 42 40 42 Mkl Torus, lower Al 899-1 40 61 32 60 46 40 46 Mk4 Al 907-1 40 47 64 91 16 40 40 Torus, upper 86747-2 40 67 37 90 36 40 40 Mk2 C2702-1 40 63 75 61 10 40 40 A1888-2 40 61 52 53 10 40 40 C2588-1B 40 95 94 84 10 40 40 SUPPORT SKIRT Mk40 C3885-4 40 94 66 81 10 40 40 Note: These are minimum Charpy values.

GE Nuclear Energy GE-NE-OOOD-0002-9600-03R2a Non-Proprietary Version Table 4-2: RTNDT Values for Quad Cities Unit 2 Nozzle and Weld Materials COMPONENT HEAT TEST CHARPY ENERGY (T5o-6O) DROP RTNDT TEMP. (FT-LB) ('F) WEIGHT (/F)

(OF) i NDT NOZZLES Recirc Outlet Nozzle ZT2885 40 36 39 39 38 30 38 Mk8 ZT2869-1 40 45.5 49 54 19 30 30 Recirc inlet Nozzle ZT2872 40 45.5 37.5 36 38 30 38 Mk7 E25VW 40 68 89 66 10 30 30 Steam Outlet Nozzle ZT3043-1 10 49 49 94 -18 30 30 Mk14 Feedwater (Trans.) ZT2405-5 40 52.5 61.5 70.5 10 40 40 Mk10 ZT2885 40 32 34 36.5 46 30 46 Core Spray Nozzle E26VW 40 83 78 66 10 40 40 Mk11 Drain Nozzle 212918 40 238 239 237 10 NIA 10 CRD Penetration C2393-1 40 50 53 34 42 40 42 Closure Head, Vent ZT3043 40 102 130 117 10 40 40 CRD HSR &

Core Diff. Press & Uq. Con Noz Mk2O6 &204& 13& 17 Jet Pump Instr BT2615 40 132 118 120 10 40 40 Mk19 WELDS:

Longitudinal ES 23.1 Beitline Seams Low. Int. to Lower S3986/3870 10 41 45 46 -32 -32 Girth Seam STUDS: LST Studs 6720372 10 53 35 58 70 MK61 Note: These are minimum Charpy values.

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version 4.2 A DJUS TED REFERENCE TEMPERA TURE FOR BEL TLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG1.99) [7] provides the methods for determining the ART. The RG1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beltline plates and several beltline welds was made and summarized in Tables 4-3 and 4-4 for 32 and 54 EFPY, respectively.

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

SHIFT = ARTNDT + Margin where, ARTNDT = [CF]*f (28 - 0.10 log r Margin = 2(Ci 2 + aA2)05 CF = chemistry factor from Tables I or 2 of Rev. 2 f =.1/4T fluence / 1019 Margin = 2(ai 2 + CA 2 05

)

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

ad = standard deviation on ARTNDT, 280 F for welds and 17'F for base material, except that cyaneed not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT The margin term aE has constant values in RG1.99 of 170F for plate and 280 F for weld.

However, E, need not be greater than 0.5

• ARTNOT. Since the GE/BWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb level, the value of a, is taken to be 0F for the vessel plate and most weld materials, except that a, is 130 F for the beltline electroslag weld materials [13d].

4.2.1.1 Chemistry The vessel beltline chemistries were obtained from several sources as detailed below:

• Vessel Plates: Copper - plate manufacturer [13a]; Nickel - highest value from CMTRs [12]

• Submerged Arc Welds: Copper and nickel from separate evaluation [13b & 13c]

• Electroslag Welds: Copper and nickel from separate evaluations [13d & 13c]

The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG1.99 [7], to determine a chemistry factor (CF) per Paragraph 1.1 of RG1.99 for welds and plates, respectively. Best estimate results are used for the beltline electroslag weld materials for the initial RTNDT [13d]; therefore, the standard deviation (a,) is specified.

4.2.1.2 Fluence A bounding pre-EPU (Extended Power Uprate) and EPU flux [14] for the vessel ID wall are calculated using methods consistent with Regulatory Guide 1.190. The flux in Reference 14 is determined for the pre-EPU power of 2527 MW, and for the EPU rated power of 2957 MWt.

The bounding peak fast flux for the RPV inner surface from Reference 14 is 3.12e8 n/cm2 -sfor pre-EPU and 3.46e8 n/cm2 -s for EPU conditions.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 32 EFPY Fluence Quad Cities Unit 2 began EPU operation at 21 EFPY, thereby operating for 11 EFPY at EPU conditions for 32 EFPY. The RPV ID surface fluence for 32 EFPY is calculated as follows:

3.12e8 n/cm2-s*1.01e9 s*(21/32) + 3.46e8 n/cm 2_s*1.01e9 s*(11/32)= 3.3e17 n/cm2 .

This fluence applies to the lower-intermediate plates and axial weld materials. The fluence is adjusted for the lower shell and axial welds, as well as for the lower to lower-intermediate girth weld based upon a peak / lower shell location ratio of 0.71 for pre-EPU conditions and 0.74 for EPU conditions (at an elevation of approximately 258" above vessel "0"); hence the peak ID surface fluence used for these components is 2.4e17 n/cm 2.

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

using the Quad Cities Unit 2 plant specific fluence and vessel thickness of 6.125". The 32 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

3.3e17 n/cm2

• exp (-0.24 * (6.125/4)) = 2.3e17 n/cm 2.

The 32 EFPY 1/4T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

2.4e17 n/cm2

• exp (-0.24 * (6.125/4)) = 1.6e17 n/cm 2.

54 EFPY Fluence As stated above, Quad Cities Unit 2 began EPU operation at 21 EFPY, thereby operating for 33 EFPY at EPU conditions for 54 EFPY. The RPV ID surface fluence for 54 EFPY is calculated as follows:

3.12e8 ni/cm 2-s*1.7e9 s*(21/54) + 3.46e8 n/cm 2-s*1.7e9 s*(33/54)= 5.7e17 n/cm2 .

This fluence applies to the lower-intermediate plates and axial weld materials. The fluence is adjusted for the lower shell and axial welds, as well as for the lower to lower-intermediate girth weld based upon a peak / lower shell location ratio of 0.71 for pre-EPU conditions and 0.74 for EPU conditions (at an elevation of approximately 258" above vessel "0"); hence the peak ID surface fluence used for these components is 4.1e17 n/cm2.

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version The fluence at 1/4T is calculated per Equation 3 of Regulatory Guide 1.99, Revision 2 [7]

using the Quad Cities Unit 2 plant specific fluence and vessel thickness of 6.125". The 54 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

5.7e17 n/cm2

• exp (-0.24 * (6.125/4)) = 3.9e17 n/cm 2.

The 54 EFPY 1/4T fluence for the lower shell plate and axial welds and the lower to lower-intermediate girth weld is:

4.1e17 n/cm 2

• exp (-0.24 * (6.125/4)) = 2.9e17 n/cm 2.

4.2.2 Limiting Beitline Material The limiting beitline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, RG1.99 [7] was applied to compute ART.

Tables 4-3 and 4-4 list values of beltline ART for 32 and 54 EFPY, respectively.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Table 4-3: Quad Cities Unit 2 Beltline ART Values (32 EFPY)

Lawr-Ilternwdlate lates and Axal Welds Thickrru - 6.13 inches 32 EFPY Peak I.D. fluence - 3.3E+17 t/an^2 32 EFPY Peak 1/4 Tfluence - 2.3E+17 n/cnr'2 32 EFPY Peak 1U4T fluence - 2.3E+17 ncrn'2 L wer Plates and AxIal Walds and Lower to Lower-lterndlate CGlrth Weld Thickness - 6.13 inches 32 EFPY Peak I.D. fluence - 2.4E+17 nlcnt2 32 EFPY Peak 1/4 Tfluence - 1.6E+17 utcnP2 32 EFPY Peak 1/4 T fluence - 1.6E+17 n/taIt2 Initial 1U4T 32 EFPY l Ca 32 EFPY 32 EFPY COMPONENT HEAT OR HEAT/LOT %CO %Ni CF RTndt Fluence A RTndt Mrin Shift ART OF ntnf2 OF IF OF OF PLATES:

6-1224 C1516-2 0.16 0.46 108 6 1.6E+17 16 0 E 16 33 39 6-122-10 C1501-2 0.1S 0.49 124 -10 1.6E+17 19 0 9 19 37 27 6-122.14 C1722-2 0.14 0.54 97 10 1.6E+17 15 0 7 15 29 39 Lwer-Ilnteradlate 6-139-16 C2753-2 0.08 0.50 51 10 2.3E+17 9 0 5 9 19 29 6-139-22 C2868-1 0.08 0.48 51 10 2.3E+17 9 0 5 9 19 29 6-139-25 C3307-2 0.12 0.55 E2 10 2.3E+17 15 0 8 15 30 40 WELDS:

IAwer-Intermnedblte ES 0.24 0.37 141 23 2.3E+17 26 13 13 37 63 E6 Loer ES- 0.24 0.37 141 23 1.6E+17 21 13 11 34 55 78 Gbib Lower to oer4antemnediate SAW 5398613870 Linde 124 0.05 0.96 68 -32 1.6E+17 10 0 5 10 21 .11 Ciefsatly values bed on daisafrom BAW-2259, dated Januimyl996, butadjusted. Valuesof Initial RTndt id at aeohtained frontthe etnedocumienL bs GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Table 4-4: Quad Cities Unit 2 Beltline ART Values (54 EFPY)

Lawer-Intermedlate Plates and Axial Welds Thickness - 6.13 lndhes 54 EFPY Peak ID. fluence - 5.7E+17 1/tde2 54 EFPY Peak 114Tfluence - 3.9E+17 nrtnT2 54 EFPY Peak 1/4 Tfluence - 3.9E+17 nlam`2 LAover Plates and Axial Welds and Lawer to Lwer-latertnedlate Clrth Weld Thickneh - 6.13 inches 54 EFPY Peak ID. fluence - 4.1E+17 nucner2 54 ERPY Peak 1/4 T fluence - 2.9E+17 n/cnr'2 54 EFPYPeak1/4Tfluence- 2.9E+17 ti/an'2 Initial 114T 54 EFPY °l _ 54 EFPY 54 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF RTndt Fluence A RTndt Margin Shift ART

• F n/cmn2 eF F F IF PLATES:

ZACr.

6-122-S C1516-2 0.16 0.46 108 6 2.9E+17 23 0 12 23 46 52 6-122-10 C1501-2 0.18 0.49 124 -10 2.9E+17 26 0 13 26 53 43 6-122-14 C1722-2 0.14 0.54 97 10 2.9E+17 21 0 10 21 41 51 l.wer-lntermedlate 6-139-16 C2753-2. 0.08 0.50 51 10 3.9E+17 13 0 .7 13 26 36 6-139-22 C2868-1 0.08 0.48 51 10 3.9E+17 13 0 7 13 26 36 6-139-25 C3307-2 0.12 0.55 82 10 3.9E+17 21 0 11 21 42 52 WELDS:

LAwer-lntertnedlate ES- 0.24 0.37 141 23 3.9E+17 36 13 18 45 81 104 Lawer ES- 0.24 0.37 141 23 2.9E+17 30 13 15 40 70 93 Irth Lower to Lower-Intermediate SAW 3398613870 Linde 124 0.05 0.96 68 .32 2.9E+17 15 0 7 is 29 -3 Cbhenstry values ae bmsedon datatofnmBAW-2259 dated Jammy 1996. but usted. Values of Imttal RTnd idal ootatned from the sane document GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 to which a pressure-retaining component may be subjected over its service lifetime. The ASME Code (Appendix G of Section Xl 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 portions of the vessel (i.e., upper vessel, lower vessel) include shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will also be called the non-beltline region.

For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 200 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 1I4T 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 IOCFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-5:

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Table 4-5: Summary of the 10CFR50 Appendix G Requirements Condition and Pressure C^Operating Minimum Temperature Requirement

.;, A,.

X I. Hydrostatic Pressure Test & Leak Test (Core is Not Critical) - Curve A

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

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 900 F II. Normal operation (heatup and cooldown),

including anticipated operational occurrences

a. Core not critical - Curve B

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

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

b. Core critical - Curve C

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

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

• 600 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 10CFR5O Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [15]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

((:

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version

))

4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beftline Regions Non-beitline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0E17 n/cm2) to cause any significant shift of RTNDT. Non-beltline components include nozzles (see Appendix E),

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

Detailed stress analyses of the non-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The BWR/6 stress analysis bounds for BWRI2 through BWR/5 designs, as will be demonstrated in the following evaluation. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include 1000 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 16] 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 4-6 and 4-7.

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version Table 4-6: Applicable BWR/3 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identii ton .,.'

FW Nozzle CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head (B only)

Top Head Nozzles (B only)

Recirculation Outlet Nozzle (B only)

Table 4-7: Applicable BWR/3 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity.Identification.

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 Quad Cities Unit 2 as the plant specific geometric values are bounded by the generic GE Nuclear Energy GE-NE-000O-0002-9600-03R2a 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 Quad Cities Unit 2 by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.

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

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

In a (( )) finite element analysis (( )), the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K4.

The (( )) evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section Xl Appendix G [6] and shown below. The results of that computation were K,= 143.6 ksi-in1 2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 840F. ((

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 8.0 inches; hence, ti 2 = 2.83. The resulting value obtained was:

Mm 1.85 for _ftS2 Mm= 0.926 4f for 2<Iit<3.464 = 2.6206 Mm = 3.21 for Vt>3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Klb is calculated from the equation in Paragraph G-2214.2 [6]:

Km = Mm = (( )) ksi-in"2 Kib = (2/3) Mm *Cypb = (( )) ksi-in"2 The total Ki is therefore:

K = 1.5 (Kim+ Kib) + Mm * (am + (2/3)

• cab) = 143.6 ksi-inl/2 GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific Ka is based on the K, the equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) = In [(K5-33.2) / 20.734] / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 840 F The generic curve was generated by scaling 143.6 ksi-in'2 by the nominal pressures and calculating the associated (T - RTNDT):

The highest RTNDT for the bottom head plates and welds is 460 F, as shown in Tables 4-1 and 4-2. ((

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version Second, the P-T curve is dependent on the calculated K1value, and the K1value is proportional to the stress and the crack depth as shown below:

K4 oc a (na)'2 (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is W/4. Thus, K is proportional to R/(t)" 2. The generic curve value of R/(t)' 2, based on the generic BWR/6 bottom head dimensions, is:

Generic: 11 2 = 138 / (8)"2= 49 inch"2 R/ (t) (4-2)

The Quad Cities Unit 2 specific bottom head dimensions are R = 125.7 inches and t =8 inches minimum (19], resulting in:

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Quad Cities Unit 2 specific: R / (t)' 2 = 125.7 / (8)"/ = 44 inch1 2 (4-3)

Since the generic value of R/(t)' 2 is larger, the generic P-T curve is conservative when applied to the Quad Cities Unit 2 bottom head.

4.3.2.1.2 Core Not Critical Heatup/Cooldown - Non-Beltline 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. ((

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1]

The calculated value of K1for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical. Therefore, the K1value for the core not critical condition is (143.6/1.5) 2.0 = 191.5 ksi-in11 .

Therefore, the method to solve for (T - RTNDT) for a specific K1 is based on the K,:

equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:

(T - RTNDT) = In [( 1 -33.2/ 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 102'F The generic curve was generated by scaling 192 ksi-in"' by the nominal pressures and calculating the associated (T- RTNDT):

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

-: psig) - - --(ksi-inin-:(F 1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49 -14 The highest RTNDT for the bottom head plates and welds is 460F, as shown in Tables 4-1 and 4-2. ((

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1]

As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Table 4-7 and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

((

GE Nuclear Energy GE-NE-0000-0002-960O-03R2a Non-Proprietary Version 4.3.2.1.3 Pressure Test - Non-Beltine Curve A (Using Feedwater Nozzle/Upper Vessel Region)

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

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

To evaluate the results, K, is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or Xl). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of K, is shown below using the BWR/6,251-inch dimensions:

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

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

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

a= 4(t 2 + t 2)112 =2.36inches tn = thickness of nozzle = 7.125 inches tv = thickness of vessel = 6.1875 inches r, = apparent radius of nozzle = r1 + 0.29 rc=7.09 inches r, = actual inner radius of nozzle = 6.0 inches r, = nozzle radius (nozzle corner radius) = 3.75 inches Thus, a/r, = 2.36 / 7.09 = 0.33. The value F(a/rQ), taken from Figure A5-1 of WRC Bulletin 175 for an a/re of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (7ra)2

• F(a/rQ):

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Nominal Kg = 1.5 34.97 - (r -2.36)"2 -1.4 = 200 ksi-in" 2 The method to solve for (T - RTNDT) for a specific K1 is based on the K4, equation of Paragraph A-4200 in ASME Appendix A [17] for the pressure test condition:

(T - RTNDT) = In [(K,-33.2) / 20.734] / 0.02 (T - RTNDT) = In [(200 - 33.2) /20.734] / 0.02 (T - RTNDT) = 104.20F The generic pressure test P-T curve was generated by scaling 200 ksi-in'2 by the nominal pressures and calculating the associated (T - RTNDT), ((

• 1))

GE Nuclear Energy GE-NE-00.00-0002-9600-03R2a Non-Proprietary Version The highest RTNDT for the feedwater nozzle materials is 460 F as shown in Table 4-2.

The generic pressure test P-T curve is applied to the Quad Cities Unit 2 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 46 0F.

((

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version Second, the P-T curve is dependent on the K1 value calculated. The Quad Cities Unit 2 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K4 are shown below:

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

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

The Dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding C = 35.04 ksi. The factor F (alr,) from Figure A5-1 of WRC-175 is determined where:

ea = Y,( tn 2+ tv 2)1/2 =2.35 inches tn = thickness of nozzle = 7.15 inches tv = thickness of vessel = 6.125 inches rj = apparent radius of nozzle = r1+ 0.29 rc=6.9 inches r1 = actual inner radius of nozzle = 6.0 inches rc= nozzle radius (nozzle comer radius) = 3.0 inches Thus, a/re = 2.35/6.9 = 0.34. The value F(a/r,), taken from Figure AS-1 of WRC Bulletin 175 for an a/rn of 0.34, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K, is 1.5 a (1ra)'

• F(alr,):

Nominal K1= 1.5 - 35.04* (-

• 2.35)"2- 1.4 = 200 ksi-in" 2 4.3.2.1.4 Core Not Critical Heatup/Cooldown - Non-Beltline Curve B (Using Feedwater Nozzle/Upper Vessel Region)

The feedwater nozzle was selected to represent non-beitline 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 GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.

Stresses were taken from a (( )) finite element analysis done specifically for the purpose of fracture toughness analysis (( )). Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 40'F feedwater injection, which is equivalent to hot standby, as seen in Figure 4-3.

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

Bulletin 175 [15].

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

Kip = SF -a (na)%

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

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version 1] .

Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(aIrQ) for Equation 4-4. These values are shown in Figure A5-1 of WRC Bulletin 175 [15].

The stresses used in Equation 4-4 were taken from (( )) design stress reports for the feedwater nozzle. The stresses considered are primary membrane, Cypm, and primary bending, Capb. Secondary membrane, a,,, and secondary bending, asb, stresses are included in the total K, by using ASME Appendix G [6] methods for secondary portion, KS = Mm (asm + (2/3)

• asb) (4-5)

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 K, are added to obtain the total value of stress intensity factor, K,. 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 Kg is based on the Kic equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

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

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

feedwater nozzle (( J] analysis, where feedwater injection of 400F into the vessel while at operating conditions (551.4 0F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle comer stresses were obtained from finite element analysis (( )). To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 inches was used in the evaluation.

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

vessel thickness of 6.1875 inches (i.e., apm = 20.49 ksi was revised to 20.49 ksi 7.5 inches/6.1875 inches = 24.84 ksi). These stresses, and other inputs used in the generic calculations, are shown below:

apm = 24.84 ksi Osm = 16.19 ksi CsYS = 45.0 ksi t, = 6.1875 inches apb = 0.22 ksi ab = 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, a., so the correction factor, R,is calculated to consider the nonlinear effects in the plastic region according to GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 = [as - apm + ((atotal - ays) I 30)] 1 (ato(al - apm) (4-7)

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

apm = 24.84 ksi 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, tin = 3.072. The resulting value obtained was:

Mm = 1.85 for 1irt2 Mm = 0.926 t for 2<tS3.464 = 2.845 Mm = 3.21 for ft/7>3.464 The value F(a/rQ), taken from Figure AS-1 of WRC Bulletin 175 for an a/r, of 0.33, is therefore, F(a/r ) =1.4 Kp is calculated from Equation 4-4:

Kp = 2.0 - (24.84 + 0.13) * (r - 2 . 3 6)tf

• 1.4 Kp = 190.4 ksi-in"i K4 is calculated from Equation 4-5:

KS = 2.845 - (9.44 + 2/3

• 11.10)

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version KS= 47.9 ksi-in12 The total K, is, therefore, 238.3 ksi-in" 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 K,;

this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 400 F water injected into the hot reactor vessel nozzle. In the base case that yielded a K, value of 238 ksi-in'1 , the pressure is 1050 psig and the hot reactor vessel temperature is 551.4 0F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by ffsat. - 40) / (551.4 - 40). From K the associated (T - RTNDT) can be calculated:

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

.Nominal Pressure Saturation Temp.- R Keg*- -- (T - RTNDT):

(psig) -(F)- _.: _: _- (ksi-in,) (OF):'.

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

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

The highest non-beltline RTNDT for the feedwater nozzle at Quad Cities Unit 2 is 46 0F as shown in Table 4-2. The generic curve is applied to the Quad Cities Unit 2 upper vessel GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 46*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.

The stress intensity factors (K1 ), calculated for the beltline region according to ASME Code Appendix G procedures [6], were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate; values were calculated for 100°F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits.

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Cm = PR /tn (4-8)

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

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with K1c, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

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

for the pressure test condition:

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

This relationship provides values of pressure versus temperature (from KR 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 200F/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 1000F/hr. The K, calculation for a coolant heatup/cooldown rate of 1000F/hr is described in Section 4.3.2.2.3 below.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 4.3.2.2.2 Calculations for the Beitline Region - Pressure Test 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 = 23 + 63 = 86 0F (Based on ART values in Table 4-3)

Vessel Height H = 823 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to inside of clad) R = 125.7 inches Minimum Vessel Thickness (without clad) t = 6.125 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 + (823-216.3) 0.0361 = 1127 psig Pressure stress:

a = PR/t (4-11)

= 1.127 *125.7/6.125 = 23.1 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 (6]

was based on a thickness of 6.125 inches (the minimum thickness without cladding);

hence, tl 2 = 2.47. The resulting value obtained was:

Mm = 1.85 for -A<2 Mm = 0.926 4t for 2<r_ <3.464 = 2.29 Mm = 3.21 for ,/ >3.464 The stress intensity factor for the pressure stress is Klm = Mm - a. 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 200F/hr instead of 100IF/hr.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Equation 4-9 can be rearranged, and 1.5 Kim substituted for Kc, to solve for (T - RTNDT).

Using the K, equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 52.96, and Kg,= 2.29 for a 200 F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

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

= ln[(1.5

• 52.96 + 2.29 - 33.2) / 20.734] / 0.02

= 42.50F T can be calculated by adding the adjusted RTNDT:

T = 42.5 + 86 = 128.50F for P = 1105 psig at 32 EFPY 4.3.2.2.3 Beltline Region - Core Not Critical Heatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section Xl Appendix G [6]:

Kc= 2.0

• Km +KI (4-13) where Kim is primary membrane K due to pressure and K, 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, GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G [6]. The relationship used to compute the through-wall AT, is based on one-dimensional heat conduction through an insulated flat plate:

a 2T(x,t) /ax2 = 1Ip (T(x,t) 1t)

I (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) I 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/ P+To (4-15)

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 4.3.2.2.4 Calculations for the Beltline 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 Klm as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. In addition, there is a K, term for the thermal stress.

The additional inputs used to calculate K1t are:

Coolant heatup/cooldown rate, normally 100OF/hr G = 100 OF/hr Minimum vessel thickness, including clad thickness C = 0.526 ft (6.3125 inches)

Thermal diffusivity at 5500 F (most conservative value) p = 0.354 ft2/ hr [21]

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

AT = GC 2 /2,P (4-16)

= 100 - (0.526)2/ (2 *0.354) = 39 0F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.2914) can be interpolated from. ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kt = Ml - AT = 11.39, can be calculated. The conservative value for thermal diffusivity at 5500 F is used for all calculations; therefore, K,, is constant for all pressures. Km has the same value as that calculated in Section 4.3.2.2.2.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version The pressure and thermal stress terms are substituted into Equation 4-9 to solve for (T - RTNDT):

(T - RTNDT) = ln[((2

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

= In[(2 - 52.96 + 11.39 - 33.2) / 20.734] /0.02

= 700F T can be calculated by adding the adjusted RTNDT:

T = 70 + 86 = 156 'F for P = 1105 psig at 32 EFPY 4.3.2.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT. Similar to the evaluations performed for the bottom head and upper vessel, a BWR/6 finite element analysis [18] was used to model the flange region. The local stresses were computed for determination of the stress intensity factor, K. Using a 1/4T flaw size and the K,0 formulation to determine T - RTNDT, for pressures above 312 psig the P-T limits for all flange regions are bounded by the 10 CFR50 Appendix G requirement of RTNDT + 900F (the largest T-RTNDT for the flange at 1563 psig is 730 F).

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version The approach used for Quad Cities Unit 2 for the bolt-up temperature was based on the conservative value of (RTNDT+ 60), or the LST of the bolting materials, whichever is greater. The 600 F adder is included by GE for two reasons: 1) the pre-1971 requirements of the ASME Code Section III, Subsection NA, Appendix G included the 600 F adder, and 2) inclusion of the additional 600 F requirement above the RTNDT provides the additional assurance that a 114T flaw size is acceptable. As shown in Tables 4-1 and 4-2, the limiting initial RTNDT for the closure flange region is represented by the electroslag weld materials in the upper shell at 23.10 F, and the LST of the closure studs is 700F; therefore, the bolt-up temperature value used is 830F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

10CFR50 Appendix G, paragraph IV.A.2 [8] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 900F) and Curve B temperature no less than (RTNDT + 120 0F).

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

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

The shutdown margin, provided in the Quad Cites Unit 2 Technical Specification, is calculated for a water temperature of 680F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 680F limit, further extensive calculations would be required to justify a lower temperature. The 830 F limit for the upper vessel and beltline region and the 680F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel. When the head is not GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

4.3.2.4 CORE CRITICALOPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFR50 Appendix G [8], Table 1. Table 1 of [8] requires that core critical P-T limits be 400 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 400F 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 (RTNOT + 600F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 830F, based on an RTNDT of 23.10 F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 1600 F or the temperature required for the hydrostatic pressure test (Curve A at 1105 psig). The requirement of closure region RTNDT + 1600F causes a temperature shift in Curve C at 312 psig.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 200 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, Kr at 1/4T to be less than that at 3/4T for a given metal temperature.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version The following P-T curves were generated for Quad Cities Unit 2.

• Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 and 54 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 32 and 54 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 32 and 54 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 1380 psig for Curve A and are non-beltline limited for Curve B at 32 EFPY. At 54 EFPY, the P-T curves are beltline limited above 760 psig for Curve A and above 770 psig for Curve B.

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

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves Cu

. C X Numbers for.... Numbers

..... eDescripti,, ..

-~ - resentatolof Presentation o

_;_, .___ __--_____. the -T Curves the -T Curv es Curve A _

Bottom Head Limits (CRD Nozzle) Figure 5-1 Table B-1 & 3 Upper Vessel Limits (FW Nozzle) Figure 5-2 Table B-1 & 3 Beltline Limits for 32 EFPY Figure 5-3 Table B-1 Beltline Limits for 54 EFPY Figure 5-4 Table B-3 Curve B Bottom Head Limits (CRD Nozzle) Figure 5-5 Table B-1 & 3 Upper Vessel Limits (FW Nozzle) Figure 5-6 Table B-1 & 3 Beltline Limits for 32 EFPY Figure 5-7 Table B-1 Beltline Limits for 54 EFPY Figure 5-8 Table B-3 Curve C Composite Curve for 32 EFPY** Figure 5-9 Table B-2 Composite Curve for 54 EFPY** Figure 5-10 Table B-4 A &B Composite Curves for 32 EFPY Bottom Head and Composite Curve A Figure 5-11 Table B-2 for 32 EFPY*

Bottom Head and Composite Curve B Figure 5-12 Table B-2 for 32 EFPY*

A &B Composite Curves for 54 EFPY Bottom Head and Composite Curve A Figure 5-13 Table B-4 for 54 EFPY*

Bottom Head and Composite Curve B Figure 5-14 Table B-4 for 54 EFPY*_

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100 cn In 1000 0

XL 900 INITIALRTndtVALUEIS I 49°F FOR BOTTOM HEADl J

co 800 o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 20'F/R W 600 z

500 un 400 co, LU 300 200 100 0 I I I -' I I I' I  ! 4 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

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

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

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100 icn la

- 1000 I

0- 900 0 lINITIAL RTndt VALUE IS I

-j l146°F FOR UPPER VESSEL_ I w

m 800 o 700 HEATUP/COOLDOWN RATE OF COOLANT 2 600 :s200 F/HR

7-

_ 500 Lu aco 400 w

300 200

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

(°F)

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

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

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 I. I/

INITIAL RTndt VALUE IS 23.1PF FOR BELTLINE 1100 0t a 1000 O.

R 900 I_ I /

BELTLINE CURVE LU U) 800 ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 32 63 o 700 HEATUP/COOLDOWN at 600 RATE OF COOLANT 3 < 20*F/HR

> 500 Lu Ir 3 400 _ 4 _ _4 c

LU Ir 300 200 100 0 - 6- a- S 4 - 6- a- S - 4-0 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 32 EFPY

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

GE Nuclear Energy GE-N E-OOOO-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 on a 1 000 I

0- 900 tJ BELTLINE CURVE Cn 800 ADJUSTED AS SHOWN:

EFPY SHIFT ( 0F) 54 81 I-o 700 HEATUPICOOLDOWN RATE OF COOLANT

< 200F/HR

500 w

X2 400 0:

300 200 0:

100 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 54 EFPY

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100 W

cm en D 1000

a. 900 INITIAL RTndt VALUE IS I 0

I- 160.6*F FOR BOTTOM HEAD I

-. 1j V) 800 o 700 HEATUP/COOLDOWN I- RATE OF COOLANT

. 100F/HR W 600 Z

> 500 W.'

Ca 400 co 300 200 100 0

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100 Z.

- 1000 D- 900 INITIAL RTndt VALUE IS 0 I 46*F FOR UPPER VESSEL ILU to 800 o 700 HEATUP/COOLDOWN

- 0 RATE OF COOLANT 0

[F1 JF/HR w

0: 500

=U 400 200 1

w

3 00 300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Umits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTORVESSELMETALTEMPERATURE rF)

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-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUE IS 23.1iF FOR BELTUNE 1100 1000 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT ( 0F) 1-0Z 900 32 63 0a w

800 m

tz t0 700 3 HEATUP/COOLDOWN RATE OF COOLANT 600 < 1000F/HR n

0.

500 400 300 I312 PS I 7 200 100 ___-_ X _

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

Figure 5-7: Beltline P-T Curve for Core Not Critical [Curve B] up to 32 EFPY

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUE IS 23.10F FOR BELTUNE 1100 la D* 1000 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (¶F) 0 900 54 81 0

I-

-l n 800 o 700 HEATUPICOOLDOWN RATE OF COOLANT 0: 600 c 100 0F/HR Z

500 Lu Uj 400 Lu CZ 300 200 100 0

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

Figure 5-8: Beltline P-T Curve for Core Not Critical [Curve B] up to 54 EFPY

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

i

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 23.1°F FOR BELTUNE, 46SF FOR UPPER VESSEL, 1200 AND 49@F FOR BOTTOM HEAD 1100 atm BELTUNE CURVE c 1000 ADJUSTED AS SHOWN:

EFPY SHIFT (OF)

n. 900 32 63 o

(n 800 Con HEATUP/COOLDOWN RATE OF COOLANT

< 100F1HR o 700

= 60 w 600 2

,40 Lu 1 500 nCU400 Lu 300 200

-BELTUNE AND NOI-BELTLINE 100 LIMITS 0

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

Figure 5-9: Core Critical P-T Curves [Curve C] up to 32 EFPY

[IOO0 F/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 ----

INITIAL RTndt VALUES ARE 1300 23.1PF FOR BELTUNE, 46°F FOR UPPER VESSEL, 1200 AND 49°F FOR BOTTOM HEAD 1100-CL 1000 _BELTLINE CURVE 1000ADJUSTED AS SHOWN:

9 -EFPY SHIFT (°F) 900o581 o _

I-I 800___HEATUP/COOLDOWN RATE OF COOLANT

< 1000 F/HR o 700-c7C-___ ______ ________

600 w

00 43600 200- b___ -mum-t Temperature 3F-BELTLINE AND 100 _ I I -

_ _.NON-BELTLINE LIMITS 0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-10: Core Critical P-T Curves [Curve C] up to 54 EFPY

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 23.1OF FOR BELTLINE, 460 F FOR UPPER VESSEL, AND 1100 490 F FOR BOTTOM HEAD Is 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (OF) 0 32 63

2. 700 n

o 700 HEATUPICOOLDOWN LU RATE OF COOLANT

< 20¶FIHR 20

-UPPER VESSEL AND BELTLINE LIMITS

---

BOTTOM HEAD 100 CURVE 0

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

Figure 5-11: Composite Pressure Test P-T Curves [Curve A] up to 32 EFPY

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

• GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Nbn-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1OF FOR BELTLINE, 460 F FOR UPPER VESSEL, 1100 AND 60.6*F FOR BOTTOM HEAD D 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

0 900 EFPY SHIFT (0F) 0 I- 32 63 0'

C,'

800

'U o 700 HEATUP/COOLDOWN al- RATE OF COOLANT W- 600 . 100FIHR, 2

a)40

=i 500

'U In 400

'U 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 (OF)

Figure 5-12: Composite Core Not Critical P-T Curves [Curve B] up to 32 EFPY

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

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 23.10F FOR BELTLINE, 46 0F FOR UPPER VESSEL, AND 1100 49OF FOR BOTTOM HEAD in 0.

C 1000 8ELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (0F) aL I-Z5 900 54 81 us 800 u) o 700 I-I HEATUPICOOLDOWN RATE OF COOLANT to 600 ' 200F/HR z

_j 500 CIV 400 tL 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 (tF)

Figure 5-13: Composite Pressure Test P-T Curves [Curve A] up to 54 EFPY

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

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1°F FOR BELTLINE, 46 0 F FOR UPPER VESSEL, 1100 AND 60.6°F FOR BOTTOM HEAD 0.

in C 1000 BELTLINE CURVES III ADJUSTED AS SHOWN:

0. 900 EFPY SHIFT (°F) 0 54 81 I-

-II LU U) 800 U) o 700 C.1 HEATUP/COOLDOWN RATE OF COOLANT 2 600 < 100OFIHR z

I--

i 500 ILl

° 400 LU 300

-UPPER VESSEL 200 AND BELTLINE LIMITS 100 HEAD -

CURVE 0

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

Figure 514: Composite Core Not Critical P-T Curves [Curve B] up to 54 EFPY

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

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

6.0 REFERENCES

1. B.J. Branlund, "Pressure-Temperature Curves for ComEd Quad Cities 2", GE-NE, San Jose, CA, May 2000 (GE-NE-B13-02057-00-01R2, Revision 2) (GE Proprietary).

2. GE Drawing Number 921D265, "Reactor Thermal Cycles - Reactor Vessel", GE-APED, San Jose, CA, Revision 1. Dresden and Quad Cities RPV Thermal Cycle Diagram (GE Proprietary).

3. GE Drawing Number 158B7279, "Nozzle Thermal Cycles - Reactor Vessel", GE-APED, San Jose, CA, Revision 1. Dresden and Quad Cities Nozzle Thermal Cycle Diagram (GE Proprietary).

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

5. "Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels Section Xl, Division 1", Code Case N-588 of the ASME Boiler &

Pressure Vessel Code, Approval Date December 12, 1997. (Note this reference is not used in this report because the girth welds are not limiting.)

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.

- 67 -

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

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

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

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

12. QA Records & RPV CMTR's:

Quad Cities 2 - (QA Records & RPV CMTR's Quad Cities Unit 2 GE PO#205-55599 (B&W) and 205-H4502 (CB&I), Mfg by B&W, RDM and CB&I)",

General Electric Company Atomic Power Equipment Department (APED) Quality Control - Procured Equipment, RPV QC", Mt. Vernon, Indiana, and subvendor Rotterdam Drydock, Rotterdam, Holland; Chicago Bridge and Iron, Memphis, Tennessee.

13. a). Letter, J.F. Longnecker (Lukens Steel) to T.A. Caine (GE), "Copper Content of Reactor Vessel Plates", dated August 27, 1985.

b). L.B. Gross, "Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds", BAW-2121P, April 1991.

c) Letter from R.M. Krich to the NRC, "Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity - Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 - LaSalle County Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. NPF-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.

d) "Evaluation of RTNDT, USE and Chemical Composition of Core Region Electroslag Welds for Quad Cities Units 1 and 2", BAW-2259, January 1996.

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version

14. a) S. Sitaraman, "Dresden and Quad Cities Neutron Flux Evaluation", GE-NE, San Jose, CA, March 2003, (GE-NE-0000-0011-0531-R2, Revision 2)(GE Proprietary Information).

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

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

Welding Research Council Bulletin 175, August 1972.

16. ((

))

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

18. ((

JI

19. Bottom Head and Feedwater Nozzle Dimensions: "Certified Design Document for Quad Cities I & II, B&W Contract No. 610-0122-51/52, GE Order No. 205-55599, Vol. 1 of 6", Nuclear Power Generation Division, Babcock & Wilcox Co., Mt. Vernon, Indiana, November 1970 (GE VPF 1744-211-1).

20. ((

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

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

22. Letter, M. Banerjee, USNRC to J.L. Skolds, Exelon Generation Company, LLC, "Dresden Nuclear Power Station, Units 2 and 3 - Issuance of Amendments Regarding Pressure and Temperature Limits (TAC Nos. MB7850 and MB7851)",

November 26, 2003 (GE Proprietary).

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version A-2

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 600F. Also Inconel discontinuities require no fracture toughness evaluations. The RPV penetrations of the nozzles listed in Table A-1 and bound the RPV penetration for the nozzles listed below, therefore, no further fracture toughness evaluation is performed for these nozzles. Nozzles and appurtenances < 2.5' or made from Inconel are not included in Table A-1 and are listed below. The Top Head Lifting Lugs are also not included in Table A-1 because the loads only occur on these components when the reactor is shutdown during an outage.

Components not requiring a fracture toughness evaluation are listed below:

Nozzle orRTD LS Appurtenance Nozzle or Appurtenance Material Reference RTNDO LST Identification MK 12 2" Instrumentation < 2.5" Alloy 600 N/A N/A 1, 2 & 5 Penetration in RPV Shell MK 22 Drain- Penetration < 2.5"- Bottom SA1 05-GR 2 42 102 Head 1,2&5 MK 51 - 54 Shroud Support Attachment to RPV Alloy 600 N/A N/A Wall 1, 2 & 5 Penetration in RPV Shell l MK 77 & 81-84 Insulation Brackets - Carbon Steel N/A N/A 1, 2 & 5 RPV Shells MK 101-128 Control Rod Drive Stub Tubes - Alloy 600 N/A N/A Bottom Head 1,2 & 5 Penetration in Dollar Plate Mk 139 High and Low Pressure Seal Leak SA1 05-GR 2 Detection- 5 Penetration - 1"* - Flange Not a pressure boundary component; therefore, no fracture toughness evaluation required.

MK 210 Top Head Lifting Lugs (only loads at outage) 1, 2 & 5 Attachment to Torus Not a pressure boundary component; therefore, no fracture toughness evaluation required.

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

" N/A - Not applicable for this material type.

A-3

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX A

REFERENCES:

1. RPV Outline or As-Built

• Chicago Bridge & Iron Co. Drawing #69-4824 R4 & R5, "Vessel As-Built Dimensions", Chicago Bridge & Iron Co., (GE-NE VPF#2752-129-1 for R4 &

2752-130-1 for R5)

2. Certified Stress Report

• Certified Design Document for Quad Cities I & II, vol. 1 of 6" B&W contract No.

610-0122-51/52, GE Order No. 205-55599", Babcock & Wilcox Co, Mt. Vernon, Indiana, November, 1970, (GE-NE VPF# 1744-211-1) - Quad Cities Unit 1&2

3. GE Drawing # 104R921, Revision 7, "Reactor Assembly, Nuclear Boiler", GE-NED, San Jose, CA.) - Quad Cities Unit 1 & 2

4. Dresden/Quad Cities LR PT Curves - Design Input Request (DIR),

Robert Stachniak, 4/26/02.

5. QA Records & RPV CMTR's

• Quad Cities 2 - (QA Records & RPV CMTR's Quad Cities Unit 2 GE PO# 205-55599(B&VW) and 205-H4502 (CB&I), Mfg by B&W, RDM and CB&I)"General Electric Company Atomic Power Equipment Department (APED) Quality Control

- Procured Equipment, RPV QC" Project: Quad Cities II, Purchase Order: 205-55599-Il, Vendor: Babcock & Wilcox, Location: Mt. Vernon, Indiana, and sub-vendor Rotterdam Drydock, Rotterdam, Holland; Chicago Bridge and Iron, Memphis, Tenn.

A-4

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-1. Quad Cities Unit 2 P-T Curve Values for32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6 and 5-7 BOTTOM UPPER  : 32 EFPY BOTTOM UPPER -32 EFPY

-HEAD - VESSEL 'BELTLINE HEAD VESSEL BELTLINE

. PRESSURE

. . CURVE A CURVE A -CURVE A CURVE B CURVE B CURVE B

-. (PSIG) (0F) - -(°F) (oF): (0F) - (f) ( 0F) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 83.1 83.1 140 68.0 83.1 83.1 68.0 83.4 83.1 150 68.0 83.1 83.1 68.0 86.2 83.1 160 68.0 83.1 83.1 68.0 88.9 83.1 170 68.0 83.1 83.1 68.0 91.5 83.1 180 68.0 83.1 83.1 68.0 93.9 83.1 190 68.0 83.1 83.1 68.0 96.2 83.1 200 68.0 83.1 83.1 68.0 98.3 83.1 210 68.0 83.1 83.1 68.0 100.3 83.1 220 68.0 83.1 83.1 68.0 102.3 83.1 230 68.0 83.1 83.1 68.0 104.1 83.1 B-2

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-1. Quad Cities Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 °F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6 and 5-7 BOTTOM UPPER 32 EFPY- BOTTOM - UPPER 32 EFPY

-HEAD VESSEL -B- ELTLINE HEAD- VESSEL BELTLINE PRESSURE CURVEA CURVE A -CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (0F) - -(OF) (OF) - ((OF)

F) (0F) -

240 68.0 83.1 83.1 68.0 105.9 83.1 250 68.0 83.1 83.1 68.0 107.6 83.1 260 68.0 83.1 83.1 68.0 109.2 83.1 270 68.0 83.1 83.1 68.0 110.8 83.1 280 68.0 83.1 83.1 68.0 112.3 83.1 290 68.0 83.1 83.1 68.0 113.8 83.1 300 68.0 83.1 83.1 68.0 115.2 83.1 310 68.0 83.1 83.1 68.0 116.5 83.1 312.5 68.0 83.1 83.1 68.0 116.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380

• 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.0 143.1 143.1 430 68.0 113.1 113.1 68.0 143.1 143.1 440 68.0 113.1 113.1 68.0 143.1 143.1 450 68.0 113.1 113.1 68.0 143.1 143.1 460 68.0 113.1 113.1 68.0 143.1 143.1 470 68.0 113.1 113.1 68.6 143.1 143.1 480 68.0 113.1 113.1 71.1 143.1 143.1 B-3

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-1. Quad Cities Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 *F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6 and 5-7 BOTTOM UPPER - - 32EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG)- (*F) (0 F) (F) I(0 F).: (0 F) -( 0 F)-

113.1 143.1 143.1 490 68.0 113.1 73.4 500 68.0 113.1 113.1 75.6 143.1 143.1 510 68.0 113.1 113.1 77.8 143.1 143.1 520 68.0 113.1 113.1 79.8 143.1 143.1 530 68.0 113.1 113.1 81.8 143.1 143.1 540 68.0 113.1 113.1 83.7 143.1 143.1 550 68.0 113.1 113.1 85.5 143.1 143.1 560 68.0 113.1 113.1 87.3 143.1 143.1 570 68.0 113.1 113.1 89.0 143.1 143.1 580 68.0 113.1 113.1 90.6 143.1 143.1 590 68.0 113.1 113.1 92.2 143.6 143.1 600 68.0 113.1 113.1 93.8 144.1 143.1 610 68.0 113.1 113.1 95.3 144.6 143.1 620 68.0 113.1 113.1 96.7 145.0 143.1 630 68.0 113.1 113.1 98.1 145.4 143.1 640 68.0 113.1 113.1 99.5 145.8 143.1 650 68.0 113.1 113.1 100.8 146.2 143.1 660 68.0 113.1 113.1 102.1 146.7 143.1 670 68.0 113.1 113.1 103.4 147.1 143.1 680 68.0 113.1 113.1 104.7 147.5 143.1 690 68.0 113.1 113.1 105.9 147.9 143.1 700 69.2 113.1 113.1 107.0 148.3 143.1 710 70.7 113.1 113.1 108.2 148.7 143.1 720 72.1 113.1 113.1 109.3 149.1 143.1 730 73.5 113.1 113.1 110.4 149.5 143.1 740 74.8 113.1 113.1 111.5 149.9 143.1 750 76.1 113.1 113.1 112.6 150.2 143.1 B-4

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-1. Quad Cities Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6 and 5-7 BOTTOM UPPER -:32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE - CURVEA CURVE A - CURVE A - CURVE B CURVE B CURVE B (PSIG) (0F) (F) - (F) - (0F) (OF) (0F) 760 77.4 113.1 113.1 113.6 150.6 143.1 770 78.6 113.1 113.1 114.6 151.0 143.1 780 79.8 113.1 113.1 115.6 151.4 143.1 790 81.0 113.1 113.1 116.6 151.8 143.1 800 82.2 113.9 113.1 117.5 152.1 143.1 810 83.3 114.6 113.1 118.5 152.5 143.1 820 84.4 115.4 113.1 119.4 152.9 143.1 830 85.5 116.1 113.1 120.3 153.2 143.1 840 86.5 116.8 113.1 121.2 153.6 143.1 850 87.6 117.5 113.1 122.0 153.9 143.1 860 88.6 118.2 113.1 122.9 154.3 143.1 870 89.6 118.9 113.1 123.7 154.6 143.1 880 90.5 119.6 113.1 124.6 155.0 143.1 890 91.5 120.3 113.1 125.4 155.3 143.1 900 92.4 120.9 113.1 126.2 155.7 143.1 910 93.4 121.6 113.1 127.0 156.0 143.8 920 94.3 122.2 113.1 127.7 156.4 144.5 930 95.1 122.9 113.9 128.5 156.7 145.2 940 96.0 123.5 114.9 129.3 157.0 145.9 950 96.9 124.1 115.8 130.0 157.4 146.6 960 97.7 124.7 116.8 130.7 157.7 147.2 970 98.6 125.3 117.7 131.5 158.0 147.9 980 99.4 125.9 118.6 1322 158.4 148.5 990 100.2 126.5 119.4 132.9 158.7 1492 1000 101.0 127.1 120.3 133.6 159.0 149.8 1010 101.7 127.7 121.1 134.2 159.3 150.5 1020 102.5 128.2 122.0 134.9 159.6 151.1 B-5

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-1. Quad Cities Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 IF/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6 and 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE ' HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A' CURVE B CURVE B CURVE B (PSIG) (0F) ( 0F) (OF) (OF) - ( 0F) (OF) 1030 103.3 128.8 122.8 135.6 160.0 151.7 1040 104.0 129.4 123.6 136.2 160.3 152.3 1050 104.7 129.9 124.4 136.9 160.6 152.9 1060 105.4 130.5 125.2 137.5 160.9 153.5 1070 106.2 131.0 126.0 138.1 161.2 154.1 1080 106.9 131.5 126.7 138.8 161.5 154.7 1090 107.6 132.1 127.5 139.4 161.8 155.2 1100 108.2 132.6 128.2 140.0 162.1 155.8 1105 108.6 132.8 128.6 140.3 162.3 156.1 1110 108.9 133.1 128.9 140.6 162.4 156.3 1120 109.6 133.6 129.6 141.2 162.7 156.9 1130 110.2 134.1 130.4 141.8 163.0 157.4 1140 110.9 134.6 131.1 142.3 163.3 158.0 1150 111.5 135.1 131.7 142.9 163.6 158.5

.1160 112.1 135.6 132.4 143.5 163.9 159.0 1170 112.8 136.1 133.1 144.0 164.2 159.6 1180 113.4 136.6 133.7 144.6 164.5 160.1 1190 114.0 137.1 134.4 145.1 164.7 160.6 1200 114.6 137.5 135.0 145.7 165.0 161.1 1210 115.2 138.0 135.7 146.2 165.3 161.6 1220 115.8 138.5 136.3 146.8 165.6 162.1 1230 116.3 138.9 136.9 147.3 165.9 162.6 1240 116.9 139.4 137.5 147.8 166.2 163.1 1250 117.5 139.8 138.1 148.3 166.4 163.6 1260 118.0 140.3 138.7 148.8 166.7 164.1 1270 118.6 140.7 139.3 149.3 167.0 164.5 1280 119.1 141.2 139.9 149.8 167.2 165.0 B-6

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-1. Quad Cities Unit 2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A For Figures 5-1, 5-2, 5-3, 5-5, 5-6 and 5-7 BOTTOM UPPER :32EFPY - BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE .,CURVE A CURVE A CURVE A CURVEB CURVE B CURVEB (PSIG) (0F) (OF) (OF) (0F) (OF) (OF) 1290 119.7 141.6 140.5 150.3 167.5 165.5 1300 120.2 142.0 141.0 150.8 167.8 165.9 1310 120.7 142.5 141.6 151.3 168.1 166.4 1320 121.3 142.9 142.2 151.8 168.3 166.8 1330 121.8 143.3 142.7 152.2 168.6 167.3 1340 122.3 143.7 143.3 152.7 168.8 167.7 1350 122.8 144.1 143.8 153.2 169.1 1682 1360 123.3 144.6 144.3 153.6 169.4 168.6 1370 123.8 145.0 144.9 154.1 169.6 169.0 1380 124.3 145.4 145.4 154.5 169.9 169.5 1390 124.8 145.8 145.9 155.0 170.1 169.9 1400 125.3 146.2 146.4 155.4 170.4 170.3 B-7

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-2. Quad Cities Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 *F/hrfor Curve A FOR FIGURES 5-9, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY- - 32 EFPY 32 EFPY PRESSURE CURVE A CURVEA CURVE B CURVE B - CURVE C

--(PSIG) - (OF (F ('F. (OF):

0 68.0 *83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 86.0 70 68.0 83.1 68.0 83.1 93.2 80 68.0 83.1 68.0 83.1 99.2 90 68.0 83.1 68.0 83.1 104.3 100 68.0 83.1 68.0 83.1 108.8 110 68.0 83.1 68.0 83.1 112.9 120 68.0 83.1 68.0 83.1 116.7 130 68.0 83.1 68.0 83.1 120.2 140 68.0 83.1 68.0 83.4 123.4 150 68.0 83.1 68.0 86.2 126.2 160 68.0 83.1 68.0 88.9 128.9 170 68.0 83.1 68.0 91.5 131.5 180 68.0 83.1 68.0 93.9 133.9 190 68.0 83.1 68.0 96.2 136.2 200 68.0 83.1 68.0 98.3 138.3 210 68.0 83.1 68.0 100.3 140.3 B-8

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-2. Quad Cities Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 OF/hr for Curve A FOR FIGURES 5-9, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

- HEAD

- BELTLINEAT HEAD BELTLINE AT BELTLINEAT-32 EFPY - - 32 EFPY -- 32 EFPY

.PRESSURE- CURVE A CURVEA .-.CURVE B CURVE B CURVE C (PSIG)

(Pl(

0F) -

4t. (OF O

- (OF) - ((F 0 )(F 220 68.0 83.1 68.0 102.3 142.3 230 68.0 83.1 68.0 104.1 144.1 240 68.0 83.1 68.0 105.9 145.9 250 68.0 83.1 68.0 107.6 147.6 260 68.0 83.1 68.0 109.2 149.2 270 68.0 83.1 68.0 110.8 150.8 280 68.0 83.1 '68.0 112.3 152.3 290 68.0 83.1 68.0 113.8 153.8 300 68.0 83.1 68.0 115.2 155.2 310 68.0 83.1 68.0 116.5 156.5 312.5 68.0 83.1 68.0 116.9 156.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.0 143.1 183.1 430 68.0 113.1 68.0 143.1 183.1 440 68.0 113.1 68.0 143.1 183.1 B-9

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-2. Quad Cities Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 'F/hr for Curves B & C and 20 OF/hr for Curve A FOR FIGURES 5-9,5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT -:HEAD BELTLINE AT BELTINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE. CURVE A CURVE A CURVE B CURVE B - CURVE C (PSIG) - (F-) (F (a (OF) (OF) 450 68.0 113.1 68.0 143.1 183.1 460 68.0 113.1 68.0 143.1 183.1 470 68.0 .113.1 68.6 143.1 183.1 480 68.0 113.1 71.1 143.1 183.1 490 68.0 113.1 73.4 143.1 183.1 500 68.0 113.1 75.6 143.1 183.1 510 68.0 113.1 77.8 143.1 183.1 520 68.0 113.1 79.8 143.1 183.1 530 68.0 113.1 81.8 143.1 183.1 540 68.0 113.1 83.7 143.1 183.1 550 68.0 113.1 85.5 143.1 183.1 560 68.0 113.1 87.3 143.1 183.1 570 68.0 113.1 89.0 143.1 183.1 580 68.0 113.1 90.6 143.1 183.1 590 68.0 113.1 92.2 143.6 183.6 600 68.0 113.1 93.8 144.1 184.1 610 68.0 113.1 95.3 144.6 184.6 620 68.0 113.1 96.7 145.0 185.0 630 68.0 113.1 98.1 145.4 185.4 640 68.0 113.1 99.5 145.8 185.8 650 68.0 113.1 100.8 146.2 186.2 660 68.0 113.1 102.1 146.7 186.7 670 68.0 113.1 103.4 147.1 187.1 680 68.0 113.1 104.7 147.5 187.5 690 68.0 113.1 105.9 147.9 187.9 B-10

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-2. Quad Cities Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A FOR FIGURES 5-9, 5-11 and 5-12

-:BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY -;

PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (0F  ; (0F- (0 F)- (OF), -

700 69.2 113.1 107.0 148.3 188.3 710 70.7 113.1 108.2 148.7 188.7 720 72.1 113.1 109.3 149.1 189.1 730 73.5 113.1 110.4 149.5 189.5 740 74.8 113.1 111.5 149.9 189.9 750 76.1 113.1 112.6 150.2 190.2 760 77.4 113.1 113.6 150.6 190.6 770 78.6 113.1 114.6 151.0 191.0 780 79.8 113.1 115.6 151.4 191.4 790 81.0 113.1 116.6 151.8 191.8 800 82.2 113.9 117.5 152.1 192.1 810 83.3 114.6 118.5 152.5 192.5 820 84.4 115.4 119.4 152.9 192.9 830 85.5 116.1 120.3 153.2 193.2 840 86.5 116.8 121.2 153.6 193.6 850 87.6 117.5 122.0 153.9 - 193.9 860 88.6 118.2 122.9 154.3 194.3 870 89.6 118.9 123.7 154.6 194.6 880 90.5 119.6 124.6 155.0 195.0 890 91.5 120.3 125.4 155.3 195.3 900 92.4 120.9 126.2 155.7 195.7 910 93.4 121.6 127.0 156.0 196.0 920 94.3 122.2 127.7 156.4 196.4 930 95.1 122.9 128.5 156.7 196.7 940 96.0 123.5 129.3 157.0 197.0 B-11

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-2. Quad Cities Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 OF/hr for Curve A FOR FIGURES 5-9, 5-11 and 5-12 BOTTOM UPPER RPV.& BOTTOM UPPER RPV & UPPER RPV &

..HEAD BELTLINE AT-. :HEAD BELTLINEAT -'BELTLINEAT 32 EFPY 32 EFPY - 32 EFPY .

PRESSURE CURVEA CURVEA CURVEB CURVEB - CURVE C (PSIG) (0 F) (0 F) - - (°

( (FF (OF 950 96.9 124.1 130.0 157.4 197.4 960 97.7 124.7 130.7 157.7 197.7 970 98.6 .125.3 131.5 158.0 198.0 980 99.4 125.9 132.2 158.4 198.4 990 100.2 126.5 132.9 158.7 198.7 1000 101.0 127.1 133.6 159.0 199.0 1010 101.7 127.7 134.2 159.3 199.3 1020 102.5 128.2 134.9 159.6 199.6 1030 103.3 128.8 135.6 160.0 200.0 1040 104.0 129.4 136.2 160.3 200.3 1050 104.7 129.9 136.9 160.6 200.6 1060 105.4 130.5 137.5 160.9 200.9 1070 106.2 131.0 138.1 161.2 201.2 1080 106.9 131.5 138.8 161.5 201.5 1090 107.6 132.1 139.4 161.8 201.8 1100 108.2 132.6 140.0 162.1 202.1 1105 108.6 132.8 140.3 162.3 202.3 1110 108.9 133.1 140.6 162.4 202.4 1120 109.6 133.6 1412 162.7 202.7 1130 110.2 134.1 141.8 163.0 203.0 1140 110.9 134.6 142.3 163.3 203.3 1150 111.5 135.1 142.9 163.6 203.6 1160 112.1 135.6 143.5 163.9 203.9 1170 112.8 136.1 144.0 164.2 204.2 1180 113.4 136.6 144.6 164.5 204.5 B-12

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-2. Quad Cities Unit 2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & Cand 20 *F/hr for Curve A FOR FIGURES 5-9,5-11 and 5-12

-BOTTOM UPPER RPV& BOTTOM UPPER RPV& -UPPER RPV &

-- HEAD BELTLINEAT HEAD BELTLINEAT BELTLINE AT 32E 32 EFP 32 EFPY

, PRESSURE--CURVEA -CURVEAA .CURVE B CURVE B - CURVE C-(PSIG) -(F) (0F)-(°F (O - (0F-1190 114.0 137.1 145.1 164.7 204.7 1200 114.6 137.5 145.7 165.0 205.0 1210 115.2 138.0 146.2 165.3 205.3 1220 115.8 138.5 146.8 165.6 205.6 1230 116.3 138.9 147.3 165.9 205.9 1240 116.9 139.4 147.8 166.2 206.2 1250 117.5 139.8 148.3 166.4 206.4 1260 118.0 140.3 148.8 166.7 206.7 1270 118.6 140.7 149.3 167.0 207.0 1280 119.1 141.2 149.8 167.2 207.2 1290 119.7 141.6 150.3 167.5 207.5 1300 120.2 142.0 150.8 167.8 207.8 1310 120.7 142.5 151.3 168.1 208.1 1320 121.3 142.9 151.8 168.3 208.3 1330 121.8 143.3 152.2 168.6 208.6 1340 122.3 143.7 152.7 168.8 208.8 1350 122.8 144.1 153.2 169.1 209.1 1360 123.3 144.6 153.6 169.4 209.4 1370 123.8 145.0 154.1 169.6 209.6 1380 124.3 145.4 154.5 169.9 209.9 1390 124.8 145.9 155.0 170.1 210.1 1400 125.3 146.4 155.4 170.4 210.4 B-13

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-3. Quad Cities Unit 2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6 and 5-8 BOTTOM UPPER 54 EFPY BOTTOM- UPPER 54 EFPY HEAD VESSEL - BELTLINE -HEAD VESSEL BELTLINE

.,PRESSURE CURVE A- -CURVE A CURVE A- CURVE B --CURVE B -CURVE B 0F) 1;(PSIG --F - (OF) (0F- (OF F) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 83.1 83.1 140 68.0 83.1 83.1 68.0 83.4 83.1 150 68.0 83.1 83.1 68.0 86.2 83.1 160 68.0 83.1 83.1 68.0 88.9 83.1 170 68.0 83.1 83.1 68.0 91.5 83.1 180 68.0 83.1 83.1 68.0 93.9 83.1 190 68.0 83.1 83.1 68.0 96.2 83.1 200 68.0 83.1 83.1 68.0 98.3 83.1 210 68.0 83.1 83.1 68.0 100.3 83.1 220 68.0 83.1 83.1 68.0 102.3 83.1 230 68.0 83.1 83.1 68.0 104.1 83.1 B-14

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-3. Quad Cities Unit 2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6 and 5-8 BOTTOM UPPER - 54 EFPY - BOTTOM UPPER 54 EFPY

-HEAD VESSEL-; BELTLINE HEAD :VESSEL BELTLINE 1PRESSURE - - CURVE A CURVE A - CURVE A CURVE B - - CURVE B -CURVE B

- (F (F( 0F) (OF): (

[`- (PSIG) 240 68.0 83.1 83.1 68.0 105.9 83.1 250 68.0 83.1 83.1 68.0 107.6 83.1 260 68.0 83.1 83.1 68.0 109.2 83.1 270 68.0 83.1 83.1 68.0 110.8 83.1 280 68.0 83.1 83.1 68.0 112.3 83.1 290 68.0 83.1 83.1 68.0 113.8 83.1 300 68.0 83.1 . 83.1 68.0 115.2 83.1 310 68.0 83.1 83.1 68.0 116.5 83.1 312.5 68.0 83.1 83.1 68.0 116.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.0 143.1 143.1 430 68.0 113.1 113.1 68.0 143.1 143.1 440 68.0 113.1 113.1 68.0 143.1 143.1 450 68.0 113.1 113.1 68.0 143.1 143.1 460 68.0 113.1 113.1 68.0 143.1 143.1 470 68.0 113.1 113.1 68.6 143.1 143.1 480 68.0 113.1 113.1 71.1 143.1 143.1 B-15

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-3. Quad Cities Unit 2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 IF/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6 and 5-8

- OTOM B UPPER 54 EFPY BOTTOM UPPER 54 EF

- HEAD --VESSEL BELTLINE HEAD VESSEL -- BELT INE i PRESSURE CURVE A CURVE A - -CURVE A: CURVE B. -: CURVE B : CURVE B (PSIG) - (°F (F (F) (OF) (OF (OF) 490 68.0 113.1 113.1 73.4 143.1 143.1 500 68.0 113.1 113.1 75.6 143.1 143.1 510 68.0 113.1 113.1 77.8 143.1 143.1 520 68.0 113.1 113.1 79.8 143.1 143.1 530 68.0 113.1 113.1 81.8 143.1 143.1 540 68.0 113.1 113.1 83.7 143.1 143.1 550 68.0 113.1 113.1 85.5 143.1 143.1 560 68.0 113.1 113.1 87.3 143.1 143.1 570 68.0 113.1 113.1 89.0 143.1 143.1 580 68.0 113.1 113.1 90.6 143.1 143.1 590 68.0 113.1 113.1 92.2 143.6 143.1 600 68.0 113.1 113.1 93.8 144.1 143.1 610 68.0 113.1 113.1 95.3 144.6 143.1 620 68.0 113.1 113.1 96.7 145.0 143.1 630 68.0 113.1 113.1 98.1 145.4 143.1 640 68.0 113.1 113.1 99.5 145.8 143.1 650 68.0 113.1 113.1 100.8 146.2 143.1 660 68.0 113.1 113.1 102.1 146.7 143.1 670 68.0 113.1 113.1 103.4 147.1 143.1 680 68.0 113.1 113.1 104.7 147.5 143.1 690 68.0 113.1 113.1 105.9 147.9 143.1 700 69.2 113.1 113.1 107.0 148.3 144.0 710 70.7 113.1 113.1 108.2 148.7 145.0 720 72.1 113.1 113.1 109.3 149.1 146.0 730 73.5 113.1 113.1 110.4 149.5 146.9 740 74.8 113.1 113.1 111.5 149.9 147.9 750 76.1 113.1 113.1 112.6 150.2 148.8 B-16

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-3. Quad Cities Unit 2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6 and 5-8 I ',BOTTOM: UPPER 54 EFPY BOTTOM UPPER 54.EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE 'CURVE A ' CURVEA ' ' CURVE A: CURVE B CURVE B CURVE B -

(PSIG) (°F) (IF- (a ' i (F , (ff, (OF 760 77.4 113.1 113.1 113.6 150.6 149.7 770 78.6 113.1 113.2 114.6 151.0 150.6 780 79.8 113.1 114.6 115.6 151.4 151.5 790 81.0 113.1 116.0 116.6 151.8 152.4 800 82.2 113.9 117.3 117.5 152.1 153.2 810 83.3 114.6 118.6 118.5 152.5 154.1 820 84.4 115.4 119.9 119.4 152.9 154.9 830 85.5 116.1 121.1 120.3 153.2 155.7 840 86.5 116.8 122.3 121.2 153.6 156.5 850 87.6 117.5 123.4 122.0 153.9 157.3 860 88.6 118.2 124.6 122.9 154.3 158.1 870 89.6 118.9 125.7 123.7 154.6 158.8 880 90.5 119.6 126.8 124.6 155.0 159.6 890 91.5 120.3 127.9 125.4 155.3 160.3 900 92.4 - 120.9 128.9 126.2 155.7 161.1 910 93.4 121.6 129.9 127.0 156.0 161.8 920 94.3 122.2 130.9 127.7 156.4 162.5 930 95.1 122.9 131.9 128.5 156.7 163.2 940 96.0 123.5 132.9 129.3 157.0 163.9 950 96.9 124.1 133.8 130.0 157.4 164.6 960 97.7 124.7 134.8 130.7 157.7 165.2 970 98.6 125.3 135.7 131.5 158.0 165.9 980 99.4 125.9 136.6 132.2 158.4 166.5 990 100.2 126.5 137.4 132.9 158.7 167.2 1000 101.0 127.1 138.3 133.6 159.0 167.8 1010 101.7 127.7 139.1 134.2 159.3 168.5 1020 102.5 128.2 140.0 134.9 159.6 169.1 B-I17

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-3. Quad Cities Unit 2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6 and 5-8 O U ' E BOTTOM UP-PERR' 54 EFPY HEAD VESSEL' BELTLINE HAD VESSEL ,BELTLINE PRESSURE CURVE A: CURVE A: CURVE A CURVE B CURVE B ,- CURVE B (PSIG) - (OF) (on ' (°-" (f') (F ' (OF) 1030 103.3 128.8 140.8 135.6 160.0 169.7 1040 104.0 129.4 141.6 136.2 160.3 170.3 1050 104.7 129.9 142.4 136.9 160.6 170.9 1060 105.4 130.5 143.2 137.5 160.9 171.5 1070 106.2 131.0 144.0 138.1 161.2 172.1 1080 106.9 131.5 144.7 138.8 161.5 172.7 1090 107.6 132.1 145.5 139.4 161.8 173.2 1100 108.2 132.6 146.2 140.0 162.1 173.8 1105 108.6 132.8 146.6 140.3 162.3 174.1 1110 108.9 133.1 146.9 140.6 162.4 174.3 1120 109.6 133.6 147.6 141.2 162.7 174.9 1130 110.2 134.1 148.4 141.8 163.0 175.4 1140 110.9 134.6 149.1 142.3 163.3 176.0 1150 111.5 135.1 149.7 142.9 163.6 176.5 1160 112.1 135.6 150.4 143.5 163.9 177.0 1170 112.8 136.1 151.1 144.0 164.2 177.6 1180 113.4 136.6 151.7 144.6 164.5 178.1 1190 114.0 137.1 152.4 145.1 164.7 178.6 1200 114.6 137.5 153.0 145.7 165.0 179.1 1210 115.2 138.0 153.7 146.2 165.3 179.6 1220 115.8 138.5 154.3 146.8 165.6 180.1 1230 116.3 138.9 154.9 147.3 165.9 180.6 1240 116.9 139.4 155.5 147.8 166.2 181.1 1250 117.5 139.8 156.1 148.3 166.4 181.6 1260 118.0' 140.3 156.7 148.8 166.7 182.1 1270 118.6 140.7 157.3 149.3 167.0 182.5 1280 119.1 141.2 157.9 149.8 167.2 183.0 B-18

GE Nuclear Energy GE-NE-0000-0002-96OD-03R2a Non-Proprietary Version TABLE B-3. Quad Cities Unit 2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6 and 5-8 BOTTOM UPPER-- 54 EFPY BOTTOM - UPPER 54 EFPY

-HEAD VESSEL BELTLINE HEAD VESSEL - BELTLINE i PRESSURE CURVE A CURVE A -- CURVE A - CURVE B CURVE B- CURVE B (PSIG) -(nF (n (F) (OF) (OF) 1290 119.7 141.6 158.5 150.3 167.5 183.5 1300 120.2 142.0 159.0 150.8 167.8 183.9 1310 120.7 142.5 159.6 151.3 168.1 184.4 1320 121.3 142.9 160.2 151.8 168.3 184.8 1330 121.8 143.3 160.7 152.2 168.6 185.3 1340 122.3 143.7 161.3 152.7 168.8 185.7 1350 122.8 144.1 161.8 153.2 169.1 186.2.

1360 123.3 144.6 162.3 153.6 169.4 186.6 1370 123.8 145.0 162.9 154.1 169.6 187.0 1380 124.3 145.4 163.4 154.5 169.9 187.5 1390 124.8 145.8 163.9 155.0 170.1 187.9 1400 125.3 146.2 164.4 155.4 170.4 188.3 B-19

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B4. Quad Cities Unit 2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14

'BOTTOM ' UPPER RPV &^ BOTTOM' UPPER RPV & UPPER RPV&

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT

54 EFPY 54 EFPY - 54 EFPY -

i:PRESSURE CURVEA CURVE A .- - CURVE B. CURVE B CURVE C' Y-(PSIG) '-- (OF)' ° '-tF' (OF) ' - ' tF ' "

0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 86.0 70 68.0 83.1 68.0 83.1 932 80 68.0 83.1 68.0 83.1 992 90 68.0 83.1 68.0 83.1 104.3 100 68.0 83.1 68.0 83.1 108.8 110 68.0 83.1 68.0 83.1 112.9 120 68.0 83.1 68.0 83.1 116.7 130 68.0 83.1 68.0 83.1 120.2 140 68.0 83.1 68.0 83.4 123.4 150 68.0 83.1 68.0 86.2 126.2 160 68.0 83.1 68.0 88.9 128.9 170 68.0 83.1 68.0 91.5 131.5 180 68.0 83.1 68.0 93.9 133.9 190 68.0 83.1 68.0 96.2 136.2 200 68.0 83.1 68.0 98.3 138.3 210 68.0 83.1 68.0 100.3 140.3 B-20

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-4. Quad Cities Unit 2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV& BOTTOM UPPER RPV &.UPPER RPV&

HEAD BELTLINE AT; - -HEAD BELTLINE AT BELTLINE AT e -54 Y 54 EPY 54 EFPY

,:PRESSURE CURVE A - CURVE A CURVE B CURVE B CURVE C:-

(PSIG) - (0F (F) (a ) (°F (-F) 220 68.0 83.1 68.0 102.3 142.3 230 68.0 83.1 68.0 104.1 144.1 240 68.0 83.1 68.0 105.9 145.9 250 68.0 83.1 68.0 107.6 147.6 260 68.0 83.1 68.0 109.2 149.2 270 68.0 83.1 68.0 110.8 150.8 280 68.0 83.1 68.0 112.3 152.3 290 68.0 83.1 68.0 113.8 153.8 300 68.0 83.1 68.0 115.2 155.2 310 68.0 83.1 68.0 116.5 156.5 312.5 68.0 83.1 68.0 116.9 156.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 .68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.0 143.1 183.1 430 68.0 113.1 68.0 143.1 183.1 440 68.0 113.1 68.0 143.1 183.1 B-21

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B4. Quad Cities Unit 2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & -`UPPER RPV &

-HEAD BELTLINE AT -HEAD ; BELTLINE AT BELTLINE AT 54 EFPY 54EFP 54 EFPY

°.PRESSURE CURVE A -CURVEA - CURVE B CURVE B CURVE C (PSIG)j ((FO- (FF - --- .F) 450 68.0 113.1 68.0 143.1 183.1 460 68.0 113.1 68.0 143.1 183.1 470 68.0 .113.1 68.6 143.1 183.1 480 68.0 113.1 71.1 143.1 183.1 490 68.0 113.1 73.4 143.1 183.1 500 68.0 113.1 75.6 143.1 183.1 510 68.0 113.1 77.8 143.1 183.1 520 68.0 113.1 79.8 143.1 183.1 530 68.0 113.1 81.8 143.1 183.1 540 68.0 113.1 83.7 143.1 183.1 550 68.0 113.1 85.5 143.1 183.1 560 68.0 113.1 87.3 143.1 183.1 570 68.0 113.1 89.0 143.1 183.1 580 68.0 113.1 90.6 143.1 183.1 590 68.0 113.1 92.2 143.6 183.6 600 68.0 113.1 93.8 144.1 184.1 610 68.0 113.1 95.3 144.6 184.6 620 68.0 113.1 96.7 145.0 185.0 630 68.0 113.1 98.1 145.4 185.4 640 68.0 113.1 99.5 145.8 185.8 650 68.0 113.1 100.8 146.2 186.2 660 68.0 113.1 102.1 146.7 186.7 670 68.0 113.1 103.4 147.1 187.1 680 68.0 113.1 104.7 147.5 187.5 690 68.0 113.1 105.9 147.9 187.9 B-22

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B4. Quad Cities Unit 2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV&8 BOTTOM UPPER RPV & UPPER RPV &

-'-.HEAD BELTLINE AT HEAD BELTLINE AT: BELTLINE AT.:

54 EFPY 54 EFPY 54 EFPY I;PRESSURE'-CURVE A CURVE A :CURVE B CURVE B CURVE C (PSIG) - (° (0F) - (F) (°F (OF) 700 692 113.1 107.0 148.3 188.3 710 70.7 113.1 108.2 148.7 188.7 720 72.1 113.1 109.3 149.1 189.1 730 73.5 113.1 110.4 149.5 189.5 740 74.8 113.1 111.5 149.9 189.9 750 76.1 113.1 112.6 150.2 190.2 760 77.4 113.1 113.6 150.6 190.6 770 78.6 113.2 114.6 151.0. 191.0 780 79.8 114.6 115.6 151.5 191.5 790 81.0 116.0 116.6 152.4 192.4 800 82.2 117.3 117.5 153.2 193.2 810 83.3 118.6 118.5 154.1 194.1 820 84.4 119.9 119.4 154.9 194.9 830 85.5 121.1 120.3 155.7 195.7 840 86.5 122.3 121.2 156.5 196.5 850 87.6 123.4 122.0 157.3 197.3 860 88.6 124.6 122.9 158.1 198.1 870 89.6 125.7 123.7 158.8 198.8 880 90.5 126.8 124.6 159.6 199.6 890 91.5 127.9 125.4 160.3 200.3 900 92.4 128.9 126.2 161.1 201.1 910 93.4 129.9 127.0 161.8 201.8 920 94.3 130.9 127.7 162.5 202.5 930 95.1 131.9 128.5 163.2 203.2 940 96.0 132.9 129.3 163.9 203.9 B-23

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-4. Quad Cities Unit 2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT:- HEAD BELTLINE AT BELTLINE AT

`54 EFPY ' 54 EFPY 54 EFPY

'PRESSURE CURVEAA CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) -  : (,F) - (° - (°F)- (OF) 950 96.9 133.8 130.0 164.6 204.6 960 97.7 134.8 130.7 165.2 205.2 970 98.6 .135.7 131.5 165.9 205.9 980 99.4 136.6 132.2 166.5 206.5 990 100.2 137.4 132.9 167.2 207.2 1000 101.0 138.3 133.6 167.8 207.8 1010 101.7 139.1 134.2 168.5 208.5 1020 102.5 140.0 134.9 169.1 209.1 1030 103.3 140.8 135.6 169.7 209.7 1040 104.0 141.6 136.2 170.3 210.3 1050 104.7 142.4 136.9 170.9 210.9 1060 105.4 143.2 137.5 171.5 211.5 1070 106.2 144.0 138.1 172.1 212.1 1080 106.9 144.7 138.8 172.7 212.7 1090 107.6 145.5 139.4 173.2 213.2 1100 108.2 146.2 140.0 173.8 213.8 1105 108.6 146.6 140.3 174.1 214.1 1110 108.9 146.9 140.6 174.3 214.3 1120 109.6 147.6 141.2 174.9 214.9 1130 110.2 148.4 141.8 175.4 215.4 1140 110.9 149.1 142.3 176.0 216.0 1150 111.5 149.7 142.9 176.5 216.5 1160 112.1 150.4 143.5 177.0 217.0 1170 112.8 151.1 144.0 177.6 217.6 1180 113.4 151.7 144.6 178.1 218.1 B-24

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE B-4. Quad Cities Unit 2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-10. 5-13 and 5-14

-BOTTOM UPPER RPV &UPPER RPVPPER RPV & UPPER RPV &

HEAD BELTLINEAT HEAD -BELTLINE AT BELTLINEAT

--:-54 EFPY; 5 EFPY 54 EFPY PRESSURE CURVE A' CURVE A CURVE B CURVE B CURVE C (PSIG) (0F) - (0F- ((F- (0F) (0F 1190 114.0 152.4 145.1 178.6 218.6 1200 114.6 153.0 145.7 179.1 219.1 1210 115.2 153.7 146.2 179.6 219.6 1220 115.8 154.3 146.8 180.1 220.1 1230 116.3 154.9 147.3 180.6 220.6 1240 116.9 155.5 147.8 181.1 221.1 1250 117.5 156.1 148.3 181.6 221.6 1260 118.0 156.7 148.8 182.1 222.1 1270 118.6 157.3 149.3 182.5 222.5 1280 119.1 157.9 149.8 183.0 223.0 1290 119.7 158.5 150.3 183.5 223.5 1300 120.2 159.0 150.8 183.9 223.9 1310 120.7 159.6 151.3 184.4 224.4 1320 121.3 160.2 151.8 184.8 224.8 1330 121.8 160.7 152.2 185.3 225.3 1340 122.3 161.3 152.7 185.7 225.7 1350 122.8 161.8 153.2 186.2 226.2 1360 123.3 162.3 153.6 186.6 226.6 1370 123.8 162.9 154.1 187.0 227.0 1380 124.3 163.4 154.5 187.5 227.5 1390 124.8 163.9 155.0 187.9 227.9 1400 125.3 164.4 155.4 188.3 228.3 B-25

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version CA NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

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

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

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

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

C-2

GE Nuclear Energy GE-N E-OOOD-0002-960O-03R2a Non-Proprietary Version C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by <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-NuclearHeatup/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 200 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-N E-0000-0002-9600-03R2a 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)

C4

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX C

REFERENCES:

1. T.A. Caine, "Pressure-Temperature Curves Per Regulatory Guide 1.99, Revision 2 for the Dresden and Quad Cities Nuclear Power Stations", SASR 89-54, Revision 1, August 1989.

C-5

GE Nuclear Energy GE-NE-OOOD-0002-9600-03R2a Non-Proprietary Version APPENDIX D GE SIL 430 D-1

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 212oF for Tech steam pressure to from main steam instrument Spec 100OF/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 212 0F for Tech Must comply with SIL 251 Spec IOOOF/hr heatup to avoid vessel stratification.

and cooldown rate.

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

D-2

GE Nuclear Energy GE-N E-000O-0002-9600-03R2a 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 100IF/hr correlated RHR inlet temperature cooldown rate when in coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow. Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta T's will be indicated coolant temperature. Delta T limit is I00OF for BWR/6s and 145OF 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-000O-0002-960O0-3R2a 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 BWRI6s 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-NE-000O-0002-9600-03R2a 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-0000-0002-9600-03R2a Non-Proprietary Version Product

Reference:

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

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

SILs pertain only to GE BWRs. GE prepares SILs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, if any, of information contained in SILs to any plant or facility other than GE BWRs as designed and furnished by GE. Determination of applicability of information contained In any SIL to a specific GE BWR and implementation of recommended action are responsibilities of the owner of that GE BWR.SILs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of its owner. Such operation involves activities of which GE has no knowledge and over which GE has no control. Therefore, GE makes no warranty or representation expressed or implied with respect to the accuracy, completeness or usefulness of information contained In SlLs. 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-OOOD-0002-9600-03R2a Non-Proprietary Version APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a 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 1.0e17 n/cm2 . Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 n/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 # 2 - Top of Active Fuel (TAF): 360.3" (from vessel 0) [1]

Shell # 1 - Bottom of Active Fuel (BAF): 216.3" (from vessel 0) [1]

Top of Recirc Outlet Nozzle N1 in Shell # 1: 188" (from vessel 0) [2]

Centerline of Recirc Outlet Nozzle N1 in Shell # 1: 161.5" (from vessel 0) [3]

Top of Recirc Inlet Nozzle N2 in Shell # 1: 193.3" (from vessel 0) [2]

Centerline of Recirc Inlet Nozzle N2 in Shell # 1: 181" (from vessel 0) [3]

Girth Weld between Shell Ring #2 and Shell Ring #3: 391.5" (from vessel 0) [3,4]

From [2], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -23" below BAF and the top of the recirculation outlet nozzle is -28" below BAF), and no nozzles are within the BAF-TAF region of the reactor vessel. The girth weld between Shell Rings #2 and#3 is -31" above TAF. Therefore, if it can be shown that the peak fluence at these locations is less than 1.0e17 n/cm 2, it can be safely concluded that all nozzles and welds, other than E-2

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version those included in Tables 4-3 and 4-4, are outside the beltline region of the reactor vessel.

Based on the bounding 32 and 54 EFPY axial flux profile for pre- and post-EPU [5], the RPV fluence level dropped to less than 1.0e17 n/cm2 at the RPV ID at -1" below the BAF and at -6" above TAF. The beltline region considered in the development of the P-T curves is adjusted to include the additional 6" above the active fuel region and the additional 1" below the active fuel region. This adjusted beltline region extends from 215.3" to 366.3" above reactor vessel "0"for both 32 and 54 EFPY.

Based on the above, it is concluded that none of the Quad Cities Unit 2 reactor vessel nozzles or welds, other than those included in Tables 4-3 and 4-4, are in the beltline region.

E-3

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX E

REFERENCES:

1. Dresden/Quad Cities LR PT Curves - Design Input Request (DIR),

Robert Stachniak (Exelon), 4/26/02.

2. Babcock & Wilcox Co. (B&VV) Drawing # 151833E, Revision 1, "Recirculation Nozzles", (GE-NE VPF# 1744-138-3), Quad Cities Units 1&2.

3. Babcock & Wilcox Co. (B&W) Drawing # 26906F, Revision 7, "General Outline", (GE-NE VPF# 1744-145-8), Quad Cities Units 1 &2.

4. Babcock & Wilcox Co. (B&W) Drawing # 151825F, Revision 1, uShell course Sub-Assembly", (GE-NE VPF# 1744-132-4), Quad Cities Units I & 2.

5. S. Sitaraman, "Dresden and Quad Cities Neutron Flux Evaluation", GE-NE, San Jose, CA, March 2003, (GE-NE-0000-0011-0531-R2, Revision 2)(GE Proprietary Information).

E-4

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Prop'rietary Version APPENDIX F CORE NOT CRITICAL CALCULATION FOR THE BOTTOM HEAD (CRD PENETRATION)

F-1

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE OF CONTENTS The following outline describes the contents of this Appendix:

F.1 Executive Summary F.2 Scope F.3 Analysis Methods F.3.1 Applicability of the ASME Code Appendix G Methods F.3.2 Finite Element Fracture Mechanics Evaluation F.3.3 ASME Code Appendix G Evaluation F.4 Results F.5 Conclusions F.6 References F.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 [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. ((

F-2

GE Nuclear Energy GE-NE-O0ob0002-9600-03R2a Non-Proprietary Version F.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 [1]. This Appendix discusses the finite element analysis and the Appendix G [1] calculations separately below.

F.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 Xl, Appendix G methods [1] considering the specific bottom head geometry is discussed first, followed by a detailed discussion of the finite element analysis and Appendix G evaluation [1].

F.3.1 Applicabilityof the ASME Code Appendix G Methods The methods described in the ASME Code Section Xl, Appendix G [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 % 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 F-3

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version more detailed (( )) finite element fracture mechanics analysis ((

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

F.3.2 Finite Element Fracture Mechanics Evaluation An advanced (( )) finite element analysis of a BWR bottom head geometry was performed to determine the mode I stress intensity at the tip of a % thickness postulated flaw. ((

1))

Finite Elements ((f]

All Finite Element Analyses were performed using ANSYS Version 6.1 [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.

F-4

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

))

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-0 F) units are:

F-5

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version

[C

))

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.

Loads Two loads cases were independently analyzed.

F-6

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version

1. Pressure Loadina -

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. (( ]l Thermal Transient

((

1]

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-OF) and temperatures (T) are in OF.

F-7

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version ha Figure 3. Regions To WMich Thermal Loads Are Applied

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

b. Regions 2 and 3:

Time (min) 2 h3 T 0 496 413

(( 341 354 496 413 (( ]

3] 496 413 ]

3))

Temperature Plot vs. Time (min.)

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

d. Region 5: h = 0.2, T = 100 The peak thermal gradients were used to compute the thermal stresses based on a uniform reference temperature of 70 0F.

F-8

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a 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, (( ))

[c The cracks considered for this analysis ((

))

F-9

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version Er I

))

F-10

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

[I

• )) I F-I1

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Stress Intensity Factor Computation

((

1]

F-12

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version I

3]

Benchmarking (( ]1 Methodology

((

F-13

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

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

Pressure Loading Analysis Results

((

. II F-14

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Benchmarking Of Pressure Loading Results Pressure Loading analyses ((

))

F-15

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version

• ] I F-16

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1[

)) I F-17

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a 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 l.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. ((

31 Thermal Gradients [ ]

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

F-18

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version

[I

,, I' 1'I F-19

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version It can be seen that the peak times and values based on each gradient are:

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

' 1[

I I ))

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.

3]

F.3.3 ASME Code Appendix G Evaluation The peak stress intensities for the pressure and thermal load cases evaluated above are used as inputs to the ASME Code Appendix G evaluation methodology [1] to calculate a T-RTNDT. The Core Not Critical Bottom Head P-T curve T-RTNDT is calculated using the formulas listed below:

F-20

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version K1 = SFpK 1 p + SFT.Ylt SF =2.0 p

SF= 1.0 T-R K-332.')

K TRTNDT 20.734 0.02 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, Note that the stress intensity is defined in units of: ksi*inlr2 F.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:

))

F-21

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Note that the pressure stress intensity has been adjusted by the factor [f 3] to account for the vessel pressure at which the maximum thermal stress occurred. The finite element results summarized above were calculated using a vessel pressure f[

))

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

))

F.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 ((

R

]

F.6 References

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

2. ANSYS User's Manual, Version 6.1.

F-22

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version APPENDIX G BOUNDING P-T CURVES FOR QUAD CITIES UNITS I & 2 G-1

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version This appendix contains P-T curves that bound the limiting material characteristics of both Quad Cities Unit 1 and Quad Cities Unit 2. Composite and individual curves are presented for both 32 and 54 EFPY similar to those provided within the main body of this report. Table G-1 provides the figure numbers and the corresponding tabulation for each P-T curve presented in this appendix.

G-2

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version Table G-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure Table Numbers Curve urve escripion NmberS for, o Presentation of Presentation of

________ ____>e___________ the,-T <Curves the P-TCurvesl Curve A Bottom Head Limits (CRD Nozzle) Figure G-1 Table G-2 & 4 Upper Vessel Limits (FW Nozzle) Figure G-2 Table G-2 & 4 Beltline Limits for 32 EFPY Figure G-3 Table G-2 Beitline Limits for 54 EFPY Figure G-4 Table G-4 Curve B Bottom Head Limits (CRD Nozzle) Figure G-5 Table G-2 & 4 Upper Vessel Limits (FW Nozzle) Figure G-6 Table G-2 & 4 Beltline Limits for 32 EFPY Figure G-7 Table G-2 Beltline Limits for 54 EFPY Figure G-8 Table G-4 Curve C _

Composite Curve for 32 EFPY** Figure G-9 Table G-3 Composite Curve for 54 EFPY** Figure G-10 Table G-5 A &B Composite Curves for 32 EFPY Bottom Head and Composite Curve A Figure G-1 1 Table G-3 for 32 EFPY*

Bottom Head and Composite Curve B Figure G-12 Table G-3 for 32 EFPY*

A &B Composite Curves for 54 EFPY Bottom Head and Composite Curve A Figure G-13 Table G-5 for 54 EFPY*

Bottom Head and Composite Curve B Figure G-14 Table G-5 for 54 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.

G-3

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100 n

C, 1 000

0. 900 INITIAL RTndt VALUE IS 600F FOR BOTTOM HEAD o 800 III o% 700 HEATUP/COOLDOWN RATE OF COOLANT

600 < 200F/HR 0: 600 w

I-t) 400 0.

300 200 100 0

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

Figure G-1: Bounding Quad Cities 1&2 Bottom Head P-T Curve for Pressure Test [Curve A] [200F/hr or less coolant heatup/cooldown]

G-4

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100

&Is in C 1000 0- 900 0 INITIAL RTndt VALUE IS 148°F FOR UPPER VESSELI IK-60 a) 800 o 700 HEATUPICOOLDOWN LU. RATE OF COOLANT W 600 < 20°1IHR 2

500 1

w U1 U) 400 Lu Ix 300 200

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

(°F)

Figure G-2: Bounding Quad Cities 1&2 Upper Vessel P-T Curve for Pressure Test [Curve A]

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

G-5

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUE IS 23.10F FOR BELTLINE 1100 D 1000 0 900 0

-II BELTLINE CURVE co 800 ADJUSTED AS SHOWN:

co EFPY SHIFT (°F) 32 63 o 700 L-HEATUP/COOLDOWN w 600 RATE OF COOLANT 2

' 20*F/HR M 500 V) 400 0

300 200 100 0

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

Figure G-3: Bounding Quad Cities 1&2 Beitline P-T Curve for Pressure Test [Curve A] up to 32 EFPY [20 0F/hr or less coolant heatup/cooldown]

G-6

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200

.. INITIAL RTndt VALUE IS 23.1°F FOR BELTUNE 1100

• II0 1000

-j ILl 900 0 BELTLINE CURVE 0

800 ADJUSTED AS SHOWN:

EFPY SHIFT ('F)

I- 54 81 a 700 to m

HEATUP/COOLDOWN n 600 RATE OF COOLANT um 500 II

_S 31 _ _,_ , IC __

< 200F/HR 400 c

300 200 ~_____SL_ __ __ __TU

_ _F 100

-~i SIIIIISII~ _

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

Figure G4: Bounding Quad Cities 1&2 Beltline P-T Curve for Pressure Test [Curve A] up to 54 EFPY [20°F/hr or less coolant heatup/cooldown]

G-7

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 __ ____ 1 1200

.1100

.0 in C1000 I I I I1-0 900 / INITIAL RTndt VALUE IS C) 800 / 174.6 0F FOR BOTTOM HEAD I o 700 HEATUPICOOLDOWN I- RATE OF COOLANT c 1000F/HR a 600 z

Z 500 _ _ _ 7 (0

a: 400 HEADI 300 200 100 +

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

Figure G-5: Bounding Quad Cities 1&2 Bottom Head P-T Curve for Core Not Critical [Curve B]

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

G-8

GE Nuclear Energy GE-N E-OOOO-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 1100 0.

in a- 1 000 IL 900 INITIAL RTndt VALUE IS 0 48 0F FOR UPPER VESSEL a) 800 ax ot 700 HEATUPICOOLDOWN RATE OF COOLANT I9 600 500 i X400 en 0:

3L0 3)00 200 100 0

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

Figure G-8: Bounding Quad Cities 1&2 Beltline P-T Curve for Core Not Critical [Curve B] up to 54 EFPY

[100°Ffhr or less coolant heatup/cooldown]

G-11

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 23.16F FOR BELTLINE, 48*F FOR UPPER VESSEL, 1200 AND 60°F FOR BOTTOM HEAD 1100

& 1000 BELTUNE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (0F)

a. 900 32 63 0

-j II Xn 800 HEATUP/COOLDOWN RATE OF COOLANT l 1O00F1HR o 700 a 600 Z

m j 500 LU to 0~

400 cL 300 200 NELTUNE AND NON-BELTLINE 100 LIMITS 0

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

Figure G-9: Bounding Quad Cities 1&2 Core Critical P-T Curves [Curve C]

up to 32 EFPY [100°F/hr or less coolant heatup/cooldown]

G-12

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 23.1OF FOR BELTLINE, 48*F FOR UPPER.

VESSEL, 1200 AND 600F FOR BOTTOM HEAD 1100 of BELTUNE CURVE c1000 a ADJUSTED AS SHOWN:

EFPY SHIFT (°F) c- 900 54 81 0

I-w (n 800 HEATUP/COOLDOWN u)

RATE OF COOLANT c 100OFIHR o 700 U-a: 600 z

3i U.'

5DO en 400 300 200 BELTLINE AND NON-BELTLINE 100 LIMITS 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTORVESSELMETALTEMPERATURE OF)

Figure G-10: Bounding Quad Cities 1&2 Core Critical P-T Curves [Curve C]

up to 54 EFPY [100OF/hr or less coolant heatup/cooldown]

G-13

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 23.1°F FOR BELTLINE, 48@F FOR UPPER VESSEL, AND 1100 60@F FOR BOTTOM HEAD Is D 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

0. 900 32 63 0

LU U_ 800 U'

o

-J 700 HEATUPICOOLDOWN RATE OF COOLANT 3600 .5200FIHR z

3 500 10 400 LU 300

-UPPER VESSEL 200 AND BELTLINE LIMITS

---

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

Figure G-11: Bounding Quad Cities 1&2 Composite Pressure Test P-T Curves [Curve A]

up to 32 EFPY [20 °F/hr or less coolant heatup/cooldown]

G-14

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1*F FOR BELTLINE, 48°F FOR UPPER VESSEL, 1100 AND 74.6*F FOR BOTTOM HEAD CL2cX1000 BELTLINE CURVES Is ADJUSTED AS SHOWN:

0 900 EFPY SHIFT (0F) 32 63 0 0 IL 600 R

-z 200 HEATUP/COOLDOWN RATE OF COOLANT

0. :S I00°F1HR 3"00 2n00

-UPPER VESSEL AND BELTLINE LIMITS 100 BOTTOM HEAD -

CURVE 0

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

(°F)

Figure G-12: Bounding Quad Cities 1&2 Composite Core Not Critical P-T Curves [Curve B]

up to 32 EFPY [100OF/hr or less coolant heatup/cooldown]

G-15

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 23.1PF FOR BELTLINE.

48¶F FOR UPPER VESSEL, AND 1100 600F FOR BOTTOM HEAD C,.

1000 BELTLINE CURVES 0

0 ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 900 54 81 0

-J aL 800 Ua IL R

700 0 HEATUP/COOLDOWN LU RATE OF COOLANT 600 .<200 F/HR aW 500 0

400 300

-UPPER VESSEL 200 AND BELTLINE LIMITS

- - BOTTOM HEAD -

100 CURVE 0

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

Figure G-13: Bounding Quad Cities 1&2 Composite Pressure Test P-T Curves [Curve A]

up to 54 EFPY [20°F/hr or less coolant heatup/cooldown]

G-16

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1°F FOR BELTLINE, 48¶F FOR UPPER VESSEL, 1100 AND 74.6*F FOR BOTTOM HEAD In D 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

M 900 EFPY SHIFT (°F) 0 54 81 I-J W) 800 to o 700 14.- HEATUP/COOLDOWN RATE OF COOLANT a: 600 c 1OF/HR 2

I-_

> 500 a) 400 Ua 9L 300

-UPPER VESSEL 200 AND BELTLINE LIMITS HEAD .

100 CURVE 0

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

Figure G-14: Bounding Quad Cities 1&2 Composite Core Not Critical P-T Curves [Curve B]

up to 54 EFPY [100lF/hr or less coolant heatup/cooldown]

G-17

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-2. Bounding Quad Cities 1&2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 OF/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6 and G-7

-BOTTOM UPPER - - 32 EFPY BOTTOM UPPER -- 32 EFPY

'HEAD VESSEL BELTLINE - HEAD -. VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVEA - CURVE B CURVE B - CURVE B -

(PSIG) (0 F) (°F - --(0F) (0 F F-- (OF) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 83.1 83.1 140 68.0 83.1 83.1 68.0 85.4 83.1 150 68.0 83.1 83.1 68.0 88.2 83.1 160 68.0 83.1 83.1 68.0 90.9 83.1 170 68.0 83.1 83.1 68.0 93.5 83.1 180 68.0 83.1 83.1 68.0 95.9 83.1 190 68.0 83.1 83.1 68.0 98.2 83.1 200 68.0 83.1 83.1 68.0 100.3 83.1 210 68.0 83.1 83.1 68.0 102.3 83.1 220 68.0 83.1 83.1 68.0 104.3 83.1 230 68.0 83.1 83.1 68.0 106.1 83.1 240 68.0 83.1 83.1 68.0 107.9 83.1 G-18

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-2. Bounding Quad Cities 1&2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 0F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6 and G-7

-BOlTOM -'UPPER"' 32EFPY BOTTOM UPPER'- 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A '-'CURVE A CURVE B CURVE B ' CURVE B (PSIG) (Of)  :( 0F) (0f (OF) (0F) (0'F) 250 68.0 83.1 83.1 68.0 109.6 83.1 260 68.0 83.1 83.1 68.0 111.2 83.1 270 68.0 83.1 83.1 68.0 112.8 83.1 280 68.0 83.1 83.1 68.0 114.3 83.1 290 68.0 83.1 83.1 68.0 115.8 83.1 300 68.0 83.1 83.1 68.0 117.2 83.1 310 68.0 83.1 83.1 68.0 118.5 83.1 312.5 68.0 83.1 83.1 68.0 118.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.1 143.1 143.1 430 68.0 113.1 113.1 71.4 143.1 143.1 440 68.0 113.1 113.1 74.4 143.1 143.1 450 68.0 113.1 113.1 77.3 143.1 143.1 460 68.0 113.1 113.1 80.0 143.1 143.1 470 68.0 113.1 113.1 82.6 143.1 143.1 480 68.0 113.1 113.1 85.1 143.1 143.1 490 68.0 113.1 113.1 87.4 143.1 143.1 G-19

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-2. Bounding Quad Cities 1&2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6 and G-7

-BOTTOM

-. . - UPPER : ;32 EFPY BOTTOM UPPER - 32 EFPY HEAD -. VESSEL BELTLINE HEAD - VESSEL BELTLINE PRESSURE. CURVEA CURVE A 'CURVE A CURVE B CURVE B' CURVE B (PSIG) (°F) (°F) (°F (°F- (°F- (°F 500 68.0 113.1 113.1 89.6 143.1 143.1 510 68.0 113.1 113.1 91.8 143.1 143.1 520 68.0 113.1 113.1 93.8 143.1 143.1 530 68.0 113.1 113.1 95.8 143.1 143.1 540 68.0 113.1 113.1 97.7 143.1 143.1 550 68.0 113.1 113.1 99.5 143.1 143.1 560 68.0 113.1 113.1 101.3 143.4 143.1 570 68.0 113.1 113.1 103.0 144.1 143.1 580 68.0 113.1 113.1 104.6 144.9 143.1 590 68.0 113.1 113.1 106.2 145.6 143.1 600 68.0 113.1 113.1 107.8 146.1 143.1 610 68.0 113.1 113.1 109.3 146.6 143.1 620 68.0 113.1 113.1 110.7 147.0 143.1 630 68.6 113.1 113.1 112.1 147.4 143.1 640 70.5 113.1 113.1 113.5 147.8 143.1 650 72.2 113.1 113.1 114.8 148.2 143.1 660 73.9 113.1 113.1 116.1 148.7 143.1 670 75.6 113.1 113.1 117.4 149.1 143.1 680 77.2 113.1 113.1 118.7 149.5 143.1 690 78.7 113.1 113.1 119.9 149.9 143.1 700 80.2 113.1 113.1 121.0 150.3 143.1 710 81.7 113.1 113.1 122.2 150.7 143.1 720 83.1 113.1 113.1 123.3 151.1 143.1 730 84.5 113.1 113.1 124.4 151.5 143.1 740 85.8 113.1 113.1 125.5 151.9 143.1 750 87.1 113.1 113.1 126.6 152.2 143.1 760 88.4 113.1 113.1 127.6 152.6 143.1 G-20

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-2. Bounding Quad Cities 1&2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OFthr for Curves B & C and 20 OF/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6 and G-7 BOlTOM UPPER;- 32 EFPY - BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD. -VESSEL BELTLINE PRESSURE -CURVEA CURVE A CURVE A CURVEB CURVEB CURVEB (PSIG) (OF) ( -F) (.F- (0F (0F (0F) 770 89.6 113.6 113.1 128.6 153.0 143.1 780 90.8 114.3 113.1 129.6 153.4 143.1 790 92.0 115.1 113.1 130.6 153.8 143.1 800 93.2 115.9 113.1 131.5 154.1 143.1 810 94.3 116.6 113.1 132.5 154.5 143.1 820 95.4 117.4 113.1 133.4 154.9 143.1 830 96.5 118.1 113.1 134.3 155.2 143.1 840 97.5 118.8 113.1 135.2 155.6 143.1 850 98.6 119.5 113.1 136.0 155.9 143.1 860 99.6 120.2 113.1 136.9 156.3 143.1 870 100.6 120.9 113.1 137.7 156.6 143.1 880 101.5 121.6 113.1 138.6 157.0 143.1 890 102.5 122.3 113.1 139.4 157.3 143.1 900 103.4 122.9 113.1 140.2 157.7 143.1 910 104.4 123.6 113.1 141.0 158.0 143.8 920 105.3 124.2 113.1 141.7 158.4 144.5 930 106.1 124.9 113.9 142.5 158.7 145.2 940 107.0 125.5 114.9 143.3 159.0 145.9 950 107.9 126.1 115.8 144.0 159.4 146.6 960 108.7 126.7 116.8 144.7 159.7 147.2 970 109.6 127.3 117.7 145.5 160.0 147.9 980 110.4 127.9 118.6 146.2 160.4 148.5 990 111.2 128.5 119.4 146.9 160.7 149.2 1000 112.0 129.1 120.3 147.6 161.0 149.8 1010 112.7 129.7 121.1 148.2 161.3 150.5 1020 113.5 130.2 122.0 148.9 161.6 151.1 1030 114.3 130.8 122.8 149.6 162.0 151.7 G-21

GE Nuclear Energy GE-NE-000O-0002-9600-03R2a Non-Proprietary Version TABLE G-2. Bounding Quad Cities 1&2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6 and G-7

- -BOTOM - EFPY BOTTOM .UPPER

.32 UPPER'; 32;EFP

,HEAD - VESSEL BELTLINE HEAD 'VESSEL' BELTLINE.:

,,PRESSURE CURVE V A CURVE B' CURVE B CURVE B

..'(PSIG) (OF): (OF) (°F) ' (°F-'(

1040 115.0 131.4 123.6 150.2 162.3 152.3 1050 115.7 131.9 124.4 150.9 162.6 152.9 1060 116.4 132.5 125.2 151.5 162.9 153.5 1070 117.2 133.0 126.0 152.1 163.2 154.1 1080 117.9 133.5 126.7 152.8 163.5 154.7 1090 118.6 134.1 127.5 153.4 163.8 155.2 1100 119.2 134.6: 128.2 154.0 164.1 155.8 1105 119.6 134.8 128.6 154.3 164.3 156.1 1110 119.9 135.1 128.9 154.6 164.4 156.3 1120 120.6 135.6 129.6 155.2 164.7 156.9 1130 121.2 136.1 130.4 155.8 165.0 157.4 1140 121.9 136.6 131.1 156.3 165.3 158.0 1150 122.5 137.1 131.7 156.9 165.6 158.5 1160 123.1 137.6 132.4 157.5 165.9 159.0 1170 123.8 138.1 133.1 158.0 166.2 159.6 1180 124.4 138.6 133.7 158.6 166.5 160.1 1190 125.0 139.1 134.4 159.1 166.7 160.6 1200 125.6 139.5 135.0 159.7 167.0 161.1 1210 126.2 140.0 135.7 160.2 167.3 161.6 1220 126.8 140.5 136.3 160.8 167.6 162.1 1230 127.3 140.9 136.9 161.3 167.9 162.6 1240 127.9 141.4 137.5 161.8 168.2 163.1 1250 128.5 141.8 138.1 162.3 168.4 163.6 1260 129.0 142.3 138.7 162.8 168.7 164.1 1270 129.6 142.7 139.3 163.3 169.0 164.5 1280 130.1 143.2 139.9 163.8 169.2 165.0 1290 130.7 143.6 140.5 164.3 169.5 165.5 G-22

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-2. Bounding Quad Cities 1&2 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B &C and 20 eF/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6 and G-7

-BOTOM UPPER 32 EFPY BOTTOM - UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL -: BELTLINE PRESSURE CURVEA -CURVEA - CURVEAA CURVEBB CURVEB  ; CURVE B -

(PSIG) -(FF (0F)  :(- - (F (F) - F 1300 131.2 144.0 141.0 164.8 169.8 165.9 1310 131.7 144.5 141.6 165.3 170.1 166.4 1320 132.3 144.9 142.2 165.8 170.3 166.8 1330 132.8 145.3 142.7 166.2 170.6 167.3 1340 133.3 145.7 143.3 166.7 170.8 167.7 1350 133.8 146.1 143.8 167.2 171.1 168.2 1360 134.3 146.6 144.3 167.6 171.4 168.6 1370 134.8 147.0 144.9 168.1 171.6 169.0 1380 135.3 147.4 145.4 168.5 171.9 169.5 1390 135.8 147.8 145.9 169.0 172.1 169.9 1400 136.3 148.2 146.4 169.4 172.4 170.3 G-23

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & Cand 20 0F/hr for Curve A for Figures G-9, G-11 and G-12 BOTTOM- UPPER RPV & BOTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY XPRESSURE CURVE A CURVEA CURVE B CURVE B CURVE C-(PSIG) (on ( -F) (F) (f (

0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 88.0 70 68.0 83.1 68.0 83.1 95.2 80 68.0 83.1 68.0 83.1 101.2 90 68.0 83.1 68.0 83.1 106.3 100 68.0 83.1 68.0 83.1 110.8 110 68.0 83.1 68.0 83.1 114.9 120 68.0 83.1 68.0 83.1 118.7 130 68.0 83.1 68.0 83.1 122.2 140 68.0 83.1 68.0 85.4 125.4 150 68.0 83.1 68.0 88.2 128.2 160 68.0 83.1 68.0 90.9 130.9 170 68.0 83.1 68.0 93.5 133.5 180 68.0 83.1 68.0 95.9 135.9 190 68.0 83.1 68.0 98.2 138.2 200 68.0 83.1 68.0 100.3 140.3 210 68.0 83.1 68.0 102.3 142.3 G-24

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-11 and G-12 BOTTOM UPPER RPV &, BOTTOM UPPER RPV & UPPER RPV &

-HEAD BELTLI NEAT A TN A -BELT NE AT

'32 EFPY 32 EFPY '- 32 EFPY

'PRESSURE CURVE ACURVEA -, CURVE B CURVE B -'.CURVE C:

(PSIG) " (( F): ' .( (OF).. (OF) 220 68.0 83.1 68.0 104.3 144.3 230 68.0 83.1 68.0 106.1 146.1 240 68.0 83.1 68.0 107.9 147.9 250 68.0 83.1 68.0 109.6 149.6 260 68.0 83.1 68.0 111.2 151.2 270 68.0 83.1 68.0 112.8 152.8 280 68.0 83.1 68.0 114.3 154.3 290 68.0 83.1 68.0 115.8 155.8 300 68.0 83.1 68.0 117.2 157.2 310 68.0 83.1 68.0 118.5 158.5 312.5 68.0 83.1 68.0 118.9 158.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.1 143.1 183.1 G-25

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures G-9, G-11 and G-12 BOTTOM UPPER R-PV & BOTTOM UPPER RPV &' UPPER RPV&

HEAD BELTLINE AT, HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY. 32 EFPY

!PRESSURE CURVEA CURVE A CURVE B CURVE B: CURVE C

'(PSIG) (0F ' (° ' - (OF)' (F ' (° 430 68.0 113.1 71.4 143.1 183.1 440 68.0 113.1 74.4 143.1 183.1 450 68.0 113.1 77.3 143.1 183.1 460 68.0 113.1 80.0 143.1 183.1 470 68.0 113.1 82.6 143.1 183.1 480 68.0 113.1 85.1 143.1 183.1 490 68.0 113.1 87.4 143.1 183.1 500 . 68.0 113.1 89.6 143.1 183.1 510 68.0 113.1 91.8 143.1 183.1 520 68.0 113.1 93.8 143.1 183.1 530 68.0 113.1 95.8 143.1 183.1 540 68.0 113.1 97.7 143.1 183.1 550 68.0 113.1 99.5 143.1 183.1 560 68.0 113.1 101.3 143.4 183.4 570 68.0 113.1 103.0 144.1 184.1 580 68.0 113.1 104.6 144.9 184.9 590 68.0 113.1 106.2 145.6 185.6 600 68.0 113.1 107.8 146.1 186.1 610 68.0 113.1 109.3 146.6 186.6 620 68.0 113.1 110.7 147.0 187.0 630 68.6 113.1 112.1 147.4 187.4 640 70.5 113.1 113.5 147.8 187.8 650 72.2 113.1 114.8 148.2 188.2 G-26

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 IF/hr for Curves B & C and 20 0F/hr for Curve A for Figures G-9, G-1I and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

-.HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVEA 'CURVEB CURVE B CURVE C (PSIG) (F-) (FJ  ; (an (° - (°F) 660 73.9 113.1 116.1 148.7 188.7 670 75.6 113.1 117.4 149.1 189.1 680 77.2 113.1 118.7 149.5 189.5 690 78.7 113.1 119.9 149.9 189.9 700 80.2 113.1 121.0 150.3 190.3 710 81.7 113.1 122.2 150.7 190.7 720 83.1 113.1 123.3 151.1 191.1 730 84.5 113.1 124.4 151.5 191.5 740 85.8 113.1 125.5 151.9 191.9 750 87.1 113.1 126.6 152.2 192.2 760 88.4 113.1 127.6 152.6 192.6 770 89.6 113.6 128.6 153.0 193.0 780 90.8 114.3 129.6. 153.4 193.4 790 92.0 115.1 130.6 153.8 193.8 800 93.2 115.9 131.5 154.1 194.1 810 94.3 116.6 132.5 154.5 194.5 820 95.4 117.4 133.4 154.9 194.9 830 96.5 118.1 134.3 155.2 195.2 840 97.5 118.8 135.2 155.6 195.6 850 98.6 119.5 136.0 155.9 195.9 860 99.6 120.2 136.9 156.3 196.3 870 100.6 120.9 137.7 156.6 196.6 880 101.5 121.6 138.6 157.0 197.0 G-27

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B &C and 20 °F/hr for Curve A for Figures G-9, G-11 and G-12

- - BOTTOM *'UPPER RPV & BOtTOM -UPPERRPV& UPPER RPV&

HEAD BELTLINE AT ->-HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY -32 EFPY

!:PRESSURE- CURVEA CURVE A-- CURVE B CURVE B CURVEC

-(PSIG) - (°F (°F) - (°- (°F - (°F) 890 102.5 122.3 139.4 157.3 197.3 900 103.4 122.9 140.2 157.7 197.7 910 104.4 123.6 141.0 158.0 198.0 920 105.3 124.2 141.7 158.4 198.4 930 106.1 124.9 142.5 158.7 198.7 940 107.0 125.5 143.3 159.0 199.0 950 107.9 126.1 144.0 159.4 199.4 960 108.7 126.7 144.7 159.7 199.7 970 109.6 127.3 145.5 160.0 200.0 980 110.4 127.9 146.2 160.4 200.4 990 111.2 128.5 146.9 160.7 200.7 1000 112.0 129.1 147.6 161.0 201.0 1010 112.7 129.7 148.2 161.3 201.3 1020 113.5 130.2 148.9 161.6 201.6 1030 114.3 130.8 149.6 162.0 202.0 1040 115.0 131.A 150.2 162.3 202.3 1050 115.7 131.9 150.9 162.6 202.6 1060 116.4 132.5 151.5 162.9 202.9 1070 117.2 133.0 152.1 163.2 203.2 1080 117.9 133.5 152.8 163.5 203.5 1090 118.6 134.1 153.4 163.8 203.8 1100 119.2 134.6 154.0 164.1 204.1 1105 119.6 134.8 154.3 164.3 204.3 G-28

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 ¶F/hr for Curves B & C and 20 0F/hr for Curve A for Figures G-9, G-1 1 and G-12 BOTTOM UPPER RPV 8 BOTTOM UPPER RPV & UPPER RPV &

HEAD:' BELTLINE AT e HEAD BELTLINE AT BELTLINE AT

-32 EFPY 32 EFPY 32 EFPY i PRESSURE- CURVE A 'CURVE A CURVE B CURVE B CURVE C (PSIG) 0

( F) - (OF)'- (°F) -- °F) (°--F) 1110 119.9 135.1 154.6 164.4 204.4 1120 120.6 135.6 155.2 164.7 204.7 1130 121.2 136.1 155.8 165.0 205.0 1140 121.9 136.6 156.3 165.3 205.3 1150 122.5 137.1 156.9 165.6 205.6 1160 123.1 137.6 157.5 165.9 205.9 1170 123.8 138.1 158.0 166.2 206.2 1180 124.4 138.6 158.6 166.5 206.5 1190 125.0 139.1 159.1 166.7 206.7 1200 125.6 139.5 159.7 167.0 207.0 1210 126.2 140.0 160.2 167.3 207.3 1220 126.8 140.5 160.8 167.6 207.6 1230 127.3 140.9 161.3 167.9 207.9 1240 127.9 141.4 161.8 168.2 208.2 1250 128.5 141.8 162.3 168.4 208.4 1260 129.0 142.3 162.8 168.7 208.7 1270 129.6 142.7 163.3 169.0 209.0 1280 130.1 143.2 163.8 169.2 209.2 1290 130.7 143.6 164.3 169.5 209.5 1300 131.2 144.0 164.8 169.8 209.8 1310 131.7 144.5 165.3 170.1 210.1 1320 132.3 144.9 165.8 170.3 210.3 1330 132.8 145.3 166.2 170.6 210.6 G-29

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-3. Bounding Quad Cities 1&2 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A for Figures G-9, G-11 and G-12

BOTTOM UPPER RPV'&` BOTTOM UPPER RPV8& UPPER RPV&

HEAD BELTLINE AT. HEAD BELTLINE AT - BELTLINE AT-

-3232 EFP.32 EFPY 32 EFPY

>-PRESSURE CURVEA CURVE A -CURVE B CURVE B CURVE C:'

(PSIG) - (0F) ' (°F) (-F) (OF) (0F) 1340 133.3 145.7 166.7 170.8 210.8 1350 133.8 146.1 167.2 171.1 211.1 1360 134.3 146.6 167.6 171.4 211.4 1370 134.8 147.0 168.1 171.6 211.6 1380 135.3 147.4 168.5 171.9 211.9 1390 135.8 147.8 169.0 172.1 212.1 1400 136.3 148.2 169.4 172.4 212.4 G-30

GE Nuclear Energy GE-NE-OOOD-0002-9600-03R2a Non-Proprietary Version TABLE G4. Bounding Quad Cities 1&2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6 and G-8 BOTTOM .UPPER 54 EFPY BOOM UPPER 54 EFPY HEAD - VESSEL BELTLINE HEAD :VESSEL- BELTLINE PRESSURE CURVEAA CURVEA - CURVEA - CURVE B CURVE B CURVE B.

(PSIG) (-F) (0F - (0 F- (° -- (°F- (F) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 83.1 83.1 140 68.0 83.1 83.1 68.0 85.4 83.1 150 68.0 83.1 83.1 68.0 88.2 83.1 160 68.0 83.1 83.1 68.0 90.9 83.1 170 68.0 83.1 83.1 68.0 93.5 83.1 180 68.0 83.1 83.1 68.0 95.9 83.1 190 68.0 83.1 83.1 68.0 98.2 83.1 200 68.0 83.1 83.1 68.0 100.3 83.1 210 68.0 83.1 83.1 68.0 102.3 83.1 220 68.0 83.1 83.1 68.0 104.3 83.1 230 68.0 83.1 83.1 68.0 106.1 83.1 240 68.0 83.1 83.1 68.0 107.9 83.1 G-31

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE G4. Bounding Quad Cities 1&2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6 and G-8 BOTTOM UPPER 54 EFPY BOTOM -UPPER -- 54 EFPY HEAD - VESSEL BELTLINE HEAD VESSEL::. BELTLINE.

PRESSURE - CURVE A. CURVE A CURVEA CURVE B CURVE B CURVE B (PSIG): - (OF) (°F:) ( - (°F (°F) 250 68.0 83.1 83.1 68.0 109.6 83.1 260 68.0 83.1 83.1 68.0 111.2 83.1 270 68.0 83.1 83.1 68.0 112.8 83.1 280 68.0 83.1 83.1 68.0 114.3 83.1 290 68.0 83.1 83.1 68.0 115.8 83.1 300 68.0 83.1 83.1 68.0 117.2 83.1 310 68.0 83.1 83.1 68.0 118.5 83.1 312.5 68.0 83.1 83.1 68.0 118.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.1 143.1 143.1 430 68.0 113.1 113.1 71.4 143.1 143.1 440 68.0 113.1 113.1 74.4 143.1 143.1 450 68.0 113.1 113.1 77.3 143.1 143.1 460 68.0 113.1 113.1 80.0 143.1 143.1 470 68.0 113.1 113.1 82.6 143.1 143.1 480 68.0 113.1 113.1 85.1 143.1 143.1 490 68.0 113.1 113.1 87.4 143.1 143.1 G-32

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G4. Bounding Quad Cities 1&2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & Cand 20 OF/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6 and G-8

.3BOTTOM UPPER 54 EFPY BOTOM -UPPER 4 EFPY HEAD:- VESSEL - BELTLINE' HEAD - VESSEL. BELTLINE PRESSURE CURVEA: CURVEA CURVEA CURVE B CURVE B.: CURVE B (PSIG) (°FF (0F- (OF) (((FF ()

500 68.0 113.1 113.1 89.6 143.1 143.1 510 68.0 113.1 113.1 91.8 143.1 143.1 520 68.0 113.1 113.1 93.8 143.1 143.1 530 68.0 113.1 113.1 95.8 143.1 143.1 540 68.0 113.1 113.1 97.7 143.1 143.1 550 68.0 113.1 113.1 99.5 143.1 143.1 560 68.0 113.1 113.1 101.3 143.4 143.1 570 68.0 113.1 113.1 103.0 144.1 143.1 580 68.0 113.1 113.1 104.6 144.9 143.1 590 68.0 113.1 113.1 106.2 145.6 143.1 600 68.0 113.1 113.1 107.8 146.1 143.1 610 68.0 113.1 113.1 109.3 146.6 143.1 620 68.0 113.1 113.1 110.7 147.0 143.1 630 68.6 113.1 113;1 112.1 147.4 143.1 640 70.5 113.1 113.1 113.5 147.8 143.1 650 72.2 113.1 113.1 114.8 148.2 143.1 660 73.9 113.1 113.1 116.1 148.7 143.1 670 75.6 113.1 113.1 117.4 149.1 143.1 680 77.2 113.1 113.1 118.7 149.5 143.1 690 78.7 113.1 113.1 119.9 149.9 143.1 700 80.2 113.1 113.1 121.0 150.3 144.0 710 81.7 113.1 113.1 122.2 150.7 145.0 720 83.1 113.1 113.1 123.3 151.1 146.0 730 84.5 113.1 113.1 124.4 151.5 146.9 740 85.8 113.1 113.1 125.5 151.9 147.9 750 87.1 113.1 113.1 126.6 152.2 148.8 760 88.4 113.1 113.1 127.6 152.6 149.7 G-33

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE G4. Bounding Quad Cities 1&2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6 and G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE.: HEAD VESSEL BELTLINE PRESSURE CURVEAA CURVEA - CURVE A CURVE B -CURVE B CURVE B (PSIG) (0F - (°F (0F- (0F ; (°F) (F) 770 89.6 113.6 113.2 128.6 153.0 150.6 780 90.8 114.3 114.6 129.6 153.4 151.5 790 92.0 115.1 116.0 130.6 153.8 152.4 800 93.2 115.9 117.3 131.5 154.1 153.2 810 94.3 116.6 118.6 132.5 154.5 154.1 820 95.4 117.4 119.9 133.4 154.9 154.9 830 96.5 118.1 121.1 134.3 155.2 155.7 840 97.5 118.8 122.3 135.2 155.6 156.5 850 98.6 119.5 123.4 136.0 155.9 157.3 860 99.6 120.2 124.6 136.9 156.3 158.1 870 100.6 120.9 125.7 137.7 156.6 158.8 880 101.5 121.6 126.8 138.6 157.0 159.6 890 102.5 122.3 127.9 139.4 157.3 160.3 900 103.4 122.9 128.9 140.2 157.7 161.1 910 104.4 123.6 129.9 141.0 158.0 161.8 920 105.3 124.2 130.9 141.7 158.4 162.5 930 106.1 124.9 131.9 142.5 158.7 163.2 940 107.0 125.5 132.9 143.3 159.0 163.9 950 107.9 126.1 133.8 144.0 159.4 164.6 960 108.7 126.7 134.8 144.7 159.7 165.2 970 109.6 127.3 135.7 145.5 160.0 165.9 980 110.4 127.9 136.6 146.2 160.4 166.5 990 111.2 128.5 137.4 146.9 160.7 167.2 1000 112.0 129.1 138.3 147.6 161.0 167.8 1010 112.7 129.7 139.1 148.2 161.3 168.5 1020 113.5 130.2 140.0 148.9 161.6 169.1 1030 114.3 130.8 140.8 149.6 162.0 169.7 G-34

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-4. Bounding Quad Cities 1&2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G4, G-5, G-6 and G-8

• BOTTOM ;UPPER '- i54EFPY BOTTOM UPPER 54 EFPY HEAD: VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A .:CURVE A ':CURVE A -' CURVE B:' CURVE B- CURVE B."

(PSIG) (°F) (°F)' .(OF) ' (OF) (OF) (0F 1040 115.0 131.4 141.6 150.2 162.3 170.3 1050 115.7 131.9 142.4 150.9 162.6 170.9 1060 116.4 132.5 143.2 151.5 162.9 171.5 1070 117.2 133.0 144.0 152.1 163.2 172.1 1080 117.9 133.5 144.7 152.8 163.5 172.7 1090 118.6 134.1 145.5 153.4 163.8 173.2 1100 119.2 134.6 146.2 154.0 164.1 173.8.

1105 119.6 134.8 146.6 154.3 164.3 174.1 1110 119.9 135.1 146.9 154.6 164.4 174.3 1120 120.6 135.6 147.6 155.2 164.7 174.9 1130 121.2 136.1 148.4 155.8 165.0 175.4 1140 121.9 136.6 149.1 156.3 165.3 176.0 1150 122.5 137.1 149.7 156.9 165.6 176.5 1160 123.1 137.6 150.4 157.5 165.9 177.0 1170 123.8 138.1 151.1 158.0 166.2 177.6 1180 124.4 138.6 151.7 158.6 166.5 178.1 1190 125.0 139.1 152.4 159.1 166.7 178.6 1200 125.6 139.5 153.0 159.7 167.0 179.1 1210 126.2 140.0 153.7 160.2 167.3 179.6 1220 126.8 140.5 154.3 160.8 167.6 180.1 1230 127.3 140.9 154.9 161.3 167.9 180.6 1240 127.9 141.4 155.5 161.8 168.2 181.1 1250 128.5 141.8 156.1 162.3 168.4 181.6 1260 129.0 142.3 156.7 162.8 168.7 182.1 1270 129.6 142.7 157.3 163.3 169.0 182.5 1280 130.1 143.2 157.9 163.8 169.2 183.0 1290 130.7 143.6 158.5 164.3 169.5 183.5 G-35

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G4. Bounding Quad Cities 1&2 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B &C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6 and G-8 BOTTOM UPPER 54 EFPY BOTTOM '"' UPPER ' -54'EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A -CURVE A CURVEA- CURVE B ':CURVE B CURVE B (PSIG) (OF) (°F) (OF): (FT) (°F) ' -(°F) 1300 131.2 144.0 159.0 164.8 169.8 183.9 1310 131.7 144.5 159.6 165.3 170.1 184.4 1320 132.3 144.9 160.2 165.8 170.3 184.8 1330 132.8 145.3 160.7 166.2 170.6 185.3 1340 133.3 145.7 161.3 166.7 170.8 185.7 1350 133.8 146.1 161.8 167.2 171.1 186.2 1360 134.3 146.6 162.3 167.6 171.4 186.6 1370 134.8 147.0 162.9 168.1 171.6 187.0 1380 135.3 147.4 163.4 168.5 171.9 187.5 1390 135.8 147.8 163.9 169.0 172.1 187.9 1400 136.3 148.2 164.4 169.4 172.4 188.3 G-36

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 *F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT ::HEAD BELTLINE AT BELTLINE AT

-54 EFPY 54 EFPY -54 EFPY

.,PRESSURE CURVE A CURVEA - CURVE B CURVE B CURVE C (PSIG) - ((F) -(aF)F) ((FO (OF) 0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 88.0 70 68.0 83.1 68.0 83.1 95.2 80 68.0 83.1 68.0 83.1 101.2 90 68.0 83.1 68.0 83.1 106.3 100 68.0 83.1 .68.0 83.1 110.8 110 68.0 83.1 68.0 83.1 114.9 120 68.0 83.1 68.0 83.1 118.7 130 68.0 83.1 68.0 83.1 122.2 140 68.0 83.1 68.0 85.4 125.4 150 68.0 83.1 68.0 88.2 128.2 160 68.0 83.1 68.0 90.9 130.9 170 68.0 83.1 68.0 93.5 133.5 180 68.0 83.1 68.0 95.9 135.9 190 68.0 83.1 68.0 98.2 138.2 200 68.0 83.1 68.0 100.3 140.3 210 68.0 83.1 68.0 102.3 142.3 G-37

GE Nuclear Energy GE-NE-0000-0002-960O-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 *F/hr for Curves B & C and 20 *F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD - BELTLINE AT -HEAD. BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY -54 EFPY

£PRESSURE- CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (0 F)° (OF (- -F) (OF) (OF):.-'

220 68.0 83.1 -68.0 104.3 144.3 230 68.0 83.1 68.0 106.1 146.1 240 68.0 83.1 68.0 107.9 147.9 250 68.0 83.1 68.0 109.6 149.6 260 68.0 83.1 68.0 111.2 151.2 270 68.0 83.1 68.0 112.8 152.8 280 68.0 83.1 68.0 114.3 154.3 290 68.0 83.1 68.0 115.8 155.8 300 68.0 83.1 68.0 117.2 157.2 310 68.0 83.1 68.0 118.5 158.5 312.5 68.0 83.1 68.0 118.9 158.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.1 143.1 183.1 G-38

GE Nuclear Energy GE-N E-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B &C and 20 °Flhr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTOM UPPER RPV & UPPER RPV &

--HEAD BELTLINE AT. . HEAD BELTLINE AT BELTLINE AT

-54 EFPY -54 EFPY. -54 EFPY PRESSURE CURVE A CURVEA - CURVE B CURVE B CURVE CVE (PSIG) (OF) - ( -F) (°F) - (°F) :

430 68.0 113.1 71.4 143.1 183.1 440 68.0 113.1 74.4 143.1 183.1 450 68.0 113.1 77.3 143.1 183.1 460 68.0 113.1 80.0 143.1 183.1 470 68.0 113.1 82.6 143.1 183.1 480 68.0 113.1 85.1 143.1 183.1 490 68.0 113.1 87.4 143.1 183.1 500 68.0 113.1 89.6 143.1 183.1 510 68.0 113.1 91.8 143.1 183.1 520 68.0 113.1 93.8 143.1 183.1 530 68.0 113.1 95.8 143.1 183.1 540 68.0 113.1 97.7 143.1 183.1 550 68.0 113.1 99.5 143.1 183.1 560 68.0 113.1 101.3 143.4 183.4 570 68.0 113.1 103.0 144.1 184.1 580 68.0 113.1 104.6 144.9 184.9 590 68.0 113.1 106.2 145.6 185.6 600 68.0 113.1 107.8 146.1 186.1 610 68.0 113.11 109.3 146.6 186.6 620 68.0 113.1 110.7 147.0 187.0 630 68.6 113.1 112.1 147.4 187.4 640 70.5 113.1 113.5 147.8 187.8 650 72.2 113.1 114.8 148.2 188.2 G-39

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT.:': HEAD BELTLINE AT BETLINE AT

'54 EFPY 54 EFPY -54 EFPY

,PRESSURE CURVE A CURVEA CURVE B CURVE B CURVE C (PSIG) - (° - (°F: (°F) ( (°F) 660 73.9 113.1 116.1 148.7 188.7 670 75.6 113.1 117.4 149.1 189.1 680 77.2 113.1 118.7 149.5 189.5 690 78.7 113.1 119.9 149.9 189.9 700 80.2 113.1 121.0 150.3 190.3 710 81.7 113.1 122.2 150.7 190.7 720 83.1 113.1 123.3 151.1 191.1 730 84.5 113.1 124.4 151.5 191.5 740 85.8 113.1 125.5 151.9 191.9 750 87.1 113.1 126.6 152.2 192.2 760 88.4 113.1 127.6 152.6 192.6 770 89.6 113.6 128.6 153.0 193.0 780 90.8 114.6 129.6 153.4 193.4 790 92.0 116.0 130.6 153.8 193.8 800 93.2 117.3 131.5 154.1 194.1 810 94.3 118.6 132.5 154.5 194.5 820 95.4 119.9 133.4 154.9 194.9 830 96.5 121.1 134.3 155.7 195.7 840 97.5 122.3 1352 156.5 196.5 850 98.6 123.4 136.0 157.3 197.3 860 99.6 124.6 136.9 158.1 198.1 870 100.6 125.7 137.7 158.8 198.8 880 101.5 126.8 138.6 159.6 199.6 G-40

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV &, BOTTOM:- UPPER RPV & UPPER RPV &'

HEAD BELTLINE AT HEAD BELTLINEAT BELTLINE AT 54EFPY 54 EFPY- 54 EFPY

.'PRESSURE CURVE A CURVE A - CURVE B -CURVEBB CURVE C:

(PSIG), (OF) (IF) (,.

.(OF) (IF (0F) 890 102.5 127.9 139.4 160.3 200.3 900 103.4 128.9 140.2 161.1 201.1 910 104.4 129.9 141.0 161.8 201.8 920 105.3 130.9 141.7 162.5 202.5 930 106.1 131.9 142.5 163.2 203.2 940 107.0 132.9 143.3 163.9 203.9 950 107.9 133.8 144.0 164.6 204.6 960 108.7 134.8 144.7 165.2 205.2 970 109.6 135.7 145.5 165.9 205.9 980 110.4 136.6 146.2 166.5 206.5 990 111.2 137.4 146.9 167.2 207.2 1000 112.0 138.3 147.6 167.8 207.8 1010 112.7 139.1 148.2 168.5 208.5 1020 113.5 140.0 148.9 169.1 209.1 1030 114.3 140.8 149.6 169.7 209.7 1040 115.0 141.6 150.2 170.3 210.3 1050 115.7 142.4 150.9 170.9 210.9 1060 116.4 143.2 151.5 171.5 211.5 1070 117.2 144.0 152.1 172.1 212.1 1080 117.9 144.7 152.8 172.7 212.7 1090 118.6 145.5 153.4 173.2 213.2 1100 119.2 146.2 154.0 173.8 213.8 1105 119.6 146.6 154.3 174.1 214.1 G-41

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 ¶F/hr for Curves B & C and 20 0F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINEAT.. HEAD BELTLINE AT:: -BELTLINE AT 54 EFPY - 54 EFPY - 54 EFPY

,PRESSURE CURVEA .CURVE A CURVE B CURVE B CURVE C (PSIG) ( 0) 0 (F - (° (°F) (°F) 1110 119.9 146.9 154.6 174.3 214.3 1120 120.6 147.6 155.2 174.9 214.9 1130 121.2 148.4 155.8 175.4 215.4 1140 121.9 149.1 156.3 176.0 216.0 1150 122.5 149.7 156.9 176.5 216.5 1160 123.1 150.4. 157.5 177.0 217.0 1170 123.8 151.1 158.0 177.6 217.6 1180 124.4 151.7 158.6 178.1 218.1 1190 125.0 152.4 159.1 178.6 218.6 1200 125.6 153.0 159.7 179.1 219.1 1210 126.2 153.7 160.2 179.6 219.6 1220 126.8 154.3 160.8 180.1 220.1 1230 127.3 154.9 161.3 180.6 220.6 1240 127.9 155.5 161.8 181.1 221.1 1250 128.5 156.1 162.3 181.6 221.6 1260 129.0 156.7 162.8 182.1 222.1 1270 129.6 157.3 163.3 182.5 222.5 1280 130.1 157.9 163.8 183.0 223.0 1290 130.7 158.5 164.3 183.5 223.5 1300 131.2 159.0 164.8 183.9 223.9 1310 131.7 159.6 165.3 184.4 224.4 1320 132.3 160.2 165.8 184.8 224.8 1330 132.8 160.7 166.2 185.3 225.3 G-42

GE Nuclear Energy GE-NE-0000-0002-9600-03R2a Non-Proprietary Version TABLE G-5. Bounding Quad Cities 1&2 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 *FIhrfor Curves B & C and 20 OF/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & -:UPPERRPV &

HEAD: BELTLINEAT :.-HEAD BELTLINE AT BELTLINE AT 54 EFPY. . - EFPY

- -54 '54 EFPY

'PRESSURE CURVEA CURVE A :.:-:. CURVE B CURVEB.: B CURVE C

• t (PSIG) - .-:. (on -O-:-)on -I: (OF 1340 133.3 161.3 166.7 185.7 225.7 1350 133.8 161.8 167.2 186.2 226.2 1360 134.3 162.3 167.6 186.6 226.6 1370 134.8 162.9 168.1 187.0 227.0 1380 135.3 163.4 168.5 187.5 227.5 1390 135.8 163.9 169.0 187.9 227.9 1400 136.3 164.4 169.4 188.3 228.3 G-43