ML050870587

Enclosure 7, NRC-05-0011, GE Report NEDC-33133, Revision 0, Pressure-Temperature Curves for DTE Energy Fermi Unit 2 February 2005

ML050870587
Person / Time
Site:	Fermi
Issue date:	02/28/2005
From:	Branlund B, Frew B, Tilly L General Electric Nuclear Energy Owners Group
To:	Office of Nuclear Reactor Regulation
References
DRF-0000-0030-4057, FOIA/PA-2010-0209, NEDO-33133, NRC-05-0011 NEDC-33133, Rev 0
Download: ML050870587 (191)

Text

ENCLOSURE 7 TO NRC-05-0011 REQUEST TO REVISE TS 3.4.10 REACTOR COOLANT SYSTEM PRESSURE & TEMPERATURE LIMITS GE REPORT NEDC-33133, REVISION 0 "PRESSURE-TEMPERATURE CURVES FOR DTE ENERGY FERMI UNIT 2" FEBRUARY 2005 NON-PROPRIETARY VERSION

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GE Nuclear Energy Engineering and Technology NEDO-33133 General Electric Company DRF 0000-0030-4057 6705 Vallecitos Road Revision 0 Sunol, CA 94586 Class I February 2005 Non-ProprietaryVersion Pressure-Temperature Curves

- For DTE Energy.

Fermi Unit 2 Prepared by: Ll riffy L.J. Tilly, Senior Engineer Structural Analysis & Hardware Design Verified by: CB(D 'Frew B.D. Frew, Principal Engineer Structural Analysis & Hardware Design Approved by: (13 (B ranfund B.J. Branlund, Principal Engineer Structural Analysis & Hardware Design

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version REPORT REVISION STATUS l1 Revision I Purpose . I I 0 l Initial Issue

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GE Nuclear Energy NEDO-33133 GE Nucl ear EnergyNEO313 Non-Proprietary Version IMPORTANT NOTICE This is a non-proprietary version of the document NEDC-33133P, 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 fl )).

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 Detroit Edison and GE, Fermi 2 Energy Center Extended Power Uprate Project, WIN #8 PIT Curves, effective 5/18104, 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 Detroit Edison, 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, 2005

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GE Nuclear Energy NEDO) 33133 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.

Fermi 2 is currently licensed to P-T curves for 32 EFPY [1]; the P-T curves in this report represent 24 and 32 effective full power years (EFPY), where 32 EFPY represents the end of the 40 year license, and 24 EFPY is provided as an intermediate point between the current EFPY and 32 EFPY. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the use of Kc from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress.

This report incorporates a fluence [4] calculated in accordance with the GE Ucensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190. This fluence was based upon a projection for the planned implementation of power uprate that provides additional conservatism.

The latest information from the BWRVIP Integrated Surveillance Program that is applicable to Fermi Unit 2 has been utilized.

The P-T curves presented in this report reflect changes from those currently licensed [1].

These P-T curves have been generated to incorporate a revised fluence [4].

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.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 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 beitline region (Region B)

• Upper vessel (Regions A & B)

• Lower vessel (Regions B &C)

For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 1000 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 200F/hr or less must be maintained at all times (see Appendix C for additional guidance).

The P-T curves apply for both heatup and cooldown and for both the 1/4T and 314T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, KI,, at 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 24 and 32 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 24 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

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GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE OF CONTENTS

1.0 INTRODUCTION

I 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITLAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 16 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 21

5.0 CONCLUSION

S AND RECOMMENDATIONS 56

6.0 REFERENCES

75

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GE Nuclear Energy NEDO-331 33 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 UPPER SHELF ENERGY (USE)

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

APPENDIX I FRACTURE MECHANICS EVALUATION FOR FLAW INDICATION 124 CONTAINED IN RPV LOWER-INTERMEDIATE SHELL VERTICAL WELD 15-308B

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE FERMI 2 RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2: CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 3i FIGURE 4-3: FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 41 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE Al 120 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [20 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 24 EFPY [20 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 61 FIGURE 5-4: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE Al UP TO 32 EFPY 120 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-5: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 24 EFPY [20 1FIHR OR LESS COOLANT HEATUP/COOLDONWrN] 63 FIGURE 5-6: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 32 EFPY [20 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64 0 °F/HR FIGURE 5-7: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [10 OR LESS COOLANT HEATUP/COOLDOWN1 65 FIGURE 5-8: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 66 FIGURE 5-9: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 24 EFPY 1100-F/HR OR LESS COOLANT HEATUP/COOLDOWN] 67 FIGURE 5-10: BELTLTNE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 32 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 68 FIGURE 5-11: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 24 EFPY

[100 0F/HR OR LESS COOLANT HEATUP/COOLDOWN] 69 FIGURE 5-12: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 32 EFPY

[100 0 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 70 FIGURE 5-13: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 24 EFPY [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 71 FIGURE 5-14: COMPOSITE CORE CRITICAL P-T CURVES [CURVE C] UP TO 32 EFPY [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 72

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE OF FIGURES, Continued FIGURE 5-15: LIMITING P-T CURVES ICURVES A, B, AND C] UP TO 24 EFPY 120°F/HR OR LESS COOLANT HEATUP/COOLDOWN FOR CURVE A AND 1000F/HR OR LESS COOLANT HEATUP/COOLDONW'N FOR CURVES B AND C] FIGURE 5-16: LIMITING P-T CURVES

[CURVES A, B, AND C] UP TO 32 EFPY 120 °F/HR OR LESS COOLANT HEATUP/COOLDOWN FOR CURVE A AND I000F/HR OR LESS COOLANT HEATUPICOOLDOWN FOR CURVES B AND Cl 73 FIGURE 5-16: LIMITING P-T CURVES [CURVES A, B, AND Cl UP TO 32 EFPY 120 0FiHR OR LESS COOLANT HEATUP/COOLDOWN FOR CURVE A AND I00F/HR OR LESS COOLANT HEATUP/COOLDOWN FOR CURVES B AND C] 74

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE OF TABLES TABLE 4-1: RTNDT VALUES FOR FERMI 2 PLATE AND FLANGE MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR FERMI 2 NOZZLE MATERIALS 12 TABLE 4-3: RTNDT VALUES FOR FERMI 2 WELD MATERIALS 13 TABLE 4-4: RTNDT VALUES FOR FERMI 2 APPURTENANCE AND BOLTING MATERIALS 15 TABLE 4-5: FERMI 2 BELTLINE ART VALUES (24 EFPY) 19 TABLE 4-6: FERMI 2 BELTLINE ART VALUES (32 EFPY) 20 TABLE 4-7:

SUMMARY

OF THE IOCFR50 APPENDIX G REQUIREMENTS 23 TABLE 4-8: APPLICABLE BWR/4 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 25 TABLE 4-9: APPLICABLE BWR/4 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 25 TABLE 4-10: PRESSURE TEST CRD PENETRATION K1 AND (T - RTNDT) AS A FUNCTION OF PRESSURE 29 TABLE 4-11: CORE NOT CRITICAL CRD PENETRATION K, AND (T - RTNDT) AS A FUNCTION OF PRESSURE 32 TABLE 4-12: PRESSURE TEST FEEDWATER NOZZLE K1 AND (T - RTNT) AS A FUNCTION OF PRESSURE 37 TABLE 4-13: CORE NOT CRITICAL FEEDWATER NOZZLE KEAND (T - RTNDT) AS A FUNCTION OF PRESSURE 45 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 58

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GE Nuclear Energy NEDO-33133 Non-Proprietary Version

1.0 INTRODUCTION

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

Complete P-T curves were developed for 24 and 32 effective full power years (EFPY),

where 32 EFPY represents the end of the 40-year license, and 24 EFPY is provided as an intermediate point between the current EFPY and 32 EFPY. The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B. This report incorporates a fluence calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14], and is in compliance with Regulatory Guide 1.190. This fluence was based upon a projection for the planned implementation of power uprate that provides additional conservatism. The latest information from the BWRVIP Integrated Surveillance Program that is applicable to Fermi Unit 2 has been utilized.

The P-T curves presented in this report reflect changes from those currently licensed [1].

These P-T curves have been generated to incorporate a revised fluence [4].

The methodology used to generate the P-T curves in this report is presented in Section 4.3. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the use of Kc from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. P-T curves are developed using geometry of the RPV shells and discontinuities, the initial RTNDT of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.

The initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. The Charpy energy data used to determine the initial RTNDT values are tabulated from the Certified Material Test Report (CMTRs). The data and methodology used to determine initial RTNDT are documented in Section 4.1.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Adjusted Reference Temperature (ART) is the reference temperature when including irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 [71 provides the methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 24 and 32 EFPY are included in Section 4.2. The peak ID fluence values of 7.13e17 n/cm 2 (24 EFPY) and 9.68e17 n/cm 2 (32 EFPY) used in this report are discussed in Section 4.2.1.2. Beltline chemistry values are discussed in Section 4.2.1.1.

Appendix A includes comprehensive documentation of the RPV discontinuities considered in this evaluation. This appendix also documents the non-beltline discontinuity curves that are used to protect each discontinuity.

Guidelines and requirements for operating and temperature monitoring are included in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles requiring fracture toughness evaluation are either included in the development of the P-T curves or are outside the beltline region.

Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Appendix G contains an evaluation of the vessel wall thickness discontinuities in the beltline and bottom head regions. Appendix H provides a core-not-critical calculation for the bottom head (CRD penetration). Upjohn welds occur in various welds in the Fermi Unit 2 vessel; these welds are known to contain flaws. The limiting flaw, which exists in the beitline region, has been evaluated and is discussed in Appendix I of this report.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 2.0 SCOPE OF THE ANALYSIS A detailed description of the P-T curve bases is included in Section 4.3. The 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda was used in this evaluation. The P-T curve methodology includes the following: 1) the use of K1c from Figure A-4200-1 of Appendix A to determine T-RTNDT, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. 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].

The beltline region in the Fermi Unit 2 vessel includes a thickness discontinuity between the lower and lower-intermediate shells. This discontinuity is noted in Appendix A and evaluated in Appendix G. 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 Fermi 2 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8]. In addition, there are thickness discontinuities in the bottom head, which are also noted in Appendix A and evaluated in Appendix G.

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 GE Nuclear Energy NEDO-331 33 Non-Proprietary Version methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles requiring fracture toughness evaluation are either included in the development of the P-T curves or are outside the beltline region. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Appendix G contains an evaluation of the vessel wall thickness discontinuities in the beltline and bottom head regions.

Appendix H provides a core-not-critical calculation for the bottom head (CRD penetration). Upjohn welds occur in various welds in the Fermi Unit 2 vessel; these welds are known to contain flaws. The limiting flaw, which exists in the beltline region, has been evaluated and is discussed in Appendix I of this report.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:

The hydrostatic pressure test will be conducted at a maximum pressure of 1055 psig [13].

The shutdown margin, provided in the Definitions Section of the Fermi 2 Technical Specification [13], is calculated for a water temperature of 680F.

GE Nuclear Energy NEDO-33133 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 applicable ASME Code for the Fermi 2 RPV is 1968 Edition with Summer 1969 Addenda. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section 1II,Subsection NB-2300 and are summarized as follows:

a. Test specimens shall be longitudinally oriented CVN specimens.

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

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

The current requirements used to establish an initial RTNDT value are significantly different. For plants constructed according to the ASME Code after Summer 1972, the requirements per the ASME Code Section 1I1,Subsection NB-2300 are as follows:

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

b. RTNDT is defined as the higher of the dropweight NDT or 600F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion are met.

c. Bolt-up in preparation for a pressure test or normal operation shall be performed at or above the highest RTNDT of the materials in the closure flange region or lowest service temperature (LST) of the bolting material, whichever is greater.

10CFR50 Appendix G [8) states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses GE Nuclear Energy NEDO-331 33 Non-Proprietary Version must be supplemented in an approved manner. Additional details are contained in the Fermi 2 UFSAR, Sections 3.1.2.4.2 and 5.2.4. GE developed methods for analytically converting fracture toughness data for vessels constructed before 1972 to comply with current requirements. These methods were developed from data in WRC Bulletin 217 [9] and from data collected to respond to NRC questions on FSAR submittals in the late 1970s. In 1994, these methods of estimating RTNDT were submitted for generic approval by the BWR Owners' Group [10], and approved by the NRC for generic use [11].

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

To establish the initial RTNDT temperatures for the Fermi 2 vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDT calculations for vessel plate, forging, and for bolting material LST are summarized in the remainder of this section. The initial RTNDT for all materials remain unchanged from values previously reported for Fermi 2 with one exception. Lower-intermediate shell plate heat C4564-1 was previously reported to have an initial RTNDT of -12 0F. As demonstrated in Table 4-1 and in the calculation shown below, using the methods mentioned above and documented in [10] and [11], the initial RTNDT for heat C4564-1 is determined to be -1 00F.

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 [121). For Fermi 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 Fermi 2 beltline plate heat C4564-1 in the lower-intermediate shell course; the lowest Charpy energy and test temperature from the CMTRs is 45 ft-lb at 100F. The estimated 50 ft-lb longitudinal test temperature is:

TsOL = 10 0F + [ (50 - 45) ft-lb 20F/ft-lb ] = 20'F GE Nuclear Energy NEDO-33133 Non-Proprietary Version The transition from longitudinal data to transverse data is made by adding 30IF to the 50 ft-lb longitudinal test temperature; thus, for this case above, T50T = '20'F + 30'F = 50'F.

The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5oT- 60'F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is -20 0F. Thus, the initial RTNDT for plate heat C4564-1 is -10OF.

For the Fermi 2 beltline weld heat 12008 Linde 1092 with flux lot 3833 (contained in the lower shell course), the CVN results are used to calculate the initial 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 12008 Linde 1092 has a lowest Charpy energy of 47 ft-lb at 100F as recorded in weld qualification records. Therefore, T5oT= 10 0F+[(50-47)* 20F/ft-lb] = 160F The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T50T- 600F). For Fermi 2, the dropweight testing to establish NDT was not available and -50 0F was assumed. The value of (T50T - 600 F) in this example is -440F; therefore, the initial RTNDT was -440F.

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 limiting heat in the feedwater nozzle at Fermi 2 (Heat AV3504 98-9202), the NDT is -100F and the lowest CVN data is 34 ft-lb at 100F. The corresponding value of (T50T - 600F) is:

(TSOT - 60°F) = {[10 + (50- 34) ft-lb

• 20F/ft-lb] + 300F) - 600F = 120F.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (ToT- 600F), which is 120F.

In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNDT. For the bottom head dollar plate heat of Fermi 2 (Heat C4504-2), the NDT is 100F and the lowest CVN data was 40 ft-lb at 400F. The corresponding value of (T50T -

60°F) was:

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

• 20F/ft-lb J+ 300F }-600F = 300F.

Therefore, the initial RTNDT was 30'F.

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. If the required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section 1II,Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 600F is the LST for the bolting materials. All of the available reported Charpy data for the Fermi 2 closure studs met the 45 ft-lb requirements at 100F. However, MLE data was not reported and information for all stud materials was not available. Therefore, the limiting LST for the bolting material is 700F, The highest RTNDT in the closure flange region is 120F, for the upper shell. Thus, the higher of the LST and the RTNDT +60°F is 72 0F, the bolt-up limit in the closure flange region.

The initial RTNDT values for the Fermi 2 reactor vessel (refer to Figure 4-1 for the Fermi 2 Schematic) materials are listed in Tables 4-1, 4-2, 4-3 and 4-4. This tabulation includes beltline, closure flange, feedwater nozzle, and bottom head materials that are considered in generating the P-T curves. The values presented in these tables and used to determine the initial RTNDT were obtained from the Fermi 2 vessel CMTRs [12].

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

/ \ TOP HEAD TOP HEAD FLANGE SHELL FLANGE SHELL #4 0 0 S HELL #3 TOP OF SHELL #2 ACTIVE FUEL (TAF) 366.3- WELDS

. / IRTHWELD BOTTOM OF SHELL#1 ACTIVE FUEL (BAF) 216.3- (

BOTTOM HEAD SUPPORT SKIRT Notes: (1) Refer to Tables 4-1, 4-2, 4-3 and 44 for reactor vessel components and their heat identifications.

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

Figure 4-1: Schematic of the Fermi 2 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table 4-1: RTNDT Values for Fermi 2 Plate and Flange Materials

. Test Drop Component Heat Temp Charpy Energy (Tswr60) Weight RTNDT CopFen)Tm (ft~b) (OF) NDT (OF)

__ ___ __ ___ _(OF ) _ _ _ _

Top Head & Flange Shell Flange G3701 AWH 113 2V-708 10 89 89 90 -20 10 10 Top Head Flange G3702 ACR 108 10 196 191 101 -20 10 10 Top Head Dollar C3732 C5445-1 10 79 86 79 -20 -10 -10 Top Head Lower Torus Plates  ;

G3731-1 C5445-2 10 95 91 97 -20 -10 -10 G3731-2 C5445-2 10 97 96 101 -20 -10 -10 Top Head Upper Torus Plates G3730 C5445-1 10 107 85 102 -20 -10 -10 Shell Courses Upper Shell Plates G3703-1 C4568-1 40 68 65 56 10 -10 10 G3703-2 C4564-2 40 64 49 54 12 -10 12 G3703-3 C4560-2 10 53 63 52 -20 -10 -10 G3703-4 C4554-2 40 74 75 87 10 -10 10 Upper Intermediate Plates G3704-1 C4574-1 10 70 68 66 -20 -10 -10 G3704-2 C4578-2 10 59 45 52 -10 -10 -10 G3704-3 C4578-1 10 44 56 68 -8 -10 -8 Lower-intermediate Plates G3703-5 C4564-1 10 60 45 59 -10 -20 -10 G3705-1 B8614-1 10 62 64 56 -20 -20 -20 G3705-2 C4574-2 10 48 49 60 -16 -30 -16 G3705-3 C4568-2 10 46 67 63 -12 -30 -12 Lower Shell Plates G3706-1 C4540-2 10 64 76 74 -20 -10 -10 G3706-2 C4560-1 10 85 79 99 -20 -10 -10 G3706-3 C4554-1 10 59 65 68 -20 -10 -10 Bottom Head Bottom Head Dollar G3708 C3424-1 10 41 48 57 -2 -10 -2 Bottom Head Upper Torus Plates G3711-1 C4526-1 -40 57 60 55 -70 -10 -10 G3712-1 C4504-3 -40 70 64 56 -70 -10 -10 Bottom Head Lower Torus Plates C3709-1 C5050-2 10 58 70 83 -20 -10 -10 C3710-1 C4504-1 10 74 70 74 -20 -10 -10 C3710-2 C4504-2 40 40 48 42 30 10 30 NOTE: These are minimum Charpy values.

- 11 -

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

- Table 4-2: RTNDT Values for Fermi 2 Nozzle Materials Drop Component . Test Charpy Energy .(TsorGO) Weight RTN.r Heat I Flux I Lot ('F) _ (_F)

_4b) hhDT ('F(

Recirculation Outlet Noze G3717-1 AJF 181 10 77 96 59 -20 -20 .20 G3717-2 AJF 193 10 91.5 89 85 -20 -30 -20 Recirculation Inlet Nozzle G3718-1 AV35059A-9239 10 54 45 38 4 0 4 G3718-2 AV3505 9A-9240 10 34 32 36 16 -10 16 G3718-3 AV3502 SA-9365 10 35 32 36 16 20 20 G3718-5 AV3503 9A-9368 10 42 36 60 8 -30 8 G3718 6 AV3503 9A-9369 10 49 40 49 0 -30 0 G3718-7 AV3504 9A-9371 10 58 28 37 24 -40 24 G3718-8 AV3857 9D-9407 10 48 32 46 16 20 20 G3718-9 AV3857 9D-9406 10 47 64 58 -14 10 10 G3718-10 - AV3857 9D9408 10 82 46 55 -12 10 10 G5218-4 AV3934 9E-9011 10 47 60 88 -14 40 40 Steam Outlet Nozzle G3714-1 AV3496 9A-9234 10 66 36 85 8 10 10 G3714-2 AV3507 9A-9235 10 74 75 36 8 0 8 G3714-3 AV3510 9A-9236 10 36 34 32 16 10 16 G3714-4 AV3511 9A-9237 10 52 32 42 16 10 16 Feedwater Nozzle G3715-1 AV3508 9A-9228 10 82 91 58 -20 0 0 G3715-2 AV3508 9A-9229 10 68 58 50 -20 0 0 G3715-3 AV3508 9A-9230 10 62 67 76 -20 -30 -20 G3715-4 AV3509 9A-9232 10 60 42 64 -4 -10 -4 G3715-5 AV3509 9A-9231 10 38 54 50 4 0 4 G3715-6 AV3504 9B-9202 10 39 46 34 12 -10 12 Core Spray Nozzle G3720-1 AV2997 9A-9363 10 67 52 96 -20 -10 -10 G3720-2 AV2997 9A-9364 10 70 99 84 -20 .10 -10 Instrumentation Nozzle G3811-1 Q2014W969C-1 10 54 59 73 -20 10 10.

G3811-2 02Q14W 969C 10 54 59 73 -20 10 10 Top Head Vent Nozzle G3810 0206W 986C 10 92 95 88 -20 10 10 Jet Pump Nozzle G3719-1 EV.98068L-9211A 10 99 124 105 -20 -20 -20 G3719-2 EV-9806 8L-9211 B 10 99 124 105 -20 -20 -20 CRO HYD Return Nozzle G3716 - AV3138 8L-9104 10 42 40 48 0 10 10 Core AP Nozzle Alloy 600 G3738 NX9492 .. Q(2)

Replacement Instrument Nozzles G3806 2127273 10 36 43 30 20 40 40 G3806R 6397860 10 250 230 247 -20 40 40 High Pressure Leak Detector Nozzle G4546 10 (1)

Drain Nozzle G3739 0101VW738T 10 39 25 32 30 40 40 CRD Stub Tubes G3736-1 through -5 Alloy 600 (2)

(1) Information for this heat Is not available; the purchase specification requirements are used for evaluation of this component.

(2) Alloy 600 components do not require fracture toughness evaluation; see Appendix A for addrdonal information.

NOTE: These are minimum Charpy values.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table 4-3: RTNDT Values for Fermi 2 Weld Materials Drop Beltilne - Axial Component

-o-pHeat

__I__ _Flux_ _ Heat__

_I__

Heat or Lot _

Tes Temp (F)('F___

Charpy

(.F) Energy j

(Ter60)

JWeight NOT DTF RTpj (F1 (4

.ower Shell 13253 Linde 1092 2-307 A, B. C Lot 3833 10 79 79 82 -50 -50 12008 Linde 1092 Lot 3833 10 62 47 62 -44 .44 Lower-lntermecate Snell 33A277 Linde 124 15-308 A, B. C. D Lot3878 10 83 94 87 -50 -50 Beltline - Girth Lower-Intermediate Shell to Lower Shell 10137 Linde 0091 1-313 LdI3999 10 101 108 107 -50 -50 Non-Beftline - Axial Upper-Intermediate Shell 20291 & 12008 Linde 1092 2-308 A through C Lot3833 10 62 47 62 -44 , -44 HADHI 10 112 110 114 -50 -50

__EOAG 10 173 133 135 -50 -50 Upper Shell 1-308 A through D 34B009 Linde 124 Lot 3687 10 55 65 57 -50 .50 348009 Unde 124 Lot 3688 10 69 70 63 -50 .50 Bottom Head Upper Torus Menclonal Welds 14306 A throuch K HADH 10 112 110 114 .50 -50 Bottom Head Lower Torus Mendonal Welds 2-306 A through G LACH-2 10 125 119 119 -50 -50 HADH 10 112 110 114 -50 .50 Top Head Upper Torus Mendonal Welds 2319 A throuah E EOEJ 10 136 170 150 -50 - -50 Top Head Lower Torus Mendonal Welds 1419 A through H DOAJ 10 166 145 158 -50 -50 AOFJ 10 112 105 115 -50 -50 EOEJ 10 136 170 150 -50 .50 Non-Beitlne - Grth Top Head AssemrDy 3-319.4-319.5-319 33A277 Unde 0091 Lot 3977 10 111 106 113 -50 -50 90099 Linde 0091 Lot 3977 10 56 30 52 -10 .- 10 Shell Flange to Upper Shell 13-308 21935 Lnde 1092 Lot 3889 10 51 70 74 -50 -50 Upper Shell to Upper-Intermediate Shell 305424 Linde 1092 4-308-A Lot 3889 10 82 87 92 -50 -50 Jpper-lntermedlate Shel to Lower-Intermediate Shell 1P3571 Unde 1092 Lot 3958 4-308-B Tandem 10 79 68 64 -50 -50 1P3571 Unde 1092 Lot 3958 Single 10 40 46 46 -30 -30 Lower Shell to Bottom Head 9-307 90099 Linde 0091 Lot 3977 10 56 30 52 -10 -10 90136 Linde 0091 Lot 3995 10 110 109 107 -50 -50 Bottom Head Asserroty 1P2809 Linde 1092 3-30655406. 6306 Lot3854 10 102 102 103 -50 -50 Support Slur, to Bottom Head 4-309 21935 Linde 1092 Lot 3869 10 62 59 60 -50 _ 50 Note: These are minimum Charpy values.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table 4-3 : RTNDT Values for Fermi 2 Weld Materials, Continued

.Heat Test Drop Component or Temp Charpy Energy (Tswr,-6) Weight RTF.,T HeatlIFluxlILot f-t)() NDT

F F Nozzle Welds Recirculation Outlet LOBI 10 123 104 115 -50 -50 5-314 A B CBAI 10

_BAI 10 120 105 128 -50 -50 Recarculation Inlet 13-314 A through K LOEH 10 113 123 140 -50 . .50 13-3140 (Replacement) BBAI 10 97 100 77 450 . -50 LACH 10 125 119 119 -50 -50 HOGI 10 91 93 94 -50 . -50 IAGI 10 142 157 170 -50 . -50 Steam Outlet FAGt 10 135 121 136 -50 .50 8-316 Athrouoh D LACH 10 125 119 119 -50 . -50 Feecwater Nozzle FACI 10 4-316 A through F HOGI 10 91 93 94 -50 . -50 LOEH 10 113 123 140 -50 . -50 COFI 10 96 97 89 -50 . 50 Core Spray Nozzles IAGI 10 142 157 170 -50 -50 14-316 A &8B BBAI 10 97 100 77 -50 . -50 Top Head Instrumert Nozzle 14-318 A & B ABEA 10 BOIA 10 99 110 1i13 -50 . -50 Top Head Vent Nozzle ABEA 10 2-318 BOIA 10 99 110 113 -50 -50 Jet Pumnp Nozzle LACH 10 125 119 119 -50 -50 19-314 A & B LOEH 10 113 123 140 -50 . -50 CRD HYD Retum Nozzle 15-315 IAGI 10 142 157 170 -50 . -50 Gore AP Nozzle 9-315 Inconel 182 Instrument Nozzles 4-315 A through F Inconel 182 Drain Nozzle 17-315 CAFJ 10 85 101 108 -50 . -50 Stub Tubes 1-310 Inconel 182 .

Appurtenance Welds Slablizer Brackets ICJJ 10 121 120 128 .50 . -50 10-324 A through H DBIJ 10 129 117 122 -50 . -50 HOCJ 10 165 174 140 -50 -50

- GBCJ 10 126 143 121 -50 . -50 Stearn Dryer Hdc1Down Brackets to Top Head 1 _

10-319 KAHJ 10 108 116 107 40 -SO Basin Seal Skirt 6-324 A through D IBEJ 10 160 151 145 -50 . -50 LOAJ 10 152 125 104 -50 - -50 7-324 A through D; 8-324 HOKJ 10 110 177 154 -50 - -50 KACJ 10 102 81 108 .- 50 - -50 Thermocoupie Pacs 1-325 GBJJ 10 203 160 239 -50 - -50 2-325 COEJ 10 129 95 e1 -50 -50 3-325; 4-325; 6-325 FCJJ 10 180 224 171 450 . -50 BBJJ 10 107 102 53 -50 . -50 COCA 10 120 139 137 -50 . -50 HOKJ 10 110 .177 154 .50 . -50 FOIA 10 182 224 218 -50 . .50 7-325: 8-325 BOLH 10 159 138 123 -50 . .50 Top Head Lifting Lugs GBCJ 10 126 143 121 -50 -50 8-319 A through D DBIJ 10 129 117 122 .50 . -50 NOTE: These are minimum Charpy values.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table 4-4: RTNDT Values for Fermi 2 Appurtenance and Bolting Materials Test Drop Component Heat Temp Charpy Energy (T-I)eFmNTpF (TsoT-60) Weight RTHOT

'F)

• Misc Appurtenances:

Support Skirt Forging S8530 AHC 178 10 81 100 102 -20 30 30 Shroud Support Alloy 600 33726 580608-IX __. (1)

Stabilizer Brackets C-S-1 A4516-1 40 58 49 53 12 10 12 C62 C5313-2 10 57 45 52 -10 -30 -10 Guide Rod Brackets G3772 Stainless Steel (1)

Steam Dryer Support Lugs G3775 Stainless Steel (1)

Steam Dryer Hold Down Brackets G4871 C2588-2D 10 122 129 107 -20 - -20 D5591 C6195-4 10 83 69 61 -20 - -20 Core Spray Brackets G3774 Stainless Steel (1)

Basin Seal Skirt G3818 C2588-2B 10 (2)

G3819 22A459 _ 10 (2)

Surveillance Specimen Brackets G3776 Stainless Steel (1)

G3777 Stainless Steel (1)

Feedwatcr Sparger Brackets G3773 Stainless Steel (1)

Top Head Lifting Lugs 33732 _ 40 (2)

Test Min Lat Charpy Energy LST Component Heat Temp Exp (ft4b) (-F) rF) (mils)

Closure Studs 33778-1 14677 10 50 50 52 70 33778-2 67156 10 55 54 55 - 70 G3778-3 (3) 10 - - - - 70 Closure Nuts G3779-1 48192 10 58 59 54 - 70 G3779-2 (3) 10 - 70 Closure Washers Closure Washers 35252 (3) 10 - 70 Bushings 34853 (3) 10 - 70 (1) Information for this heat is not available; the purchase specification requirements are used for evaluation of this component.

(2) Alloy 600 and Stainless Steel components do not require fracture toughness evaluation; see Appendix A for additional infonnatior (3) Information for this component Is not available; ASME Code requirements are applied as defined In Section 4.1.2 of this report.

NOTE: These are minimum Charpy values.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE The adjusted reference temperature (ART) of the limiting. beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RGI.99) provides the methods for determining the ART. The RG1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beltline plates and welds was performed and is summarized in Tables 4-5 and 4-6 for 24 and 32 EFPY, respectively.

4.2.1 Regulatory Guide 1.99, Revision 2 (RGI.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(O.28-0.10 log I Margin = 2(oal2 + 2 C;A ) 0.5 CF = chemistry factor from Tables 1 or 2 of RG1.99 f = Y4T fluence /10 1 9 Margin = 2(C012 + C;,2) 0.5 is, standard deviation on initial RTNDT, which istaken to be 0F.

CFA standard deviation on ARTNDT, 280F for welds and 17'F for base material, except that a,, need not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT The margin term cr, has constant values of 170 F for plate and 280F for weld as defined in RG1.99. However, a,&need not be greater than 0.5 *ARTNDT. Since the GE/BWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in 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 weld materials.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version 4.2.1.1 Chemistry The vessel beltline chemistries were obtained from [13] and are consistent with all known available sources of data for the'beltline materials, including the Certified Material Test Reports (CMTR) [12], and the 1991 P-T curve report [1]. Chemistries for the surveillance materials evaluated in Tables 4-5 and 4-6 were obtained from the Integrated Surveillance Program [13]."

The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG1.99, to determine a chemistry factor (CF) per Paragraph 1.1 of RG1.99 for welds and plates, respectively.

For weld heat 13253, 12008, both the chemistry for the Fermi 2 vessel weld and the chemistry from the Integrated Surveillance Program (ISP) are presented.

Heat CE-2(WM), which has been determined to be Heat 13253, 12008, is the surveillance weld material as'defined by the Integrated Surveillance Program (ISP);

chemistry and adjusted CF information defined by this program were provided by [13].

For this material, an adjusted CF used in calculating the adjusted reference temperature for 24 and 32 EFPY was obtained by multiplying the ISP least-squares fit CF developed in accordance with RG1.99 as defined by BWRVIP-102 [5] by the ratio of the RG1.99 CF for the vessel weld chemistry to the RG1.99 CF for the ISP surveillance chemistry. This results in an adjusted CF of: 326.96 (224 /206.6) = 354.5.

4.2.1.2 Fluence The fluence used in this evaluation reflects implementation of power uprate for Cycle 11.

Delay of implementation of the power uprate will increase the conservatism for the beltline P-T curves. The peak fluence for the RPV inner surface, used for determination of the P-T curves, is 9.68e17 n/cm 2 for 32 EFPY. For 24 EFPY, the peak fluence for the RPV inner surface is 7.13e17 n/cm2. The basis for all fluence values used in this report is contained in [4]. Calculations for 1/4T fluence are performed in accordance with RG1.99 [7]. The fluence used in developing the P-T curves is conservatively based upon operation at 3430 MWt for 12.04 EFPY and 3952 M~M for 19.96 EFPY.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version The peak fluence for the elevation of the girth weld between the lower and lower-intermediate shell plates is also provided in [4]. This fluence is applied to this girth weld and all plates and welds in the lower shell. Axial fluence distribution factors of 0.64 and 0.65 are applied for the 32 and 24 EFPY fluences, respectively. The slight difference is due to the amount of time that Fermi Unit 2 will operate at the EPU power level.

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

Tables 4-5 and 4-6 list values of beltline ART for 24 and 32 EFPY, respectively.

Surveillance capsule material data is available from the Integrated Surveillance Program (ISP) to represent the Fermi Unit 2 vessel. These materials are included in the ART calculations provided in Tables 4-5 and 4-6, and in the determination of the limiting material that is represented in the beltline P-T curves.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table 4-5: Fermi 2 Beltline ART Values (24 EFPY)

Lower-Intermediate Shell Plates and Axial Wekds Thickness in hiches= 6.125 24 EFPY Peak I.D. fluence - 7.13E+17 nicm12 24 EFPY Peak 114T fluence = 4.94E+17 n/cm^2 24 EFPYPeak114Tfluence= 4.94E+17 nrcmA2 Lower Shell Plates and Axbal Welds & Lowerto Lower4ntermediate GIrth Weld Thickness in kiches- 7.125 24 EFPY Peak l.D. fluence = 4.66E+17 ncrn^2 Axial Distribution Factor at Elevation 24 EFPY Peak 1/4 T fluence = 3.04E+17 n/cm^2 or Girth Weld = 065 24 EFPY Peak 114T fluence = 3.04Et17 ri/cm`2 A4usted Inital 14 T 24 EFPY 24 EFPY 24 EFPY COMPONENT HEATORHEAT&LOT SCu %Ni CF CF (1) RTndt Fluence A RTnrlt a G Margin Shift ART

.I._._._____.*F nicm12 IF _ F F IF PLATES:

Lower Shell 03706-1 C4540-2 0.08 0.62 51 -10 3.04E+17 11 0 6 11 23 13 G3706-2 C4560-1 0.11 0.57 74 -10 3.04E+17 16 0 8 16 33 23 G3706-3 C4554-1 0.12 0.56 82 *10 3.04E+17 18 0 9 18 36 26 LowerF4ntermediate Shell G3703-5 C4564-1 0.09 0.55 58 -10 4.94E+17 17 0 8 17 3.4 24 G3705-1 B8614-1 0.12 0.61 83 -20 4.94E+17 24 0 12 24 48 28 G3705-2 C4574-2 0 10 0.55 65 -16 4.94E+17 19 0 9 19 38 22 G3705-3 C4568-2 0.12 0.61 83 -12 4.94E+17 24 0 12 24 48 36 WELDS:

Lower Shell Axial Tandem 13253,12008 2-307A B. C 1092Lot3833 0.26 087 224 -44 3.04E+17 50 0 25 50 99 55 Lower4-ntermediate Shell Axial 15-308A B. C D 33A277 124Lot3878 0.32 0.50 188.5 -50 4.94E+17 55 0 27 55 110 60 Lower to Lower-Intermediate Girth 1-313 10137 0091 Lot 3999 0.23 1.00 236 *50 3.04E+17 52 0 26 52 104 54 INTEGRATED SURVEILLANCE PROGRAM (21:

Plate (3) C4114-2 0.12 0.69 84 -12 4.94E+17 24 0 12 24 49 37 Weld (4) CE-2 (WMX13253.12008) 0.21 0.86 207 354 -44 3.04E+17 78 0 28 28 106 62

() Adputed CF calkulated per RGI 99 Poston Z1 as Shownhi Secton4.2.1.1 of elt report (2) Procedures definedhi BWRvIP-102 are appliedto determrire tie ART consileriog VieIntegrated Surveillance Program.

(3)TheISP plate Isnot tie identical heatrand is presented usingtie ISP chemistry andCFand appledtote lmiting Fermni 2 pate. which toHeatC456S-2. ISPindcatesthatC4554-1 is also a imhngpate.

however cue to the reduced Ileneat ie towr hel, ts mtenal is nrobrger Smitng.

(4)The ISPweld Is the identcal heat and I spresented sing the ISP chemrrstryandcausted CFwsithe vesselweld Inetal RTariandftence. 0,t spreserted ascalculated. bis 15mulpled by 0.5 lortVe Margin catuhtlon as deed hi RGI.99, Positon 2.1.

- 19 -

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table 4-6: Fermi 2 Beltline ART Values (32 EFPY)

Lower-IntermedIate Shell Plates and Axial Welds Thickness In hIches= 6 125 32 EFPY Peak I.D. fuence= 9.68E+17 ncm^2 32EFPYPeak114Tfluence= 6.70E+17 r/cm^2 32PY Peak 114T fluence = 6.70E+17 rncm^2 Lower Shell Plates and Axial Welds & Lower to Lower-Intermediate Girth Weld Thickness in inches= 7.125 32 EFPY Peak l.D. Iluence = 6.23E.17 n/cm^2 Axial Distritution Factor at Elevation 32 EFPY Peak 11 Tfuence= 4 06E.17 n/cmA2 ofGirthWeld 064 32PYPeakl114T fuence= 406E#17 n/crn2 Adjusted Initial 114T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OiR HEAT/LOT %Cu  %'N CF CF (1) RTndt Fluence a RTndt a, oe Margin . Shift ART

.I._._.__ __. F n/cm^2 -F -F F F PLATES:

Lower Shell G3706-1 C454,-2 0.08 0.62 51. -10 4.06E.17 13 0 7 13 27 17 G3706-2 C4560-1 0.11 0.57 74 .10 4.06E1 17 19 0 10 19 38 28 G3706-3 C4554-1 0.12 0.56 82 -10 4.06E417 21 0 11 21 43 33 Lower4rntermediate Shell G3703-5 04564-1 0.09 0.55 58 -10 6.70E+17 20 0 10 20 40 30 G3705-1 B8614-1 0.12 0.61 83 .20 6.70Et17 28 0 14 26 57 37 G370S-2 C4574-2 0.10 0.55 65 -16 6.70E+17 22 0 11 22 44 28 G370S-3 C4568-2 0.12 0.61 83 -12 6.70E.17 28 0 14 28 57 45 WELDS:

Lower Shell Axial Tandem 13253, 12008 2-307 A, B. C 1092 Lot 3833 0.26 0.87 224 -44 4.06E*17 59 0 28 56 115 71 Lower-Intermediate Shell Axial 15-308 A. B.C. C D 33A277, 124 Lot 3878 0.32 0.50 188.5 .50 6.70E+17 64 0 28 56 120 70 Lower to Lower4-ntermediate Girth 1-313 10137. 0091 Lot 3999 0.23 1.00 236 -50 4.06E-17 62 0 28 56 118 68 INTEGRATED SURVEILLANCE PROGRAM 12):

Plate (3) C4114-2 0.12 069 84 -12 6.70E*17 29 0 t4 29 57 45 Weld (4) CE-2 (WM)(13253.12008) 0.21 0.86 207 354 -44 4 06E*17 93 0 28 28 121 77 (1) Adjusted CF calculated per RG1.99 Posibon 2.1 asshDan InSection 4.2.1.1 of Its report (2) ProcedruresdefredinBWRVYPt102re appliedtodeterminetheARTconsiderrgtheintegratedSurveilancePrograom.

(3)The ISPplate I nottie k ertbcl heat andispresentedusinrtie ISPchemrstryandCFandapplied tie lting FemmI2 piate,whch i5 HeatC456882. ISP undcalesthatC4554-1 s alsoa limibng plate, however oue b the reduced fluence attne wershel. tis msateiial no longer Iminrig (4) The ISP weld is tie ioenrcal heat and is presented wsingtie ISP chemisly and adjusted CF with the vessel weld Initral RTr arndluenrce.a, is presented as calculated, but Is multplied by 0.5 for the Margin calculaton as defined kmRG1.99. Position 2.1.

GE Nuclear Energy NEDO-331 33 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 Xi [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 nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 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 (see Appendix C for additional guidance).

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup.

However, as a conservative simplification, the thermal gradient stress at the 1/4T location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4T location. This approach is conservative because irradiation effects cause the allowable toughness, KIr, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

The applicable temperature is the greater of the 10CFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is provided in Table 4-7.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table 4-7: Summary of the 10CFR50 Appendix G Requirements Operating Condition and Pressure MMinimu Temperature Requirement' I. Hydrostatic Pressure Test & Leak Test (Core is Not Critical) - Curve A _

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

2. At > 20% of preservice hydrotest Larger of ASME Limits or of highest pressure closure flange region initial RTNDT + 90'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 + 600 F*

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

b. Core critical - Curve C

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

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

_ __ hydrostatic pressure test

• 60 0F adder is included by GE as an additional conservatism. as discussed in Section 4.3.2.3.

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

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

((

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

))

4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0e17 n/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 BWR/2 through BWR/5 designs, and will be demonstrated in the following evaluation. The analyses took into account mechanical loading and anticipated thermal transients that bound BWR/2 through BWR/5 designs. 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 [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWR/6 components: the feedwater nozzles (F"N and the CRD penetrations (bottom head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-8 and 4-9.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table 4-8: Applicable BWR/4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B

s. > '.,Discontinuit Identification ........... ' i FW Nozzle CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Water Level Instrumentation Nozzle Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Table 4-9: Applicable BWR/4 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity'ldentification"<,-n.

CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Shellf Support Skirt**

Shroud Support Attachments" Core AP and Liquid Control Nozzle*

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

The P-T curves for the non-beltline region were conservatively developed for a large BWRI6 (nominal inside diameter of 251 inches). The analysis is considered appropriate for Fermi Unit 2 as the plant specific geometric values are comparable to the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version The generic value was adapted to the conditions at Fermi Unit 2 by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes in the upper vessel and bott6m head, respectively, has made the analysis different from a shell analysis such as the beltline. This was the result of the stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.

1))

An evaluation was performed for the bottom head wall thickness transition discontinuities located between the bottom head lower torus and upper torus and also between the bottom head torus and Shell #1. Appendix G of this report contains a detailed description of this evaluation. It was concluded that the discontinuities are bounded by the bottom head P-T curve developed in the following sections, and no further adjustment was required. Upjohn welds occur in various welds in the Fermi Unit 2 vessel; these welds are known to contain flaws. The limiting flaw, which exists in the beltline region, has been evaluated as discussed in Appendix I of this report.

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

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

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 GE Nuclear Energy NEDO-33133 Non-Proprietary Version that computation were K, = 143.6 ksi-in"' for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 840F. ((

The limit for the coolant temperature change rate is 20 0F1hr or less.

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

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

Mm = 1.85 for tIi<2 Mm = 0.926 fti for 2<4 t7<3.464 = 2.6206 Mm = 3.21 for t:>3.464 GE Nuclear Energy NEDO-33133 Non-Proprietary Version Km is calculated from the equation in Paragraph G-2214.1 [6] and Kib is calculated from the equation in Paragraph G-2214.2 [6]:

KM = Mm- = (( )) ksi-in" 2 Kib = (2/3) Mm = [ ] ksi-in"2 The total K, is therefore:

K, = 1.5 (Kim+ Kib) + Mm (asm + (2/3) GOb) = 143.6 ksi-in'r This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific Ki is based on the Kc equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) In [(Kl - 33.2) / 20.734) / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.734] / 0.02 (T - RTNDT)- 840F The generic curve was generated'by scaling 143.6 ksi-in"'2 by the nominal pressures and calculating the associated (T - RTNDT) as shown in Table 4-10.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version I

1]

The highest RTNDT for the bottom head plates and welds is 300F, as shown in Tables 4-1 and 4-3. ((

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Second, the P-T curve is dependent on the calculated K, value, and the K, value is proportional to the stress and the crack depth as shown below:

K, moa (a) " (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is V4. Thus, K, is proportional to R/(t)"2. The generic curve value of RI(t)"2, based on the generic BWR/6 bottom head dimensions, is:

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

The Fermi Unit 2-specific bottom head dimensions are R = 127.38 inches and t =7.38 inches minimum [19], resulting in:

Fermi Unit 2-specific: R / (t) 112 = 127.38 / (7.38)"2 = 47 inch"- (4-3)

Since the generic value of RI(t) "2 is larger, the generic P-T curve is conservative when applied to the Fermi 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 NEDO-331 33 Non-Proprietary Version The calculated value of K,for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [6] for comparison with KIR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical. Therefore, the K, value for the core not critical condition is (143.6 /1.5) -2.0 = 191.5 ksi-in""2.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the K equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:

(T - RTNDT) = In [(K,-33.2) / 20.734] / 0.02 CT - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 102'F The generic curve was generated by scaling 192 ksi-in'2 by the nominal pressures and calculating the associated (T - RTNDT) as shown in Table 4-11.

Table 4-11: Core Not Critical CRD Penetration K, and (T - RTNDT) as a Function of Pressure Nominal Pressure K! T- RTNDT (psig) --- (ksi-in1' 2) (°F) -

1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49 -14 The highest RTNDT for the bottom head plates and welds is 300F, as shown in Tables 4-1 and 4-3. ((

As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Table 4-9 and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than GE Nuclear Energy NEDO-33133 Non-Proprietary Version those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

• 1]

GE Nuclear Energy NEDO-33133 Non-Proprietary Version 4.3.2.1.3 Pressure Test - Non-Beltline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, K,, 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'" for an applied pressure of 1563 psig preservice hydrotest pressure. ((

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

To evaluate the results, K, is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or Xl). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of K, is shown below using the 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 (aIr,) from Figure A5-1 of WRC-175 is 1.4 where:

a= %4 ( t n 2.36 inches

=,2)1i2 t, = thickness of nozzle = 7.125 inches t, = thickness of vessel = 6.1875 inches r, = apparent radius of nozzle = ra+ 0.29 r,=7.09 inches r, = actual inner radius of nozzle = 6.0 inches rc = nozzle radius (nozzle comer radius) = 3.75 inches GE Nuclear Energy NEDO-33133 Non-Proprietary Version Thus, air, = 2.36 / 7.09 = 0.33. The value F(a/rQ), taken from Figure A5-1 of WRC Bulletin 175 for an alr, of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K,, is 1.5 a (7ra) /*F(alr):

Nominal K,= 1.5 - 34.97 . (7T* 2.36) 12

• 1.4 = 200 ksi-in"2 The method to solve for (T - RTNDT) for a specific K 1 is based on the K1, 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) = 104.2 0F

((

))

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

))

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

((

1))

The highest RTNDT for the feedwater. nozzle materials is 120F as shown in Table 4-2.

However, the. RTNDT was increased to 250 F to consider the stresses in the bottom headfCRD and recirculation inlet nozzle together with the initial RTNDT as described below. The generic pressure test P-T curve is applied to the Fermi Unit 2 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 25 0 F.

((:

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Second, the P-T curve is dependent on the K, value calculated. The Fermi Unit 2 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K, are shown below:

Vessel Radius to base metal, R, 127 inches Vessel Thickness, t, 6.69 inches Vessel Pressure, P, 1563 psig Pressure stress: a = PR / t = 1563 psig - 127 inches / (6.69 inches) = 29,671 psi. The dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 32.64 ksi. The factor F (a/rQ)from Figure A5-1 of WRC-175 is determined where:

a=  % (tn 2+ t,,2)1/2 =2.31 inches tn= thickness of nozzle = 6.38 inches tv = thickness of vessel = 6.69 inches rn = apparent radius of nozzle = ri + 0.29 r,=7.29 inches rim= actual inner radius of nozzle = 6.13 inches rc = nozzle radius (nozzle comer radius) = 4.0 inches GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Thus, a/rn = 2.31 / 7.29 = 0.32. The value F(a/r,), taken from Figure A5-1 of WRC Bulletin 175 for an a/re of 0.32, is 1.5. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (ira) /2- F(a/rn):

Nominal K, = 1.5 - 32.64 - (i

• 2.31) 12 - 1.5 = 197.9 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-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences feedwater. flow that is colder relative to the vessel coolant.

Stresses were taken from a (( )) finite element analysis done specifically for the purpose of fracture toughness analysis (( )). Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 400F 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 (Kap) is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:

Kp = SF -a (7ca)% . F(a/r,) (4-4)

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version where SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/rI) is the shape correction factor.

GE Nuclear Energy - NEDO-33133 Non-Proprietary Version Finite element analysis of a nozzle comer flaw was performed to determine appropriate values of F(a/rn) for Equation 4-4. These values are shown in Figure A5-1 of WRC Bulletin 175 [15].

The stresses used in Equation 4-4 were taken from (( )) design stress reports for the feedwater nozzle. The stresses considered are primary membrane, apm, and primary bending, Cpb. Secondary membrane, os, and secondary bending, Usb, stresses are included in the total K1 by using ASME Appendix G [6] methods for secondary portion, Kq,:

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

GE Nuclear Energy NEDO-33133 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. K1p and Kt4 are added to obtain the total value of stress intensity factor, K1. A safety factor of 2.0 is applied to primary stresses for core not critical heatup/cooldown conditions.

Once K4 was calculated, the following relationship was used to determine (T - RTNDT).

The method to solve for (T - RTNDT) for a specific K 1 is based on the K 1, equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltlirie components was then used to establish the P-T curves.

(T - RTNDT) = In [(K4 - 33.2) / 20.734] /0.02 (4-6)

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

feedwater nozzle (( 0

)) analysis, where feedwater injection of 40 F 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 s = 45.0 ksi t, = 6.1875 inches Cypb = 0.22 ksi csb = 19.04 ksi a = 2.36 inches r, = 7.09 inches t, = 7.125 inches GE Nuclear Energy NEDO-33133 Non-Proprietary Version In this case the total stress, 60.29 ksi, exceeds the yield stress, ay,, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 (15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the inside surface temperature is used.)

R = [cys - apm + ((ototal - cry) / 30)]I (atotal - (pm) (4-7)

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

apm = 24.84 ksi asm = 9.44 ks!

Upb = 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, t"n = 3.072. The resulting value obtained was:

Mm = 1.85 for J`<2 Mm = 0.926 ft- for 2<./7<3.464 = 2.845 Mm 3.21 for t >3.464 The value F(a/rQ), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, is therefore, F (a / r.) = 1.4 Kip is calculated from Equation 4-4:

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

GE Nuclear Energy NEDO-33133 Non-Proprietary Version K4s = 2.845 - (9.44 + 2/3 - 11.10)

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

The total K, is substituted into Equation 46 to solve for (T - RTNOT):

(T - RTNDT) = In [(238.3- 33.2) / 20.734] / 0.02 (T - RTNDT) = 1150F The (( )) curve was generated by scaling the stresses used to determine the Kl; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 400F water injected into the hot reactor vessel nozzle. In the base case that yielded a Kg value of 238 ksi-in-, the pressure is 1050 psig and the hot reactor vessel temperature is 551.40 F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (T-turtion - 40) / (551.4 - 40).

From Kg the associated (T - RTNDT) can be calculated as shown in Table 4-13.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table 4-13: Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp. R (T- RTNDT)

(psig) - (F) 112)

(ksi-1ni(F) 1563 604 0.23 303 128 1400 588 0.34 283 124 1200 557 0.48 257 119 1050 551 0.58 238 115 1000 546 0.62 232 113 800 520 0.79 206 106 600 489 1.0 181 98 400 448 1.0 138 81

• Note: For each change in stress for each pressure and saturation temperature condition, there is a corresponding change to R that influences the determination of The highest non-beltline RTNDT for the feedwater nozzle at Fermi Unit 2 is 120F as shown in Table 4-2. However, the RTNDT was increased to 250F to consider the stresses in the bottom head/CRD and recirculation inlet nozzle as previously discussed. The generic curve is applied to the Fermi Unit 2 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 250F as discussed in Section 4.3.2.1.3.

[ [1 GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 4.3.2.2 CORE BELTLINE REGION The pressure-temperature (P-T) operating limits for the beltline region are determined according to the ASME Code [6]. As the beftline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature.

The stress intensity factors (K<), calculated for the beltline region according to ASME Code Appendix G procedures [6], were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate; values were calculated for 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. Due to the existence of a flaw in one of the Shell #2 axial welds, an evaluation was performed to determine requirements necessary to protect this flaw, as discussed in Appendix I and Section 4.3.2.2.2 below. Thermal stresses are calculated including clad thickness as defined by the ASME Code. As demonstrated in Tables 4-5 and 4-6, the ART is conservatively calculated using minimum wall thickness excluding clad thickness.

An evaluation was performed for the vessel wall thickness transition discontinuity located between the lower and lower-intermediate shells in the beltline region. Appendix G of this report contains a detailed description of this evaluation. It was concluded that the discontinuity is bounded by the beltline P-T curve developed in the following sections, and no further adjustment was required.

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 beltfine limits. The vessel shell, with an inside radius (R) to minimum thickness (tan) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

am PR / tin (4-8)

GE Nuclear Energy NEDO-33133 Non-Proprietary Version The stress intensity factor, Kim, is calculated using Paragraph G-2214.1 of the ASME Code.

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G (61 for comparison with Kic, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

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

for the pressure test condition:

Km

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

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

respectively).

GE's current practice for the pressure test curve is to add a stress intensity factor, K,,, for a coolant heatup/cooldown rate, specified as 200F/hr for Fermi Unit 2, to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatup/cooldown rate of 1000F/hr. The Kit calculation for a coolant heatup/cooldown rate of 100 0F/hr is described in Section 4.3.2.2.3 below.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version 4.3.2.2.2 Calculations for the Beftline Region - Pressure Test This sample calculation is for a pressure test pressure of 1055 psig at 32 EFPY. The following inputs were used in the beltline limit calculation:

A = -44 + 121 + 5 = 820 F Adjusted RTNDT = Initial RTNOT + Shift (Based on ART values in Table 4-6 and adjusted to protect the existing flaw discussed below)

Vessel Height H = 861.6 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to base metal) R= 127 inches Minimum Vessel Thickness (without clad) t = 6.125 inches Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1055 psi + (H - B) 0.0361 psi/inch = P psig (4-10)

= 1055 + (861.6-216.3) 0.0361 = 1078 psig Pressure stress:

a = PR/t (4-11)

= 1.078 - 127/6.125 = 22.35 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, t"n = 2.47. The resulting value obtained was:

Mm = 1.85 for _<2 Mm = 0.926 rt- for 2<Vt<3.464 = 2.29 Mm = 3.21 for ft>3.464 GE Nuclear Energy NEDO-33133 Non-Proprietary Version The stress intensity factor for the pressure stress is Km = Mm *a. The stress intensity factor for the thermal stress, K,,, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20°F/hr instead of 100°F/hr.

Equation 4-9 can be rearranged, and 1.5 Km substituted for K~c, to solve for (T - RTNDT).

Using the Kic equation of Paragraph A-4200 in ASME Appendix A [17], KIm = 51.2, and Kit= 2.39 for a 20°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 - 51.2 + 2.39 - 33.2) / 20.734] / 0.02

= 39.80 F T can be calculated by adding the adjusted RTNDT:

T =39.8 + 77 = 116.8 0F for P = 1055 psig at 32 EFPY As previously mentioned, a flaw exists in Weld 15-308B that is located at the upper boundary of the extended beltline region, at 374.6 inches above vessel '0'. An evaluation was performed, as detailed in Appendix I, to determine the impact of this flaw on the P-T curves, as its dimensions exceed the postulated 1/4T flaw upon which this evaluation is based. This evaluation was conservatively performed using the parameters for the minimum hydrotest pressure of 1030 psig. It was found that for 32 EFPY, the beltline hydrotest pressure test curve (Curve A) must be shifted 5FF in addition to the 770F shift defined in Table 4-6. Similarly, for 24 EFPY, the curve must be shifted such that the Curve A temperature at 1030 psig is 1130F. Because Curve A is bounded by 10CFR50 Appendix G requirements at 1030 psig, a shift of 130F is required in order to accomplish this and protect the beltline flaw.

Therefore, T is further adjusted by 50F to protect the flaw:

T= 116.80 F + 50 F = 121.80 F for P = 1055 psig at 32 EFPY.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 4.3.2.2.3 Beltline Region - Core Not CriticalHeatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section Xl Appendix G [6]:

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

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

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

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

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

'the following boundary conditions:

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

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/ 2p - GCx I f3 + 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 Kit for heatup and cooldown.

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

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

4.3.2.2.4 Calculations for the Beltline Region Core Not Critical Heatup/Cooldown This Fermi Unit 2 sample calculation is for a pressure of 1055 psig for 32 EFPY.. The core not critical heatup/cooldown curve at 1055 psig uses the same Kim calculation as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational condition rather than test condition; the operational condition necessitates the use of a higher safety factor. In addition, there is a Kt term for the thermal stress. The additional inputs used to calculate Kt are:

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Coolant heatup/cooldown rate, normally 1000F/hr G = 100 0F/hr I Minimum vessel thickness, including clad thickness C = 0.5365 ft (6.125" + 0.3125" = 6.4375")

Thermal diffusivity at 550°F (most conservative value) p = 0.354 ff/ hr [21]

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

AT = GC2 /2P (4-16)

= 100 - (0.5365)2/ (2 - 0.354) = 41'F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.2942) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Kit = Mt *AT = 11.96, can be calculated. The conservative value for thermal diffusivity at 5500 F is used for all calculations; therefore, K,, is constant for all pressures. Kim has the same value as that calculated in Section 4.3.2.2.2.

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

(T - RTNDT) = ln[((2

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

= ln[(2- 51.2+ 11.96 -33.2) /20.734] /0.02

= 68.2 0 F T can be calculated by adding the adjusted RTNDT:

T = 68.2 + 77 = 145.2 0F for P = 1055 psig at 32 EFPY As previously mentioned, a flaw exists in Weld 15-308B that is located at the upper boundary of the extended beltline region, at 374.6 inches above vessel '0'. An evaluation, as detailed in Appendix I, was performed to determine the impact of this flaw on the P-T curves, as its dimensions exceed the postulated 1/4T flaw upon which this evaluation is GE Nuclear Energy N EDO-33133 Non-Proprietary Version based. It was determined that the core not critical beltline curve calculated above bounds the requirements for protecting this flaw.

It is noted that the 32 EFPY core not critical beltline curve is bounded by upper vessel requirements at 1055 psig as can be seen in Figure 510 and Appendix B.

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 (( )) 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 Kc formulation to determine T - RTNDT, for pressures above 312 psig the P-T limits for all flange regions are bounded by the 10CFR50 Appendix G requirement of RTNDT + 900F (the largest T-RTNDT for the flange at 1563 psig is 730F). For pressures below 312 psig, the flange curve is bounded by RTNDT + 60 (the largest T - RTNDT for the flange at 312 psig is 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 Fermi Unit 2 at low pressures.

The approach used for Fermi' Unit 2 for the bolt-up temperature was based on the conservative value of (RTNDT+ 60), or the LST of the bolting materais, whichever is greater. The 600F adder is included by GE for two reasons: 1) the pre-1971 requirements of the ASME Code Section III, Subsection NA, Appendix G included the 600 F adder, and 2) inclusion of the additional 600F requirement above the RTNDT provides the additional assurance that a 1/4T flaw size is acceptable. As shown in Tables 4-1, 4-2, and 4-3, the limiting initial RTNDT for the closure flange region is represented by Shell #4 GE Nuclear Energy INEDO-33133 Non-Proprietary Version at 120F, and the LST of the closure studs is 700 F; therefore, the bolt-up temperature value used is the more conservative value of 720F. 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 Fermi 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 68'F limit, further extensive calculations would be required to justify a lower temperature. The 720F limit for the upper vessel and beltline region and the 680F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel. When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFR50 Appendix G [8], Table 1. Table 1 of [8] requires that core critical P-T limits be GE Nuclear Energy NEDO-331 33 GE Nuclear EnergyNEO313 Non-Proprietary Version 40'F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 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 (RTNDT + 600F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 720F, based on an RTNDT of 120F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 1600F or the temperature required for the hydrostatic pressure test (Curve A at 1055 psig). The requirement of closure region RTNDT + 160'F causes a temperature shift in Curve C at 312 psig.

GE Nuclear Energy NEDO-331 33 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 0F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20 0F/hr or less must be maintained at all times (see Appendix C for additional guidance).

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, Kjr, at 1/4T to be less than that at 3/4T for a given metal temperature.

GE Nuclear Energy NEDO-33133 Non-Proprietary Version The following P-T curves were generated for Fermi Unit 2:

• Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 24 and 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition.

• Separate P-T curves were developed for the upper vessel, beltline (at 24 and 32 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

• A composite P-T curve was also generated for the Core Critical condition at 24 and 32 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.

Using the fluence from Section 4.2.1.2, the P-T curves are beltline limited above 900 psig for Curve A and above 820 psig for Curve B for 24 EFPY. The 32 EFPY P-T curves are beltline limited above 840 psig for Curve A and upper vessel limited between 820 and 890 psig and beltline limited above 890 psig for Curve B.

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

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves

-Figure Table Numbers Cuve Curve Description'. Nu bersfor for

-, xNe n *, .. E' :.

-Presentaion of rf v-...6*S, Presentation of the P-T theP-TCurvesi Curves A Bottom Head Limits (CRD Nozzle) Figure 5-1 Tables B-1 & B-3 A Upper Vessel Limits (FW Nozzle) Figure 5-2 Tables B-1 & B-3 A Beitline Limits - 24 EFPY Figure 5-3 Table B-1 A Beltline Limits - 32 EFPY Figure 5-4 Table B-3 A Bottom Head and Composite Curve A- 24 EFPY Figure 5-5 Table B-2 A Bottom Head and Composite Curve A- 32 EFPY* Figure 5-6 Table B-4 B Bottom Head Limits (CRD Nozzle) Figure 5-7 Tables B-1 & B-3 B Upper Vessel Limits (FW Nozzle) Figure 5-8 Tables B-1 & B-3 B Beltline Limits - 24 EFPY Figure 5-9 Table B-1 B Beltline Limits - 32 EFPY Figure 5-10 Table B-3.

B Bottom Head and Composite Curve B - 24 EFPY* Figure 5-11 Table B-2 B Bottom Head and Composite Curve B - 32 EFPY* Figure 5-12 Table B-4 C Composite Curve 0- 24 EFPY* Figure 5-13 Table B-2 C Composite Curve C - 32 EFPY** Figure 5-14 Table B-4 ABC Limiting Curves - 24 EFPY** Figure 5-15 Table B-2 ABC Limiting Curves - 32 EFPY*** Figure 5-16 Table B-4

• 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.

The Limiting curves are the more limiting of four limits: 10CFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Limits (FW Nozzle), and Beltline Limits.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 1100

. cm I

2~1000 W

CL 900 _ INITIAL RTndt VALUE IS 0 _I 30'F FOR BOTTOM HEADI I-

-J Ut 800 U)

W.,

o 700 _ HEATUP/COOLDOWN I- RATE OF COOLANT

< 20FIHR K: 600 z

3 500

>) 400 a:

3L 300 200 - - BOTTOM HEAD LIMITS 100 0

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

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

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 1100 c 1000

c. 900 w INITIAL RTndt VALUE IS I25°F FOR UPPER VESSELI 0

I.-

HEATUPICOOLDOWN RATE OF COOLANT 0: 600 -< 20'F/HR Z

=I 500

-J in 400 uJ c

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 0F/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 1100 VW C 1000 L-I- 900 0 BELTLINE CURVE t..2 a) 800 ADJUSTED AS SHOWN:

w EFPY SHIFT (CF) 24 119 o

I.--

700 II, Us HEATUP/COOLDOWN 0: 600 RATE OF COOLANT z

< 20FIHR or 500 LU U) 400 LU 300 200

-BELTLINE LIMITS 10CFR50

___ __BOCTUP _ _ _

100 72F 0

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

Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 24 EFPY

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUE IS

-44°F FOR BELTLINE 1100 Is CL 900 0

BELTLINE CURVE CO 800 ADJUSTED AS SHOWN:

EFPY SHIFT (0F) w 32 126 o 700 I -

HEATUP/COOLDOWN W 600 RATE OF COOLANT z

< 20*F/HR 3 500

'U v) 400 300 200

-BELTLINE LIMITS 100 0

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

(°F)

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

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 -440 F FOR BELTLINE, 25*F FOR UPPER VESSEL, AND 1100 30'F FOR BOTTOM HEAD CL w

a 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

= EFPY SHIFT (°F)

CL 900 0 24 119 III

-J U) 800 U)

LU o

I--

700 C.)

I 600 Iz

> 500 a) 400 30 300 UPPER VESSEL 200 AND BELTLINE LIMITS

--- -- BOTTOM HEAD 100 CURVE 0 I I I .11 I I . I 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

Figure 5-5: Composite Pressure Test P-T Curves [Curve A] up to 24 EFPY

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 INITIAL RTndt VALUES ARE 1200 -44°F FOR BELTUNE, .

25°F FOR UPPER VESSEL, AND 1100 30'F FOR BOTTOM HEAD la 0.

w 1000 BELTLINE CURVES LU ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

X. 900 0 32 126

-j w

o 700 U HEATUP/COOLDOWN LUj RATE OF COOLANT M 600 ' 20FIHR I-i 500 Mu ca 400 3.

300

- UPPER VESSEL 200 AND BELTLINE LIMITS

---

BOTTOM HEAD 100 CURVE 0

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

Figure 5-6: Composite Pressure Test P-T Curves [Curve A] up to 32 EFPY

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 1100 Ia)

In

- 1000 Lu IL 900 lINITIAL RTndt VALUE ISl 0 144.6'F FOR BOTTOM HEADlI I-j .

LU a) 800 U)

Lu o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 100'F/HR It 600 Z

ff 500 O-Lu V)

U)

Lu 400 I- __.. ___I EDM 3L0 68_

300 200

-BOTTOM HEAD LUMITS 100 0

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

Figure 5-7: Bottom Head P-T Curve for Core Not Critical [Curve B]

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 1100 la cL 0 18000 I-

-I w

0- 900 INITIAL RTndt VALUE IS as I 25¢1 FOR UPPER VESSEL Lu cn 800 0 700 HEATUP/COOLDOWN I-L) RATE OF COOLANT

< 100FIHR I3 w 600 Z

F_

Mu a) 500 Lu 200

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

Figure 5-8: Upper Vessel P-T Curve for Core Not Critical [Curve B]

[1000F/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-33133 Non-Proprietary Version 1400 1300 1200 1100 0,

1000 BELTLINE CURVE CL ADJUSTED AS SHOWN:

EFPY SHIFT (0F) 900 24 106 0

-j 800 U)

'U 700 HEATUP/COOLDOWN RATE OF COOLANT cc 600 < 100°FIHR LU 500 0:

400 300 200

- BELTLINE LIMITS 100 0

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

(°F)

Figure 5-9: Beltline P-T Curve for Core Not Critical [Curve B] up to 24 EFPY

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200

. I3 / I1 1100 a.

I 0' 1000 BELTLINE CURVE w ADJUSTED AS SHOWN:

EFPY SHIFT (-F)

0. 900 32 121 0

Its-MU 01 800 o

I--

700 C.) HEATUP/COOLDOWN RATE OF COOLANT

_ _G _ _ / _

= 600 ' 100'F-IR Z

3 50o Lu Cn *400 U)

- 3L 300 200 10CFR50 -BELTLINE LIMITS BOTUP 100

- ___- 72F .. -I__

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

Figure 5-10: Beltline P-T Curve for Core Not Critical [Curve B] up to 32 EFPY

[1 0 OF/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE

-44'F FOR BELTLINE, 250F FOR UPPER VESSEL, 1100 AND 44.6 0F FOR BOTTOM HEAD D 1000 BELTUNE CURVES III ADJUSTED AS SHOWN:

aL 900 EFPY SHIFT (0F) 0 24 106 Vn 800 al700 Us o 700 w

HEATUPICOOLDOWN I-0 RATE OF COOLANT W- 600 < 100FIHR Z

us

i 500 v)

Cl) 400 Lu CC 300 UPPER VESSEL 200 AND BELTLINE LIMITS 100 BOTTOM HEAD CURVE 0 I I  : I I I 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (IF)

Figure 5-11: Composite Core Not Critical P-T Curves [Curve B)up to 24 EFPY

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 1300 1200 INITIAL RTndt VALUES ARE

-440 F FOR BELTLINE, 25°F FOR UPPER VESSEL, 1100 AND la 44.60F FOR BOTTOM HEAD

,, 1 000 BELTLINE CURVES

-J ADJUSTED AS SHOWN:

9L 900 EFPY SHIFT (CF) 32 121 uJ Un 800 o 700 I-I L) HEATUPJCOOLDOWN RATE OF COOLANT 600 I

Z < 100°F/HR j 500 mU U) 400 3L 300

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

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

(°F)

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

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

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 -44 0 F FOR BELTLINE, 25@F FOR UPPER VESSEL, 1200 AND 30'F FOR BOTTOM HEAD 1100

.BELTLINE CURVE 0 1000 ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

L'* 900 24 106 0

t-

"u C 800 HEATUP/COOLDOWN RATE OF COOLANT

< 1000FIHR 0 700 It' Lu M 600 2

500 cc w =30 200

-BELTLINE AND NON-BELTLINE 100 - I _ Minimum Criticality . LIMITS I I I Temp~eratur-eI72F 0

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

Figure 5-13: Composite Core Critical P-T Curves [Curve C] up to 24 EFPY

[1001F/hr or less coolant heatup/cooldown]

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version 1400 INITIAL RTndt VALUES ARE 1300 -44°F FOR BELTLINE, 25°F FOR UPPER VESSEL, 1200 AND 30°F FOR BOTTOM HEAD 1100 Dt

~-00 BELTLINE CURVE ADJUSTED AS SHOWN:

00D 1000

-J 9000 EFPY SHIFT (°F)

I 32 121 w

r~ 800 HEATUP/COOLDOWN I-.

RATE OF COOLANT

< 100°F/HR o 700 CD U 600 z

0~ 500 nf 400 rr 300

. _ I

)400 200

-BELTLINE AND NON-BELTLINE 100 - - - Minimum Criticality LIMITS 1 I - - lTemperature I 72Fl 0 25 50 75 100 125 150 175 200 225 250 :

MINIMUM REACTOR VESSEL METAL TEMPERATURE

( 0F)

Figure 5-14: Composite Core Critical P-T Curves [Curve C] up to 32 EFPY

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

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Curve A; Curve B; Curve C INITIAL RTndt VALUES 1400ARE 140- - -- 44°F FOR BELTLINE, 25*F FOR UPPER 1 - - . l VESSEL, 1300 -AND 30°F FOR BOTTOM 1200 ___ I_ HEAD 1200 CURVES A&C AND

/___ 44.6°F FOR BOTTOM 1100 .- /-- 1 HEAD CURVE B

- 1000 100 - ___ - ___ BELTLINE CURVES ADJUSTED AS SHOWN:

___ 900 EFPY

-- SHIFT ('F) o.900 - __ - - - _

0 .24 119 (Curve A)

- 24 106 (Curves B,C) w 0 U,

o 800 -_ _ - HEATUPICOOLDOWN

> RATE OF COOLANT W < 20F/HR FOR CURVE A, o 700 -_- _ _ - 100'FIHR FOR CURVES B&C c A, B, C - LIMITING CURVE t 600 - - .......ll

_ A - PRESSURE TEST WITH FUEL IN THE VESSEL

i 500 - --

. B- NON-NUCLEAR 0 -HEATUPICOOLDOWN o 400. CORE NOT CRITICAL C - NUCLEAR 300 31 .IG- _ HEATUP/COOLDOWN CORE CRITICAL 200 -A LIMITING CURVE BLU , / l- B LIMITING CURVE 100 72'F - -- C LIMITING CURVE l l . [CURVES A,B, Cl IARE VAUD UP TO 0 124 EFP 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL.METAL TEMPERATURE

(*F)

Figure 5-15: Limiting P-T Curves [Curves A, B, and C] up to 24 EFPY

[20°F/hr or less coolant heatup/cooldown for Curve A and 100°F/hr or less coolant heatup/cooldown for Curves B and C]

GE Nuclear Energy NEDO-33133 Non-Proprietary Version iE00 Curve A; Curve B; Curve C INITIAL RTndt VALUES 1400 ARE F FOR BELTLINE 25-F FOR UPPER 1300 -_ l_VESSEL,

_/_

AND 30-F FOR BOTTOM HEAD 1200 _ /_l_CURVES A&C 44.6-F FOR BOTTOM 1100D - .HEAD CURVE B 1000 - - - - - _ - BELTLINE CURVES ADJUSTED AS SHOWN:

/ / /EFPY SHIFT (-F) 0L 900 32 126 Curve A 0

I- 32 121 Curves B,C a 800- - _-

-HEATUP/COOLDOWN

> lRATE OF COOLANT O_ < 20-F/HR FOR CURVE A, o

o 700 I100-FIHR FOR CURVES B&C tit 60_ __A, B, C - LIMITING CURVE A - PRESSURE TEST WITH E I FUEL IN THE VESSEL M 500_

rx I B - NON-NUCLEAR 0 - -HEATUP/COOLDOWN cn 4_ CORE4NOT CRITICAL C- NUCLEAR 300 -_3t2 3S -. =7 F = l HEATUP/COOLDOWN

/I--/CORE CRITICAL 20 i-A LIMITING CURVE B l l-l B LIMITING CURVE 100 - - l. C LIMITING CURVE t 25 L...... l lCURVES OJl A, B.C RE VA~LID,UP TO 0

O 32 EFPy 0 25 50 75 100 125 150 175 200 225 MINIMUM REACTOR VESSEL METALTEMPERATURE (F)

Figure 5-16: Limiting P-T Curves [Curves A, B, and C] up to 32 EFPY

[20°F/hr or less coolant heatup/cooldown for Curve A and 100 0F/hr or less coolant heatup/cooldown for Curves B and C]

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

6.0 REFERENCES

1. Caine, T.A., "Implementation of Regulatory Guide 1.99 Revision 2 for the Fermi 2 Nuclear Power Plant", SASR 90-73, Revision 1, GE Nuclear Energy, San Jose, CA, January 1991.

2. GE Drawing Number 761E246, "Reactor Vessel Thermal Cycles - Reactor Vessel",

GE-NED, San Jose, CA, Revision 1 (GE Proprietary).

3. GE Drawing Number 158B8369, "Reactor Vessel Nozzle Thermal Cycles - Reactor System", GE-NED, San Jose, CA, Revision 2 (GE Proprietary).

4. Wu, T., "DTE Energy Fermi-2 Energy Center Neutron Flux Evaluation", GE-NE-0000-0031-6254-RI, Revision 1, GE Nuclear Energy, San Jose, CA, February 2005 (GE Proprietary).

5. "BWR Vessel and Internals Project BWR Integrated Surveillance Program Implementation Guidelines", BWRVIP-102, EPRI, Palo Alto, CA, June 2002 (EPRI Proprietary).

6. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section Xl of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

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

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

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

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

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

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

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

12. QA Records and RPV CMTRs for Fermi 2, GE PO #205-H0399, Contract #2667, Manufactured by Combustion Engineering, Inc., Chattanooga, Tenn.

13. Letter Number PFIP-04-020710801.26, J. Vargas (Detroit Edison) to G. Carlisle (GE),

"Retransmittal of Design Input Request Response for WIN 8 Pressure Temperature Curves", October 21, 2004.

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

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

Welding Research Council Bulletin 175, August 1972.

16. ((

17. 'Analysis of Flaws", Appendix A to Section Xl of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

18. [

19. Bottom Head and Feedwater Nozzle Dimensions:

a. Drawing Number E232-900, Revision 5, "Bottom Head Machining and Welding",

Combustion Engineering, Chattanooga, Tennessee (VPF# 1976-014).

b. Drawing Number E232-910, Revision 8, "Nozzle.Details", Combustion Engineering, Chattanooga, Tennessee (VPF # 1976-064).

20. ((

- . ))

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

21. "Materials - Properties', Part D to Section II of the ASME Boiler & Pressure Vessel Code, 1998 Edition with Addenda through 2000.

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy NEDO-33133 Non-Proprietary Version I]

A-2

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Per ASME Code Appendix G, Section G2223 (c), fracture toughness analysis to demonstrate protection against non-ductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5" or less provided the lowest service temperature is not lower than RTNDT plus 60*F. Inconel (or Alloy 600) and stainless steel discontinuities require no fracture toughness evaluations.

Nozzle or Appurtenance Material Reference Remarks MK 321-05 High Pressure Seal Leak Detector Nozzle (attached to Nozzles less than 2.5" require no Shell Flange) SA 106 Gr B 1, 27 fracture toughness evaluation.

MK 315-04 Core AP and Liquid Control Nozzle (See Table A-1 for Penetration in the Bottom Head Nozzles made from Inconel require Dollar Plate) Inconel 1, 8, 20 no fracture toughness evaluation.

Nozzles less than 2.5" in thickness MK 315-01 Instrumentation require no fracture toughness Nozzles SA 508 CLI 1,20,28 evaluation.

Nozzles less than 2.5' in thickness require no fracture toughness MK 315-14 Drain Nozzle SA 508 CL 1 1, 8, 20 evaluation.

Components made from Inconel SB- 66 require no fracture toughness MK 327-01 Shroud Support Inconel 1, 23, 29 evaluation.

Not a pressure boundary component; therefore requires no MK 324-03 Basin Seal Skirt SA 515 GR 70 1. 26 fracture toughness evaluation.

Not a pressure boundary MK 325-02 and 325-03 A 36 Carbon component; therefore requires no Thermocouple Pad Steel 1,30, 31 fracture toughness evaluation. -

MK 310-02 through 310-29 CRD Components made from Alloy 600 Stub Tubes (in Bottom Head Dollar SB-1 67 and less than 2.5" require no Plate and Lower Torus) Alloy 600 1,22,23,24 fracture toughness evaluation.

Appurtenances made from MK 323-10 and 323-11 SA351 Gr Stainless Steel require no fracture Surveillance Brackets CF8M 1, 32, 33 toughness evaluation.

A-3

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations, Continued Nozzle or Appurtenance Material Reference Remarks Appurtenances made from SA351 Gr Stainless Steel require no fracture MK 323-06 Core Spray Brackets CF8M 1, 32, 33 toughness evaluation.

Appurtenances made from MK 323-07 Feedwater Sparger SA351 Gr Stainless Steel require no fracture Brackets CF8M 1, 32, 33 toughness evaluation.

Appurtenances made from MK 323-05 Steam Dryer Support SA351 Gr Stainless Steel require no fracture Lug CF8M 1, 32,33 toughness evaluation.

Appurtenances made from SA351 Gr Stainless Steel require no fracture MK 323-02 Guide Rod Bracket CF8M 1, 32 toughness evaluation Loading only occurs during outages. Not a pressure boundary SA 533 Gr B component; therefore requires no MK 319-06 Top Head Lifting Lugs CL 1 1, 5,6,17 fracture toughness evaluation.

Not a pressure boundary MK 319-15 Steam Dryer Hold SA 533 Gr B component; therefore requires no Down Bracket CL 1 1, 5 fracture toughness evaluation.

Upjohn welds do not require fracture toughness evaluation. It is, however, noted that any Upjohn welds within the beltline region are Upjohn Welds N/A NIA covered by the beltline curves.

A-4

GE Nuclear Energy NEDO-33133 Non-Proprietary Version APPENDIX A

REFERENCES:

1. Vessel Drawings

• Drawing Number B-230-481, Revision 5, "Drawing Plan List", Combustion Engineering, Inc., Windsor, Conn. (VPF # 1976-087).

• Drawing Number E232-895, Revision 4, "General Arrangement Elevation for 251" ID BWR", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF # 1976-077).

• Drawing Number E232-907, Revision 4, "Vessel Assembly ", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF # 1976-069).

• Drawing Number E232-901, Revision 11, "Lower Vessel Shell Assembly -

Machining & Welding", Combustion Engineering, Inc., Chattanooga, Tenn.

(VPF # 1976-11)..

• Drawing Number E232-902, Revision 16, "Upper Vessel Shell Assembly -

Machining & Welding", Combustion Engineering, Inc., Chattanooga, Tenn.

(VPF # 1976-012).

• Drawing Number E232-926, Revision 6, "As-Built Dimensions for 251" ID BWR",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF# 1976-109).

• Drawing Number E233-308, Revision 3, "As-Built Dimensions for 251" ID BWR",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF # 1976-212).

• Drawing Number E232-925, Revision 2, "Material Identification", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF # 1976-080).

. QA Records and RPV CMTRs for Fermi 2, GE PO Number 205-H0399, Contract Number 2667, Manufactured by Combustion Engineering, Inc.,

• Chattanooga, Tenn.

2. Letter Number PFIP-04-0207/0801.26, J. Vargas (Detroit Edison) to G. Carlisle (GE), "Retransmittal of Design Input Request Response for WIN 8 Pressure Temperature Curves', October 21, 2004.

3. Wu, T., 'DTE Energy Fermi-2 Energy Center Neutron Flux Evaluation", GE-NE-0000-0031-6254-RI, Revision 1, GE Nuclear Energy, San Jose, CA, February 2005 (GE Proprietary).

A-5

GE Nuclear Energy NEDO-33133

'Non-Proprietary Version

4. Drawing Number SC-2667-1, Revision 0, "Closure Head Flange Ordering Sketch",

Combustion Engineering, Inc., Windsor, Conn. (VPF # 1976-002).

5. Drawing Number E232-913, Revision 4, "Closure Head Machining & Welding",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF# 1976-078).

6. Drawing Number E232-914, Revision 5, "Closure Head Final Machining",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-079).

7. Drawing Number SC-2667-2, Revision 0, "Vessel Flange Ordering Sketch",

Combustion Engineering, Inc., Windsor, Conn. (VPF # 1976-003).

8. Drawing Number E232-900, Revision 5, "Bottom Head Machining and Welding",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF#1976-014).

9. Drawing Number C-200-061-2-F, Revision B, "Recirculation Outlet Nozzle", Ladish Company, Cudahy, Wisconsin (VPF #1976-067).

10. Drawing Number C-200-061-2, Revision 3, "Recirculation Outlet Nozzle',

Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-022).

11. Drawing Number E232-908, Revision 3, "Nozzle Details", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-050).

12. Drawing Number B-245-201, Revision 1, "Recirculation Inlet Nozzle", Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-021).

13. Drawing Number E232-910, Revision 8, "Nozzle Details", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-064).

14. Drawing Number B-245-205, Revision 0, "Core Spray Nozzle", Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-052).

15. Drawing Number E232-912, Revision 3, 'Closure Head Nozzle Details",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF#1976-117).

16. Drawing Number E232-834, Revision 1, "As Built Location of Weld Seams -

Closure Head", Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-259).

17. Drawing Number E232-896, Revision 2, "General Arrangement Plans", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-075).

A-6

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

18. Drawing Number SB-2667-36, Revision 0, "Vent Nozzle Forging", Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-128).

19. Drawing Number B-245-200, Revision 0, 'Jet Pump Instrumentation Nozzle",

Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-020).

20. Drawing Number E232-909, Revision 7, "Nozzle Details", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-051).

21. Drawing Number 6-245-204, Revision 1, 'C.R.D. Hyd. System Return Nozzle",

Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-049).

22. Drawing Number E232-904, Revision 4, "Bottom Head Penetrations", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-066).

23. Drawing Number E232-905, Revision 7, Vessel Machining", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-081).

24. Drawing Number E232-927, Revision 2, "As Built Dimensions Bottom Head Penetrations", Combustion Engineering, Inc., Chattanooga, Tenn.

(VPF #1976-108).

25. Drawing Number E232-903, Revision 7, "Vessel Support Skirt Assy and Details",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF#1976-065).

26. Drawing Number E232-918, Revision 4, "Vessel External Attachments",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-111).

27. Drawing Number E232-915, Revision 2, "Miscellaneous Details", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF#1976-082).

28. Drawing Number SB-2667-37, Revision 0, "Instrumentation Nozzle Forging",

Combustion Engineering, Inc., Windsor, Conn. (VPF #1976-129).

29. Drawing Number E232-921, Revision 5, "Shroud Support Details and Assembly",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF#1976-083).

30. Drawing Number E232-919, Revision 7, "Vessel External 'Attachments",

Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-107).

31. Drawing Number E233-309, Revision 3, 'Thermocouple Arrangement - Upper Vessel", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF #1976-219).

A-7

GE Nuclear Energy NEDO-33133 Non-Proprietary Version

32. Drawing Number E232-917, Revision 8, "Vessel Internal Attachments", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF#1976-103).

33. Drawing Number E233-305, Revision 1, Internal Bracket Removal and Replacement", Combustion Engineering, Inc., Chattanooga, Tenn.

(VPF #1976-215).

Note: As-built drawings for the top head flange and vessel flange are not available.

Design sketches in References 4 and 7 have been used in lieu of the as-built drawings.

A-8

GE Nuclear Energy NEDO-33133 Non-Proprietary Version APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-I

GE Nuclear Energy N EDO-33133 Non-Proprietary Version TABLE B-1. Fermi Unit 2 P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 F/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-1, 5-2. 5-3, 5-7, 5-8 & 5-9

- BOTTOM UPPER. 24 EFPY BOTTOM UPPER 24 EFPY

. HEAD :VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B 0 0 (PSIG) ( F) - ( F) .(OF)- (IF) (OF) (OF) 0 68.0 72.0 72.0 68.0 72.0 72.0 10 68.0 72.0 72.0 68.0 72.0 72.0 20 68.0 72.0 72.0 68.0 72.0 72.0 30 68.0 72.0 72.0 68.0 72.0 72.0 40 68.0 72.0 72.0 68.0 72.0 72.0 50 68.0 72.0 72.0 68.0 72.0 72.0 60 68.0 72.0 72.0 68.0 72.0 72.0 70

• 68.0 72.0 72.0 68.0 72.0 72.0 80 68.0 72.0 72.0 68.0 72.0 72.0 90 68.0 72.0 72.0 68.0 72.0 72.0 100 68.0 72.0 72.0 68.0 72.0 72.0 110 68.0 72.0 72.0 68.0 72.0 72.0 120 68.0 72.0 72.0 68.0 72.0 72.0 130 68.0 72.0 72.0 68.0 72.0 72.0 140 68.0 72.0 72.0 68.0 72.0 72.0 150 68.0 72.0 72.0 68.0 72.0 72.0 160 68.0 72.0 72.0 68.0 72.0 72.0 170 68.0 72.0 - 72.0 68.0 72.0 72.0 180 68.0 - 72.0 72.0 68.0 72.9 72.0 190 68.0 72.0 72.0 68.0 75.2 72.0 200 68.0 72.0 72.0 68.0 77.3 72.0 210 68.0 72.0 72.0 68.0 79.3 72.0 220 68.0 72.0 72.0 68.0 81.3 72.0 B-2

GE Nuclear Energy .NEDO-33133 Non-Proprietary Version TABLE B-1. Fermi Unit 2 P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100OF/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 24 EFPY BOTTOM UPPER 24 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVE A CURVE A CURVEB CURVE B- CURVEB (PSIG) ('F) (0F) (OF) (0F) (OF) (*F) 230 68.0 72.0 72.0 68.0 83.1 72.0 240 68.0 72.0 72.0 68.0 84.9 72.0 250 68.0 72.0 72.0 68.0 86.6 72.0 260 68.0 72.0 72.0 68.0 88.2 72.0 270 68.0 72.0 72.0 68.0 89.8 72.0 280 68.0 72.0 72.0 68.0 91.3 72.0 290 68.0 72.0 72.0 68.0 92.8 72.0 300 68.0 72.0 72.0 68.0 94.2 72.0 310 68.0 72.0 72.0 68.0 95.5 72.0 312.5 68.0 72.0 72.0 68.0 95.9 72.0 312.5 68.0 102.0 102.0 68.0 132.0 132.0 320 68.0 102.0 102.0 68.0 132.0 132.0 330 68.0 102.0 102.0 68.0 132.0 132.0 340 68.0 102.0 102.0 68.0 132.0 132.0 350 68.0 102.0 102.0 68.0 132.0 132.0 360 68.0 102.0 102.0 68.0 132.0 132.0 370 . 68.0 102.0 102.0 68.0 132.0 132.0 380 68.0 102.0 102.0 68.0 132.0 132.0 390 68.0 102.0 102.0 68.0 132.0 132.0 400 68.0 102.0 102.0 68.0 132.0 132.0 410 68.0 102.0 102.0 68.0 132.0 132.0 420 68.0 102.0 102.0 68.0 132.0 132.0 430 68.0 102.0 102.0. 68.0 132.0 132.0 440 68.0 102.0 102.0 68.0 132.0 132.0 450 68.0 102.0 . 102.0 68.0 132.0 132.0 B-3

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE B-1. Fermi Unit 2 P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 OF/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 24 EFPY BOTTOM UPPER 24 EFPY HEAD VESSEL: BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVEA CURVE A , CURVE B CURVEB CURVE B 0 0 . (F)

(PSIG) ( F) ( F) (°F) (OF) (OF) 460 68.0 102.0 102.0 68.0 132.0 132.0 470 68.0 102.0 102.0 68.0 132.0 132.0 480 68.0 102.0 102.0 68.0 132.0 132.0 490 68.0 102.0 102.0 68.0 132.0 132.0 500 68.0 102.0 102.0 68.0 132.0 132.0 510 68.0 102.0 102.0 68.0 132.0 132.0 520 68.0 102.0 102.0 68.0 132.0 132.0 530 68.0 102.0 102.0 68.0 132.0 132.0 540 68.0 102.0 102.0 68.0 132.0 132.0 550 68.0 102.0 102.0 69.5 132.0 132.0 560 68.0 102.0 102.0 71.3 132.0 132.0 570 68.0 102.0 102.0 73.0 132.0 132.0 580 68.0 102.0 102.0 74.6 132.0 132.0 590 68.0 102.0 102.0 76.2. 132.0 132.0 600 68.0 102.0 102.0 77.8 132.0 132.0 610 68.0 102.0 102.0 79.3 132.0 132.0 620 68.0 102.0 102.0 80.7 132.0 132.0 630 68.0 102.0 102.0 82.1 132.0 132.0 640 68.0 102.0 102.0 83.5 132.0 132.0 650 68.0 102.0 102.0 84.8 132.0 132.0 660 68.0 102.0 102.0 86.1 132.0 132.0 670 68.0 102.0 102.0 87.4 132.0 132.0 680 68.0 102.0 102.0 88.7 132.0 132.0 690 68.0 102.0 102.0 89.9 132.0 132.0 700 68.0 102.0 102.0 91.0 132.0 132.0 B-4

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-1. Fermi Unit 2 P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 OF/hrfor Curves B & C and 20 *F/hr for Curve A for Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 24 EFPY . BOTTOM UPPER 24 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVEB CURVE B 0

(PSIG) (OF) ( F) (OF) (OF) (OF) (0 F) 710 68.0 102.0 102.0 92.2 132.0 132.0 720 68.0 102.0 102.0 93.3 132.0 132.0 730 68.0 102.0 102.0 94.4 132.0 132.0 740 68.0 102.0 102.0 95.5 132.0 132.0 750 68.0 102.0 102.0 96.6 132.0 132.0 760 68.0 102.0 102.0 97.6 132.0 132.0 770 68.0 102.0 102.0 98.6 132.0 132.0 780 68.0 102.0 102.0 99.6 132.0 132.0 790 68.0 102.0 102.0 100.6 132.0 132.0 800 68.0 102.0 102.0 101.5 132.0 132.0 810 68.0 102.0 102.0 102.5 132.0 132.0 820 68.0 102.0 102.0 103.4 132.0 132.0 830 68.0 102.0 102.0 104.3 132.2 132.0 840 68.0 102.0 102.0 105.2 132.6 132.0 850 68.6 102.0 102.0 106.0 132.9 132.0 860 69.6 102.0 102.0 106.9 133.3 132.0 870 70.6 102.0 102.0 107.7 133.6 132.0 880 71.5 102.0 102.0 108.6 134.0 132.0 890 72.5 102.0 102.0 109.4 134.3 132.0 900 73.4 102.0 102.0 110.2 134.7 132.0 910 74.4 102.0 102.2 111.0 135.0 132.0 920 75.3 102.0 103.2 111.7 135.4 132.0 930 76.1 102.0 104.1 112.5 135.7 132.0 940 77.0 102.5 105.1 113.3 136.0 132.0 950 77.9 103.1 106.0 114.0 136.4 132.0 B-5

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-1. Fermi Unit 2 P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTTOM UPPER 24 EFPY BOTTOM UPPER 24 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 960 78.7 103.7 106.9 114.7 136.7 132.0 970 79.6 104.3 107.8 115.5 137.0 132.0 980 80.4 104.9 108.7 116.2 137.4 132.0 990 81.2 105.5 109.6 116.9 137.7 132.0 1000 82.0 106.1 110.4 117.6 138.0 132.0 1010 82.7 106.7 111.3 118.2 138.3 132.0 1020 83.5 107.2 112.1 118.9 138.6 132.0 1030 84.3 107.8 112.9 119.6 139.0 132.0 1040 85.0 108.4 113.7 120.2 139.3 132.0 1050 85.7 108.9 114.5 120.9 139.6 132.0 1055 86.1 109.2 114.9 121.2 139.7 132.0 1060 86.4 109.5 .115.3 121.5 139.9 132.0 1070 87.2 110.0 116.0 122.1 140.2 132.0 1080 87.9 110.5 116.8 122.8 140.5 132.0 1090 88.6 1111 117.5 123.4 140.8 132.3 1100 89.2 111.6 118.3 124.0 141.1 132.8 1105 89.6 111.8 118.6 124.3 141.3 133.1 1110 89.9 112.1 119.0 124.6 141.4 133.4 1120 90.6 112.6 119.7 125.2 141.7 133.9 1130 91.2 113.1 120.4 125.8 142.0 134.5 1140 91.9 113.6 121.1 126.3 142.3 135.0 1150 92.5 114.1 121.8 126.9 142.6 135.5 1160 93.1 114.6 122.4 127.5 142.9 136.1 1170 93.8 115.1 123.1 128.0 143.2 136.6 1180 94.4 115.6 123.7 128.6 143.5 137.1 B-6

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-1. Fermi Unit 2 P-T Curve Values for24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 *Ffhr for Curves B & C and 20 0F/hr for Curve A for Figures 5-1, 5-2, 5-3, 5-7, 5-8 & 5-9 BOTOM UPPER 24 EFPY BOTTOM UPPER: 24 EFPY HEAD VESSEL BELTLINE - HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE . CURVE B (PSIG) .(F) ('F) (OF) - (,F)

(F)e ..'(*F, 1190 95.0 116.1 124.4 129.1 143.7 137.6 1200 95.6 116.5 125.0 129.7 144.0 138.1 1210 96.2 117.0 125.7 130.2 144.3 138.6 1220 96.8 117.5 126.3 130.8 144.6 139.1 1230 97.3 117.9 126.9 131.3 144.9 139.6 1240 97.9 118.4 127.5 131.8 145.2 140.1 1250 98.5 118.8 128.1 132.3 145.4 140.6 1260 99.0 119.3 128.7 132.8 145.7 141.0 1270 99.6 119.7 129.3 133.3 146.0 141.5 1280 100.1 120.2 129.8 133.8 146.2 142.0 1290 100.7 120.6 130.4 134.3 146.5 142.4 1300 101.2 121.0 131.0 134.8 146.8 142.9 1310 101.7 121.5 131.5 135.3 147.1 143.3 1320 102.3 121.9 132.1 135.8 147.3 143.8 1330 102.8 122.3 132.6 136.2 147.6 144.2 1340 103.3 122.7 133.2 136.7 147.8 144.7 1350 103.8 123.1 133.7 137.2 148.1 145.1 1360 104.3 123.6 134.2 137.6 148.4 145.5 1370 104.8 124.0 134.8 138.1 148.6 146.0 1380 105.3 124.4 135.3 138.5 148.9 -146.4 1390 105.8 124.8 135.8 139.0 149.1 146.8 1400 106.3 125.2 136.3 139.4 149.4 147.2 B-7

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-2. Fermi Unit 2 Composite P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-5, 5-11, 5-13 & 5-15 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT. HEAD BELTLINE AT LIMITING 24 EFPY 24 EFPY 24 EFPY PRESSURE CURVE A CURVEA 'CURVE B CURVE B CURVE C (PSIG) (*F) (OF) (OF) - (OF) (OF) 0 68.0 72.0 68.0 72.0 72.0 10 68.0 72.0 68.0 72.0 72.0 20 68.0 72.0 68.0 72.0 72.0 30 68.0 72.0 68.0 72.0 72.0 40 68.0 72.0 68.0 72.0 72.0 50 68.0 72.0 68.0 72.0 72.0 60 68.0 72.0 68.0 72.0 72.0 70 68.0 72.0 68.0 72.0 72.2 80 68.0 72.0 68.0 72.0 78.2 90 68.0 72.0 68.0 72.0 83.3 100 68.0 72.0 68.0 - 72.0 87.8 110 68.0 72.0 68.0 72.0 91.9 120 68.0 72.0 68.0 72.0 95.7 130 68.0 72.0 68.0 72.0 99.2 140 68.0 72.0 68.0 72.0 .102.4 150 68.0 72.0 68.0 72.0 105.2 160 68.0 72.0 68.0 72.0 107.9 170 68.0 72.0 68.0 72.0 110.5 180 68.0 72.0 68.0 72.9 112.9 190 68.0 72.0 68.0 75.2 115.2 200 68.0 72.0 68.0 77.3 117.3 210 68.0 72.0 68.0 79.3 i 19.3 220 68.0 72.0 68.0 81.3 121.3 E-8

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE B-2. Fermi Unit 2 Composite P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 *F/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-5, 5-11, 5-13 & 5-15 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

- HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 24 EFPY 24 EFPY 24 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG)  ;-( 0F) (OF) (OF) (OF) (OF) 230 68.0 72.0 68.0 83.1 123.1 240 68.0 72.0 68.0 84.9 124.9 250 68.0 72.0 68.0 86.6 126.6 260 68.0 72.0 68.0 88.2 128.2 270 68.0 72.0 68.0 89.8 129.8 280 68.0 72.0 68.0 91.3 131.3 290 68.0 72.0 68.0 92.8 132.8 300 68.0 72.0 68.0 94.2 134.2 310 68.0 72.0 68.0 95.5 135.5 312.5 68.0 72.0 68.0 95.9 135.9 312.5 68.0 102.0 68.0 132.0 172.0 320 68.0 102.0 68.0 132.0 172.0 330 68.0 102.0 68.0 132.0 172.0 340 68.0 102.0 68.0 132.0 172.0 350 68.0 102.0 68.0 132.0 172.0 360 68.0 102.0 68.0 132.0 172.0 370 68.0 102.0 68.0 132.0 172.0 380 68.0 102.0 68.0 132.0 172.0 390 68.0 102.0 68.0 132.0 172.0 400 68.0 102.0 68.0 132.0 172.0 410 68.0 102.0 68.0 132.0 172.0 420 68.0 102.0 68.0 132.0 172.0 430 68.0 102.0 68.0 132.0 172.0 440 68.0 102.0 68.0 132.0 172.0 450 68.0 102.0 68.0 132.0 172.0 460 68.0 102.0 68.0 132.0 172.0 B-9

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-2. Fermi Unit 2 Composite P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-5, 5-11, 5-13 & 5-15 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 24 EFPY 24 EFPY -24 EFPY PRESSURE CURVE A -CURVE A CURVE B CURVE B CURVE (PSIG) (OF) (0F) -(OF) (OF) (0F) 470 68.0 102.0 68.0 132.0 172.0 480 68.0 102.0 '68.0 132.0 172.0 490 68.0 102.0 68.0 132.0 172.0 500 68.0 102.0 68.0 132.0 172.0 510 68.0 102.0 68.0 132.0 172.0 520 68.0 102.0 68.0 132.0 172.0 530 68.0 102.0 68.0 132.0 172.0 540 68.0 102.0 68.0 132.0 172.0 550 68.0 102.0 69.5 132.0 172.0 560 68.0 102.0 71.3 132.0 172.0 570 68.0 102.0 73.0 132.0 172.0 580 68.0 102.0 74.6 132.0 172.0 590 68.0 102.0 76.2 132.0 172.0 600 68.0 102.0 77.8 132.0 172.0 610 68.0 102.0 79.3 132.0 172.0 620 68.0 102.0 80.7 132.0 172.0 630 68.0 102.0 82.1 132.0 172.0 640 68.0 102.0 83.5 132.0 172.0 650 68.0 102.0 84.8 132.0 172.0 660 68.0 102.0 86.1 132.0 172.0 670 68.0 102.0 87.4 132.0 172.0 680 68.0 102.0 88.7 132.0 172.0 690 68.0 102.0 89.9 132.0 172.0 700 68.0 102.0 91.0 132.0 172.0 710 68.0 102.0 92.2 132.0 172.0 720 68.0 102.0 93.3 132.0 172.0 B-10

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE B-2. Fermi Unit 2 Composite P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 *F/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-5, 5-11, 5-13 & 5-15 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 24 EFPY 24 EFPY 24 EFPY PRESSURE 'CURVE A CURVE A CURVE B CURVE B CURVE C 0

(PSIG) (OF) (OF) (OF) ( F) (OF) 730 68.0 102.0 94.4 132.0 172.0 740 68.0 102.0 95.5 132.0 172.0 750 68.0 102.0 96.6 132.0 172.0 760 68.0 102.0 97.6 132.0 172.0 770 68.0 102.0 98.6 132.0 172.0 780 68.0 102.0 99.6 132.0 172.0 790 68.0 102.0 100.6 132.0 172.0 800 68.0 102.0 101.5 132.0 172.0 810 68.0 102.0 102.5 132.0 172.0 820 68.0 102.0 103.4 132.0 172.0 830 68.0 102.0 104.3 132.2 172.2 840 68.0 102.0 105.2 132.6 172.6 850 68.6 102.0 106.0 132.9 172.9 860 69.6 102.0 106.9 133.3 173.3 870 70.6 102.0 107.7 133.6 173.6 880 71.5 102.0 108.6 134.0 174.0 890 72.5 102.0 109.4 134.3 174.3 900 73.4 102.0 110.2 134.7 174.7 910 74.4 102.2 111.0 135.0 175.0 920 75.3 103.2 111.7 135.4 175.4 930 76.1 104.1 112.5 135.7 175.7 940 77.0 105.1 113.3 136.0 176.0 950 77.9 106.0 114.0 136.4 176.4 960 78.7 106.9 114.7 136.7 176.7 970 79.6 107.8 115.5 137.0 177.0 980 80.4 108.7 116.2 137.4 177.4 B-11

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-2. Fermi Unit 2 Composite P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-5, 5-11, 5-13 & 5-15 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 24 EFPY 24 EFPY 24 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C 0

(PSIG) (OF) 0

( F) 0

( F) ( F) (0F) 990 81.2 109.6 116.9 137.7 177.7 1000 82.0 110.4 117.6 138.0 178.0 1010 82.7 111.3 118.2 138.3 178.3 1020 83.5 112.1 118.9 138.6 178.6 1030 84.3 112.9 119.6 139.0 179.0 1040 85.0 113.7 120.2 139.3 179.3 1050 85.7 114.5 120.9 139.6 179.6 1055 86.1 i14.9 121.2 139.7 179.7 1060 86.4 115.3 121.5 139.9 179.9 1070 87.2 116.0 122.1 140.2 180.2 1080 87.9 116.8 122.8 140.5 180.5 1090 88.6 117.5 123.4 140.8 180.8 1100 89.2 118.3 124.0 141.1 181.1 1105 89.6 118.6 124.3 141.3 181.3 1110 89.9 119.0 124.6 141.4 181.4 1120 90.6 119.7 125.2 141.7 181.7 1130 91.2 120.4 125.8 142.0 182.0 1140 91.9 121.1 126.3 142.3 182.3 1150 92.5 121.8 126.9 142.6 182.6 1160 93.1 122.4 127.5 142.9 182.9 1170 93.8 123.1 128.0 143.2 183.2 1180 94.4 123.7 128.6 143.5 183.5 1190 95.0 124.4 129.1 143.7 183.7 1200 95.6 125.0 129.7 144.0 184.0 1210 96.2 125.7 130.2 144.3 184.3 1220 96.8 126.3 130.8 144.6 184.6 B-12

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE B-2. Fermi Unit 2 Composite P-T Curve Values for 24 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 °F/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-5, 5-11, 5-13 & 5-15 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT

• LIMITING 24 EFPY 24 EFPY 24 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B. CURVE C

( -F) *(OF) ' (OF) - (OF)

(PSIG) (OF '. '

1230 97.3 126.9 131.3 144.9 184.9 1240 97.9 127.5 131.8 145.2 185.2 1250 98.5 128.1 132.3 145.4 185.4 1260 99.0 128.7 132.8 145.7 185.7 1270 99.6 129.3 133.3 146.0 186.0 1280 100.1 129.8 133.8 146.2 186.2 1290 100.7 130.4 134.3 146.5 186.5 1300 101.2 131.0 134.8 146.8 186.8 1310 101.7 131.5 135.3 147.1 187.1 1320 102.3 132.1 135.8 147.3 187.3 1330 102.8 132.6 136.2 147.6 187.6 1340 103.3 133.2 136.7 147.8 187.8 1350 103.8 133.7 137.2 148.1 188.1 1360 104.3 134.2 137.6 148.4 188.4 1370 104.8 134.8 138.1 148.6 188.6 1380 105.3 135.3 138.5 148.9

• 188.9 1390 105.8 135.8 139.0 149.1 189.1 1400 106.3 136.3 139.4 149.4 189.4 B-1 3

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-3. Fermi Unit 2 P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 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) (OF) (OF) (0F) (OF) (0F)- (OF) 0 68.0 72.0 72.0 68.0 72.0 72.0 10 68.0 72.0 72.0 68.0 72.0 72.0 20 68.0 72.0 72.0 68.0 72.0 72.0 30 68.0 72.0 72.0 68.0 72.0 72.0 40 68.0 72.0 72.0 68.0 72.0 72.0 50 68.0 72.0 72.0 68.0 72.0 72.0 60 68.0 72.0 72.0 68.0 72.0 72.0 70 - 68.0 72.0 72.0 68.0 72.0 72.0 80 68.0 72.0 72.0 68.0 72.0 72.0 90 68.0 72.0 72.0 . 68.0 72.0 72.0 100 68.0 72.0 72.0 68.0 72.0 72.0 110 68.0 72.0 72.0 68.0 72.0 72.0 120 68.0 72.0 72.0 68.0 72.0 72.0 130 68.0 72.0 72.0 68.0 72.0 72.0 140 68.0 72.0 72.0 68.0 72.0 72.0 150 68.0 72.0 72.0 68.0 72.0 72.0 160 68.0 72.0 72.0 68.0 72.0 72.0 170 68.0 72.0 72.0 68.0 72.0 72.0 180 68.0 72.0 72.0 68.0 72.9 72.0 190 68.0 72.0 72.0 68.0 75.2 72.0 200 68.0 72.0 72.0 68.0 77.3 72.0 210 68.0 72.0 72.0 68.0 79.3 72.0 220 68.0 72.0 72.0 68.0 81.3 72.0 B-14

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-3. Fermi Unit 2 P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 *F/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 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

.(0F) 0 (PSIG) ( F) (OF) (OF) (OF) (OF) 230 68.0 72.0 72.0 68.0 83.1 72.0 240 68.0 72.0 72.0 68.0 84.9 72.0 250 68.0 72.0 72.0 68.0 86.6 72.0 260 68.0 72.0 72.0 68.0 88.2 72.0 270 68.0 72.0 72.0 68.0 89.8 72.0 280 68.0 72.0 72.0 68.0 91.3 72.0 290 68.0 72.0 72.0 68.0 92.8 72.0 300 68.0 72.0 72.0 68.0 94.2 72.0 310 68.0 72.0 72.0 68.0 95.5 72.0 312.5 68.0 72.0 72.0 68.0 95.9 72.0 312.5 68.0 102.0 102.0 68.0 132.0 132.0 320 68.0 102.0 102.0 68.0 132.0 132.0 330 68.0 102.0 102.0 68.0 132.0 132.0 340 68.0 102.0 102.0 68.0 132.0 132.0 350 68.0 102.0 102.0 68.0 132.0 132.0 360 68.0 102.0 102.0 68.0 132.0 132.0 370 68.0 102.0 102.0 68.0 132.0 132.0 380 68.0 102.0 102.0 68.0 132.0 132.0 390 68.0 102.0 102.0 68.0 132.0 132.0 400 68.0 102.0 102.0 68.0 132.0 132.0 410 68.0 102.0 102.0 68.0 132.0 132.0 420 68.0 102.0 102.0 68.0 132.0 132.0 430 68.0 102.0 102.0 68.0 132.0 132.0 440 68.0 102.0 102.0 68.0 132.0 132.0 450 68.0 102.0 102.0 68.0 132.0 132.0 B-15

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-3. Fermi Unit 2 P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 'F/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 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) (Oj) (0F) (OF) (OF) (OF)- (F) 460 68.0 102.0 102.0 68.0 132.0 132.0 470 68.0 102.0 102.0 68.0 132.0 132.0 480 68.0 102.0 102.0 68.0 132.0 132.0 490 68.0 102.0 102.0 68.0 132.0 132.0 500 68.0 102.0 102.0 68.0 132.0 132.0 510 68.0 102.0 102.0 68.0 132.0 132.0 520 68.0 102.0 102.0 68.0 132.0 132.0 530 68.0 102.0 102.0 68.0 132.0 132.0 540 68.0 102.0 102.0 68.0 132.0 132.0 550 68.0 102.0 102.0 69.5 132.0 132.0 560 68.0 102.0 102.0 71.3 132.0 132.0 570 68.0 102.0 102.0 73.0 132.0 132.0 580 68.0 102.0 102.0 74.6 132.0 132.0 590 68.0 102.0 102.0 76.2 132.0 132.0 600 68.0 102.0 102.0 77.8 132.0 132.0 610 68.0 102.0 102.0 79.3 132.0 132.0 620 68.0 102.0 102.0 80.7 132.0 132.0 630 68.0 102.0 102.0 82.1 132.0 132.0 640 68.0 102.0 102.0 83.5 132.0 132.0.

650 68.0 102.0 102.0 84.8 132.0 132.0 660 68.0 102.0 102.0 86.1 132.0 132.0 670 68.0 102.0 102.0 87.4 132.0 132.0 680 68.0 102.0 102.0 88.7 132.0 132.0 690 68.0 102.0 102.0 89.9 132.0 132.0 700 68.0 102.0 102.0 91.0 132.0 132.0 B-16

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-3. Fermi Unit 2 P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 lF/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 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

.(F) (`F) (OF)

(PSIG) - (OF) (*F) (OF) 132.0 710 68.0 102.0 102.0 92.2 132.0 720 68.0 102.0 102.0 93.3 132.0 132.0 730 68.0 102.0 102.0 94.4 132.0 132.0 740 68.0 102.0 102.0 95.5 132.0 132.0 750 68.0 102.0 102.0 96.6 132.0 132.0 760 68.0 102.0 102.0 97.6 132.0 132.0 770 68.0 102.0 102.0 98.6 132.0 132.0 780 68.0 102.0 102.0 99.6 132.0 132.0 790 68.0 102.0 102.0 100.6 132.0 132.0 800 68.0 102.0 102.0 101.5 132.0 132.0 810 68.0 102.0 102.0 102.5 132.0 132.0 820 68.0 102.0 102.0 103.4 132.0 132.0 830 68.0 102.0 102.0 104.3 132.2 132.0 840 68.0 102.0 102.0 105.2 132.6 132.0 850 68.6 102.0 102.8 106.0 132.9 132.0 860 69.6 102.0 103.9 106.9 133.3 132.3 870 70.6 102.0 105.0 107.7 133.6 133.1 880 71.5 102.0 106.1 108.6 134.0 133.8 890 72.5 102.0 107.2 109.4 134.3 134.6 900 73.4 102.0 108.2 110.2 134.7 135.3 910 .74.4 102.0 109.2 111.0 135.0 136.0 920 75.3 102.0 110.2 111.7 135.4 136.7 930 76.1 102.0 111.1 112.5 135.7 137.4 940 77.0 102.5 112.1 113.3 136.0 138.1 950 77.9 103.1 113.0 114.0 136.4 138.7 B-17

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-3. Fermi Unit 2 P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 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) (oF) (*F) (9F) ( F) (*F) (*F) 960 78.7 103.7 113.9 114.7 136.7 139.4 970 79.6 104.3 114.8 115.5 137.0 140.0 980 80.4 104.9 115.7 116.2 137.4 140.7 990 81.2 105.5 116.6 116.9 137.7 141.3 1000 82.0 106.1 117.4 117.6 138.0 142.0 1010 82.7 106.7 118.3 118.2 138.3 142.6 1020 83.5 107.2 119.1 118.9 138.6 143.2 1030 84.3 107.8 119.9 119.6 139.0 143.8 1040 85.0 108.4 120.7 120.2 139.3 144.4 1050 85.7 108.9 121.5 120.9 139.6 145.0 1055 86.1 109.2 121.9 121.2 139.7 145.3 1060 86.4 109.5 122.3 121.5 139.9 145.6 1070 87.2 110.0 123.0 122.1 140.2 146.2 1080 87.9 110.5 123.8 122.8 140.5 146.7 1090 88.6 111.1 124.5 123.4 140.8 147.3 1100 89.2 111.6 125.3 124.0 141.1 147.8 1105 89.6 111.8 125.6 124.3 141.3 148.1 1110 89.9 i12.1 126.0 124.6 141.4 148.4 1120 90.6 112.6 126.7 125.2 141.7 148.9 1130 91.2 113.1 127.4 125.8 142.0 149.5 1140 91.9 113.6 128.1 126.3 142.3 150.0 1150 92.5 114.1 128.8 126.9 142.6 150.5 1160 93.1 114.6 129.4 127.5 142.9 151.1 1170 93.8 115.1 130.1 128.0 143.2 151.6 1180 94.4 115.6 130.7 128.6 143.5 152.1 B-18

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-3. Fermi Unit 2 P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 'F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-7, 5-8 & 5-10 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) .. ( F) ..(OF) (OF): (0F) 1190 95.0 116.1 131.4 129.1 143.7 152.6 1200 95.6 116.5 132.0 129.7 144.0 153.1 1210 96.2 117.0 132.7 130.2 144.3 153.6 1220 96.8 117.5 133.3 130.8 144.6 154.1 1230 97.3 117.9 133.9 131.3 144.9 154.6 1240 97.9 118.4 134.5 131.8 145.2 155.1 1250 98.5 118.8 135.1 132.3 145.4 155.6 1260 99.0 119.3 135.7 132.8 145.7. 156.0 1270 99.6 119.7 -136.3 133.3 146.0 156.5 1280 100.1 120.2 .136.8 133.8 146.2 157.0 1290 100.7 120.6 137.4 134.3 146.5 157.4 1300 101.2 121.0 138.0 134.8 146.8 157.9 1310 101.7 121.5 138.5 135.3 147.1 158.3 1320 102.3 121.9 139.1 135.8 147.3 158.8 1330 102.8 122.3 139.6 136.2 147.6 159.2 1340 103.3 122.7 140.2 136.7 147.8 159.7 1350 103.8 123.1 140.7 137.2 148.1 160.1 1360 104.3 123.6 141.2 137.6 148.4 160.5 1370 104.8 124.0 141.8 138.1 148.6 161.0 1380 105.3 124.4 142.3 138.5 148.9 161.4 1390 105.8 124.8 142.8 139.0 149.1 161.8 1400 106.3 125.2 143.3 139.4 149.4 162.2 B-19

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-4. Fermi Unit 2 Composite P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 OF/hr for Curves B & C and 20 OF/hr for Curve A for Figures 5-6, 5-12, 5-14 & 5-16 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT . HEAD BELTLINE AT LIMITING; 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A -CURVE B CURVE B CURVE C (PSIG) (`F) (OF) (OF) (OF) (OF)

• 0 68.0 72.0 68.0 72.0 72.0 10 68.0 72.0 68.0 72.0 72.0 20 68.0 72.0 68.0 72.0 72.0 30 68.0 72.0 68.0 72.0 72.0 40 68.0 72.0 68.0 72.0 72.0 50 68.0 72.0 68.0 72.0 72.0

.60 68.0 72.0 68.0 72.0 72.0 70 68.0 72.0 68.0 72.0 72.2 80 68.0 72.0 68.0 72.0 78.2 90 68.0 72.0 68.0 72.0 83.3 100 68.0 72.0 68.0 72.0 87.8 110 68.0 72.0 68.0 72.0 91.9 120 68.0 72.0 68.0 72.0 95.7 130 68.0 72.0 68.0 72.0 99.2 140 68.0 72.0 68.0 72.0 102.4 150 68.0 72.0 68.0 72.0 105.2 160 68.0 72.0 68.0 72.0 107.9 170 68.0 72.0 68.0 72.0 110.5 180 68.0 72.0 68.0 72.9 112.9 190 68.0 72.0 68.0 75.2 115.2 200 68.0 72.0 68.0 77.3 117.3 210 68.0 72.0 68.0 79.3 119.3 220 68.0 72.0 68.0 81.3 121.3 B-20

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-4. Fermi Unit 2 Composite P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 'F/hr for Curve A for Figures 5-6, 5-12, 5-14 & 5-1E BOTTOM UPPER RPV & BOTTOM; UPPER RPV&

HEAD BELTLINE AT HEAD BELTLINE AT -LIMITING 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) 0 (`F) (OF)

( F) (OF) 230 68.0 72.0 68.0 83.1 123.1 240 68.0 72.0 68.0 84.9 124.9 250 68.0 72.0 68.0 86.6 126.6 260 68.0 72.0 68.0 88.2 128.2 270 68.0 72.0 68.0 89.8 129.8 280 68.0 72.0 68.0 91.3 131.3 290 '68.0 72.0 68.0 92.8 132.8 300 68.0 72.0 68.0 94.2 134.2 310 68.0 72.0 68.0 95.5 135.5 312.5 68.0 72.0 68.0 95.9 135.9 312.5 68.0 102.0 68.0 132.0 172.0 320 68.0 102.0 68.0 132.0 172.0 330 68.0 102.0 68.0 132.0 172.0 340 68.0 102.0 68.0 132.0 172.0 350 68.0 102.0 68.0 132.0 172.0 360 68.0 102.0 68.0 132.0 172.0 370 68.0 102.0 68.0 132.0 172.0 380 68.0 102.0 68.0 132.0 172.0 390 68.0 102.0 68.0 132.0 172.0 400 68.0 102.0 68.0 132.0 172.0 410 68.0 102.0 68.0 132.0 172.0 420 68.0 102.0 68.0 132.0 172.0 430 68.0 102.0 68.0 132.0 172.0 440 68.0 102.0 68.0 132.0 172.0 450 68.0 102.0 68.0 132.0 172.0 B-21

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-4. Fermi Unit 2 Composite P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-6, 5-12, 5-14 & 5-16 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVEC 0 (OF) (OF)

(PSIG) ( F) (OF) (OF) 460 68.0 102.0 68.0 132.0 172.0 470 68.0 102.0 68.0 132.0 172.0 480 68.0 102.0 68.0 132.0 172.0 490 68.0 102.0 68.0 132.0 172.0 500 68.0 102.0 68.0 132.0 172.0 510 68.0 102.0 68.0 132.0 172.0 520 68.0 102.0 68.0 132.0 172.0 530 68.0 102.0 68.0 132.0 172.0 540 68.0 102.0 68.0 132.0 172.0 550 68.0 102.0 69.5 132.0 172.0 560 68.0 102.0 71.3 132.0 172.0 570 68.0 102.0 73.0 132.0 172.0 580 68.0 102.0 74.6 132.0 172.0 590 68.0 102.0 76.2 132.0 172.0 600 68.0 102.0 77.8 132.0 172.0 610 68.0 102.0 79.3 132.0 172.0 620 68.0 102.0 80.7 132.0 172.0 630 68.0 102.0 82.1 132.0 172.0 640 68.0 102.0 83.5 132.0 172.0 650 68.0 102.0 84.8 132.0 172.0 660 68.0 102.0 86.1 132.0 172.0 670 68.0 102.0 87.4 132.0 172.0 680 68.0 102.0 88.7 132.0 172.0 690 68.0 102.0 89.9 132.0 172.0 700 68.0 102.0 91.0 132.0 172.0 B-22

GE Nuclear Energy NEDO-33 133 Non-Proprietary Version TABLE B-4. Fermi Unit 2 Composite P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 OF/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-6, 5-12, 5-14 & 5-16 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE ATe HEAD BELTLINE AT LIMITING 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C

.. ( F .

(PSIG) (0F) (OF) (OF) (OF)  : (f) 710 68.0 102.0 92.2 132.0 172.0 720 68.0 102.0 93.3 132.0 172.0 730 68.0 102.0 94.4 132.0 172.0 740 68.0 102.0 95.5 132.0 172.0 750 68.0 102.0 96.6 132.0 172.0 760 68.0 102.0 97.6 132.0 172.0 770 68.0 102.0 98.6 132.0 172.0 780 68.0 102.0 99.6 132.0 172.0 790 68.0 102.0 100.6 132.0 172.0 800 68.0 102.0 101.5 132.0 172.0 810 68.0 102.0 102.5 132.0 172.0 820 68.0 102.0 103.4 132.0 172.0 830 68.0 102.0 104.3 132.2 172.2 840 68.0 102.0 105.2 132.6 172.6 850 68.6 102.8 106.0 132.9 172.9 860 69.6 103.9 106.9 133.3 173.3 870 70.6 105.0 107.7 133.6 173.6 880 71.5 106.1 108.6 134.0 174.0 890 72.5 107.2 109.4 134.6 174.6 900 73.4 108.2 110.2 135.3 175.3 910 74.4 109.2 111.0 136.0 176.0 920 75.3 110.2 111.7 136.7 176.7 930 76.1 111.1 112.5 137.4 177.4 940 77.0 112.1 113.3 138.1 178.1 950 77.9 113.0 114.0 138.7 178.7 B-23

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE 8-4. Fermi Unit 2 Composite P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 *F/hr for Curve A for Figures 5-6, 5-12, 5-14 & 5-16 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 32 EFPY -32 EFPY 32 EFPY PRESSURE. CURVEA CURVEA CURVE B CURVE B CURVE C (PSIG) (OF) (OF),.. (OF) (OF) (*F) 960 78.7 113.9 114.7 139.4 . 179.4 970 79.6 114.8 115.5 140.0 180.0 980 80.4 115.7 116.2 140.7 180.7 990 81.2 116.6 116.9 141.3 181.3 1000 82.0 117.4 117.6 142.0 182.0 1010 82.7 118.3 118.2 142.6 182.6 1020 83.5 119.1 118.9 143.2 183.2 1030 84.3 119.9 119.6 143.8 183.8 1040 85.0 120.7 120.2 144.4 184.4 1050 85.7 121.5 120.9 145.0 185.0 1055 86.1 121.9 121.2 145.3 185.3 1060 86.4 122.3 121.5 145.6 185.6 1070 87.2 123.0 122.1 146.2 186.2 1080 87.9 123.8 122.8 146.7 186.7 1090 88.6 124.5 123.4 147.3 187.3 1100 89.2 125.3 124.0 147.8 187.8 1105 89.6 125.6 124.3 148.1 188.1 1110 89.9 126.0 124.6 148.4 188.4 1120 90.6 126.7 125.2 148.9 188.9 1130 91.2 127.4 125.8 149.5 189.5 1140 91.9 128.1 126.3 150.0 190.0 1150 92.5 128.8 126.9 150.5 190.5 1160 93.1 129.4 127.5 151.1 191.1 1170 93.8 130.1 128.0 151.6 191.6 1180 94.4 130.7 128.6 152.1 192.1 B-24

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE B-4. Fermi Unit 2 Composite P-T Curve Values for 32 EFPY Required Metal Temperature with Required Coolant Temperature Rate at 100 0F/hr for Curves B & C and 20 0F/hr for Curve A for Figures 5-6, 5-12, 5-14 & 5-16 BOTTOM UPPER RPV& BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT - LIMITING 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C 0

(PSIG) (OF) (OF) ( F) 0

( F) (0F) 1190 95.0 131.4 129.1 152.6 192.6 1200 95.6 132.0 129.7 153.1 193.1 1210 96.2 132.7 130.2 153.6 193.6 1220 96.8 133.3 130.8 154.1 194.1 1230 97.3 133.9 131.3 154.6 194.6 1240 97.9 134.5 131.8 155.1 195.1 1250 98.5 135.1 132.3 155.6 195.6 1260 99.0 135.7 132.8 156.0 196.0 1270 99.6 136.3 133.3 156.5 196.5 1280 100.1 136.8 133.8 157.0 197.0 1290 100.7 137.4 134.3 157.4 197.4 1300 101.2 138.0 134.8 157.9 197.9 1310 101.7 138.5 135.3 158.3 198.3 1320 102.3 139.1 135.8 158.8 198.8 1330 102.8 139.6 136.2 159.2 199.2 1340 103.3 140.2 136.7 159.7 199.7 1350 103.8 140.7 137.2 160.1 200.1 1360 104.3 141.2 137.6 160.5 200.5 1370 104.8 141.8 138.1 161.0 201.0 1380 105.3 142.3 138.5 161.4 201.4 1390 105.8 142.8 139.0 161.8 201.8 1400 106.3 143.3 139.4 162.2 202.2 B-25

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear.Energy NEDO-331 33 Non-Proprietary Version C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

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

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

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

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

C-2

GE Nuclear Energy NEDO-33133 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 <200 F per hour. If the coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.

C.2.2 Curve B: Non-Nuclear HeatuplCooldown 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 200F per hour during a hydrotest and when the core is not critical.

C.2.3 Curve C: Core Critical Operation The operator must comply with this curve whenever the core is critical. An exception to this principle is for low-level physics tests; Curve B must be followed during these situations.

C.3 REACTOR OPERATION VERSUS OPERATING LIMITS For most reactor operating conditions, coolant pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the right of the P-T curves). The operations where P-T curve compliance is typically monitored closely are planned events, such as vessel bolt-up, leakage testing and startup/shutdown operations, where operator actions can directly influence vessel pressures and temperatures.

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

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

C-3

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:

• Head flange bolt-up

• Leakage test (Curve A compliance)

• Startup (coolant temperature change of less than or equal to 1000F in one hour period heatup)

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

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

C-4

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version APPENDIX D GE SIL 430 D-1

GE Nuclear Energy NEDO-33133 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 2120F for Tech steam pressure to from main steam instrument Spec 1000F/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 212 0F for Tech Must comply with SIL 251 Spec 1 0 F/hr heatup to avoid vessel stratification.

and cooldown rate.

Alternate measurement When above 212WF need to above 2120F. allow for temperature variations (up to 10-1 50F lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GE Nuclear Energy NEDO-331 33 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 1 0 F/hr correlated RHR inlet temperature cooldown rate when in coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow: Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta Ts will be indicated coolant temperature. Delta T limit is 100F for BWR/6s and 1450 F for earlier BWRs.

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

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

outside metal surface Should have drain line flow.

temperatures.

D-3

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

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

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

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

temperature limit for head bolt-up.

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

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

RPV shell outside Information only. Slow to respond to RPV surface T/Cs coolant changes. Not available on BWR/6s.

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

D-4

GE Nuclear Energy NEDO-33133 Non-Proprietary Version TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Bottom head outside 1 of 2 primary measurements Should verify that vessel surface T/Cs to comply with stratification is not Tech Spec brittle fracture present for vessel hydro.

metal temperature (see SIL No. 251).

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

during heatup.

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

D-5

GE Nuclear Energy NEDO-33133 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:

SlLs pertain only to GE BWRs. GE prepares SlLs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, ifany, of information contained in SlLs 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 BVVR 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 SlLs.

D-6

GE Nuclear Energy NEDO-33133 Non-Proprietary Version APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy E NEDO-33133 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/cm2, then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.

The following dimensions are obtained from the referenced drawings and are specified as the distance above vessel 0'":

Shell # 2 - Top of Active Fuel (TAF)* 366.3" [1,2]

Shell # 1 - Bottom of Active Fuel (BAF) 216.3" [1,2]

Shell # 2 - Top of Extended Beltline Region 374.7" [2]

Shell # 1- Bottom of Extended Beltline Region 210.5" [2]

Centerline of Recirculation Outlet Nozzle in Shell # 1 161.5" [3,4]

Top of Recirculation Outlet Nozzle N1 in Shell # 1 193.7" [3,4]

Centerline of Recirculation Inlet Nozzle N2 in Shell # 1 181.0" [3,4]

Top of Recirculation Inlet Nozzle N2 in Shell # 1 197.5" [3,4]

Centerline of 2" Instrumentation Nozzle in Shell # 2 366.0" [4,5]

From [4], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -19" below BAF and the top of the recirculation outlet nozzle is -23" below BAF). As shown in [4], the 2" Instrumentation Nozzle (MK 315-01) is contained within the core beltline region; E-2

GE Nuclear Energy NEDO-33133 Non-Proprietary Version however, this nozzle has a thickness less than 2.5" and, as noted in Table A-2, requires no fracture toughness. Therefore, if it can be shown that the peak fluence at these locations is less than 1.0e17 n/cm 2, it can be safely concluded that all nozzles and welds, other than those included in Tables 4-5 and 4-6, are outside the beltline region of the reactor vessel.

Based on the axial fluence profile, the RPV fluence at 32 EFPY drops to less than 1.0e17 n/cm2 at -6" below the BAF and at -9" above TAF [2]. The beltline region considered in the development of the P-T curves is adjusted to include the region from 210.5" to 374.7" above reactor vessel "0" for 32 EFPY.

Based on the above, it is concluded that none of the Fermi Unit 2 reactor vessel plates, nozzles, or welds, other than those included in Tables 4-5 and 4-6, are in the beltline region.

E-3

GE Nuclear Energy NEDO-33133 Non-Proprietary Version APPENDIX E

REFERENCES:

1. Letter Number PFIP-04-0207/0801.26, J. Vargas (Detroit Edison) to G. Carlisle (GE), "Retransmittal of Design Input Request Response for WIN 8 Pressure Temperature Curves", October 21, 2004.

2. Wu, T., "DTE Energy Fermi-2 Energy Center Neutron Flux Evaluation", GE-NE-0000-0031-6254-R1, Revision 1, GE Nuclear Energy, San Jose, CA, February 2005 (GE Proprietary).

3. Drawing Number E232-908, Revision 3, "Nozzle Details", Combustion Engineering, Inc., Chattanooga, Tennessee (VPF #1976-050).

4. Drawing Number E232-895, Revision 4, "General Arrangement Elevation",

Combustion Engineering, Inc., Chattanooga, Tennessee (VPF#1976-077).

5. Drawing Number E232-909, Revision 7, "Nozzle Details", Combustion Engineering, Inc., Chattanooga, Tennessee (VPF #1976-051).

E-4

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version APPENDIX F UPPER SHELF ENERGY (USE)

F-1

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy of the beltline materials. The USE must remain above 50 ft-lb at all times during plant operation, assumed here to be 32 EFPY. Calculations of 32 EFPY USE, using Regulatory Guide 1.99, Revision 2 (RG1.99) [2] methods are summarized in Table F-1.

A RG1.99 analysis is also provided for the weld materials where Position 2.2 is applied.

Surveillance capsules from the Fermi Unit 2 vessel have not been removed and tested.

Fermi Unit 2 has committed to participate in the BWRVIP Integrated Surveillance Program (ISP), and surveillance capsule data that is representative of the weld material in this vessel is available from the ISP [3]. Test results for the weld materials from [3]

were applied to the USE evaluation as seen in Table F-1.

When Fermi Unit 2 initially reported unirradiated USE values for weld heat 13253, 12008 1092 Lot 3833, insufficient test results were available and a conservative value of 62 ft-lbs was used. The 62 ft-lb value represents the lowest Charpy test result greater than 50 ft-lbs from the material certification report. The ISP has determined that the heat of material tested and designated as CE-2(WM) is the same heat of material. Sufficient unirradiated test data exists for this heat of material; the initial USE has been determined to be 119.3 ft-lbs, and this value is used in the USE evaluation for Fermi Unit 2.

The ISP test results for the weld material heat 13253, 12008 1092 Lot 3833 demonstrate that the measured % decrease in USE exceeds the RG1.99 predicted % decrease. In order to assure that the weld materials meet all RG1.99 requirements for end of license USE, Position 2.2 was applied to all beltline weld materials. Table F-2 provides the detailed information regarding the ISP test results.

Based on the results presented in Table F-1, the USE values for the Fermi Unit 2 reactor vessel beltline materials remain within the limits of RG1.99 and 10CFR50 Appendix G for 32 EFPY of operation.

F-2

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table F-I Upper Shelf Energy Evaluation for Fermi Unit 2 at 32 EFPY

. Initial 132 EFPY  % 32 EFPY

. Transverse 1/4T Decrease Transverse Material Heat or Heat/Lot Initial Longitudinal USE , USE( 11 %Cu Fluence USE12 USED3

. ._._.. (ft-4b) (fl-ib) _. (nrcm2) - . _(_-lb)

Plates:

Lower Shell .

G3706-1 C4540-2 145 94.3 0.08 4.06E+17 8 87 G3706-2 C4560-1 156 101.4 0.11 I 4.06E+171 10 91 G3706-3 C4554-1 132 85.8 0.12 1 4.06E+171 10.5 77 Lower-Intermediate Shell . .

G3703-5 C4564-1 115 74.8 0.09 l 6.70E+17 9.5 1 68 G3705-1 B8614-1 I 130 84.5 0.12 T6.70E+17 11.5 75 G3705-2 C4574-2 l 120 78.0 0.10 I 6.70E+17 10.5 70 G3705-3 C4568-2 119 77.4 1 0.12 6.70E+17 11.5 J 68 Welds:

Vertical Lower Shell Tandem 13253, 12008 2-307 A,B,C 1092 Lot 3833 N/A 119 0.26 4.06E+17 19.5 96

4) Tandem 13253, 12003261406+7 28 141 2-307A,8BC 1092 Lot3833 NMA 119 0.26 4.06E+17 32 81 Lower-Intermediate Shell 15-308AB,CD 33A277, 124 Lot 3878 N/A 94 0.32 T6.70E+17 25 71 4

15-308AB,C,D'" 33A277 124 Lot 3878 N/A 94 0.32 l 6.70E+17 36 60 Girth 1-313 10137, 0091 Lot3999 N/A 108 0.23 4.06E+17 18 89 1-313" 10137, 0091 Lot3999 N/A 108 0.23 4.06E+17l 32 l 73 Notes:

1. Transverse USE for plate materials obtained using 65% of the longitudinal USE.

2. USE Decrease obtained from RG1.99 Figure 2.

3. 32 EFPY Transverse USE = Initial Transverse USE

• 11- (% Decrease USE /100)].

4. RG1.99 Position 22 applied to the weld materials.

F-3

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Table F-2 Detailed ISP Test Results for Weld Heat 13253,12008 1092 Lot 3833 [3]

ISP Surveillance Weld USE [Heat CE-2(WM)(13253,12008)]:

%Cu = 0.21 Unirradiated USE = 119.3 ft-lb SSP Capsule E Measured USE - . 67.7 ft-lb SSP Capsule E Fluence = 1.76E+18 nlcm2 SSP Capsule G Measured USE = 70.1 ft-lb SSP Capsule G Fluence = 1.87E+18 n/cm2 SSP Capsule E Measured % Decrease = 43.3 (Charpy Curves)

SSP Capsule E RG 1.99 Predicted % Decrease = 23.6 (RG 1.99, Rev. 2, Figure 2)

SSP Capsule G Measured % Decrease 41.2 (Charpy Curves)

SSP Capsule G RG 1.99 Predicted % Decrease = 23.9 (RG 1.99. Rev. 2, Figure 2)

F-4

GE Nuclear Energy NEDO-33133 Non-Proprietary Version APPENDIX F

REFERENCES:

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

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

3. Letter Number PFIP-04-0207/0801.26, J. Vargas (Detroit Edison) to G. Carlisle (GE), "Retransmittal of Design Input Request Response for WIN 8 Pressure Temperature Curves", October 21, 2004.

F-5

GE Nuclear Energy N EDO -33133 Non-Proprietary Version APPENDIX G THICKNESS TRANSITION DISCONTINUITY EVALUATION G-1

GE Nuclear Energy NEDO -33133 Non-Proprietary Version G.1 OBJECTIVE The purpose of the following evaluation is to determine the hydrotest, heat-up/cool-down, and transient temperatures (T) for the shell thickness transition discontinuities in the beltline, the bottom head upper to lower torus, and the bottom head to lower shell, and to demonstrate that these temperatures are bounded by the appropriate P-T curves.

G.2 METHODS AND ASSUMPTIONS ANSYS finite element analyses were performed for the thickness discontinuities in the beltline and bottom head regions of the Fermi Unit 2 vessel. The purpose of this evaluation was to determine the RPV discontinuity stresses (hoop and axial) that result from the thickness transition discontinuity in the beltline region and the bottom head. The transition in the beltline is modeled as a transition from 7.125 inches minimum thickness (lower shell) to 6.125 inches minimum thickness (lower-intermediate shell) [1]. The bottom head lower torus to upper torus is modeled as a transition from 7.375 inches minimum thickness to 3.4375 inches minimum thickness, respectively [2]. Similarly, the bottom head upper torus to lower shell is modeled as a transition from 3.4375 inches minimum thickness to 7.125 inches minimum thickness, respectively [1, 2].

Four (4) load cases defined on the Fermi Unit 2 vessel thermal cycle diagram [3] were evaluated for the beltline and bottom head shell discontinuity:

1) hydrostatic test pressure at 1055 psig,

2) cool-down transient of 100OFthr, starting at 5460 F and decreasing to 700F on the inside surface wall and with an initial operating pressure of 1000 psig, and 3) a heat-up transient of 1000F/hr, starting at 70'F and increasing to 5460F on the inside surface wall and with a final operating pressure of 1000 psig. For both transient cases it was assumed that the outside RPV wall surface is insulated with a heat transfer coefficient of 0.2 BTU/hr-ft2 'F [4] and that the ambient temperature is 1000F.

These are the bounding beltline transients of those described in Region B of the Fermi Unit 2 vessel thermal cycle diagram at temperatures for which brittle fracture could occur.

G-2

GE Nuclear Energy N EDO - 33133 Non-Proprietary Version Additionally, the bottom head was analyzed for

3) (( )),and

4) ((][3]-

As discussed in Section 4.3.2.1.2 of this report, these transients represent ((

The Normal/Upset transient "Loss of AC Power Natural Circulation Restart" was also analyzed.

It was determined that the (( )) transients bound this operating condition for the bottom head region; results for the bounding conditions are presented in this appendix.

Material properties were used from the Code of construction for the RPV Materials: Shell and Bottom Head Plate Materials are ASME SA533, Grade B, Class 1 low alloy steel (LAS) and Support Skirt Materials are ASME SA508 Class 2 [5].

Methods consistent with those described in Section 4.3 were used to calculate the T - RTNDT for the shell discontinuity for a hydrotest pressure of 1055 psig and the two transient cases. The adjusted reference temperature values shown in Table 4-6 were added to the T - RTNDT to determine the temperature "T". The value of 'T" was compared to that of the beltline region for the same condition as described in Sections 4.3.2.2.1 for the hydrotest pressure case and 4.3.2.2.4 for the transient cases.

The Control Rod Drive Penetrations in the bottom head were not evaluated as a part of this discontinuity analysis; detailed analysis of the penetrations is provided in Appendix H. The stub tubes provide sufficient stiffness that the deletion' of these penetrations from this analysis is acceptable.

It is demonstrated in this analysis that Curve A for the bottom head (CRD) and beltline regions (Figures 5-1 and 5-4) bound the temperatures found for the hydrostatic test pressure temperatures from the FEA analysis. It is also shown that Curve B for the bottom head (CRD) beltline regions (Figures 5-7 and 5-10) bound the temperatures found for transient pressures from the stresses obtained in the FEA analysis. Therefore, the transition discontinuity stresses G-3

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version in the beltline, bottom head to lower shell, and bottom head upper to lower torus are bounded by the P-T curves.

The locations of maximum stress were evaluated in the beltline shell, bottom head to lower shell, and bottom head torus locations as shown in Figure G-1.

The methods of ASME Code Section Xl, Appendix G [6] are used to calculate the pressure test and thermal limits. The membrane and bending stress were determined from the finite element analysis and are shown below. The hoop stresses were more limiting than the axial stresses, and are provided in Tables G-1 through G-5 of this appendix.

The stress intensity factors, Kim and Kib, are calculated using 1998 ASME Code with Addenda through 2000 Section XI Appendix A [7] and Appendix G [6], as shown in Section 4.3.2.2.2 of this report. Therefore, Kim= Mm*Om and Klb = Mb*Gb. The values of Mm and Mb were determined from the ASME Code Appendix G [6]. The stress intensity is based on a 1/4 T radial flaw with a six-to-one aspect ratio (length of 1.5T). The flaw is oriented normal to the maximum stress direction, in this case a vertically oriented flaw since the hoop stress was limiting.

The calculated value of Kim + Kib is multiplied by a safety factor (SF) (1.5 for pressure test and 2.0 for the transient cases), per ASME Appendix G [6] for comparison with KIR, the material fracture toughness expressed as Kjc.

The relationship between Kqc and temperature relative to reference temperature (T - RTNDT) is provided in ASME Code Section Xl Appendix A [7] Paragraph A-4200, represented by the relationship (Kl units ksi-in05):

Kic = 33.2 + 20.734 exp [ 0.02 (T - RTNDT) ]; therefore, T- RTNDT = In [ (Kic-3 3.2 ) / 20.734 ] I 0.02, where Kic = SF (Kim + Kib) for the pressure test, and Kic = (SF

• Kip) + Kis for transient cases.

G-4

GE Nuclear Energy N EDO -33133 Non-Proprietary Version This relationship is derived in the Welding Research Council (WRC) Bulletin 175 [8] as the lower bound of all dynamic fracture toughness data. This relationship provides values of pressure versus temperature (from KIR and (T - RTNDT), respectively).

The RTNDT is added to the (T - RTNDT) to determine the hydrotest, heat-up, cool-down, and additional transient temperatures, AnalVsis Information:

Beltline Thin Section Thickness: t, = 6.125 inches 4(t) 2.47 inch05 Thick Section Thickness: tm, = 7.125 inches 4(t) = 2.67 inch05 Bottom Head to Lower Shell Thin Section Thickness: tmi = 3.438 inches 4(t) = 1.85 inch05 Thick Section Thickness: tnx = 7.125 inches 4(t) = 2.67 inch0 5 Bottom Head Upper Torus to Lower Torus Thin Section Thickness: tmi, = 3.438 inches

-4(t) = 1.85 inch"5 Thick Section Thickness: t.. = 7.375 inches 4(t) = 2.72 inch"5 G-5

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version Lower to Lower-Intermediate Shell: 6.125 thick I Lower Shell: 7.125- thick I

Figure G-1: Location and Wall Thickness of Evaluation Discontinuities in the Beltline and Bottom Head Regions G-6

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version Table G-1: Analysis Results for Hydrostatic Pressure Test for the Beltline Shell Discontinuity Pres.ure Primary. Primary - K Pressure Membrane Bending Mm .mm Mb =b. T-RTNDT Surfacemm MPMKb q (psig) Pm Pb 2/3 Mm Kib (0F)

(psi): (psi) ( in. ) .

1000 Inside 19860 -24 2.29 45514 1.53 -36 68.22 26.20 1000 Outside 19860 24 2.29 45514 1.53 36 68.32 26.36 1055 Inside 20952 -25 2.29 48017 1.53 -38 71.97. 31.29 1055 Outside 20952 25 2.29 48017 1.53 38 72.08 31.44 G-7

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version Table G-2: Analysis Results for Hydrostatic Pressure Test for the Bottom Head Discontinuities Primary Primary Km = T--N-

--Membrane Bending Mb = T-RTNDT Pressure Section Surface Mem b Mm Mm*Pb M= Ki (°F)

Pm Pb 2/3 Mn 0 (psi in"')

(psi) (psi) 1000 2 Inside 18000 -2363 1.85 33300 1.23 -2914 45.58 -25.79 1000 2 Outside 18000 2363 1.85 33300 1.23 2914 54.32 0.93 1055 2 Inside 18990 -2493 1.85 35132 1.23 -3075 48.09 -16.57 1055 2 Outside 18990 2493 1.85 35132 1.23 .3075 57.31 7.54 1000 3 Inside 22350 -2892 1.85 41348 1.23 -3567 56.67 6.20 1000 3 Outside 22350 2892 1.85 41348 1.23 3567 67.37 24.98 1055 3 Inside 23579 -3051 1.85 43622 1.23 -3763 59.79 12.43 1055 3 Outside 23579 3051 1.85 43622 1.23 3763 71.08 30.13 1000 5 Inside 5247 424 1.85 9707 1.23 523 15.34 -

1000 5 Outside 5247 -424 1.85 9707 1.23 -523 13.78 1055 5 Inside 5536 447 1.85 10241 1.23 551 16.19 1055 5 Outside 5536 -447 1.85 10241 1.23 -551 14.53 G-8

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version G.3 Results and Conclusions for Hydrostatic Pressure Test The results of this analysis demonstrate that Curve A remains bounding for the bottom head to lower shell and bottom head torus (Figure 5-1) and beltline shell (Figure 5-4) discontinuities.

Beltline The maximum Fermi Unit 2 plant-specific T-RTNDT calculated with the linearized stresses from the Finite Element Analysis (FEA) for the beltline thickness discontinuity is 31.44OF as shown in Table G-1. The limiting beltline weld material RTNDT (ART) at the region of the discontinuity is 770 F (see Table 4-6) at 32 EFPY, resulting in T = 108.44 0F. The limiting beltline plate RTNDT (ART) at the region of the discontinuity is 450 F (see Table 4-6) at 32 EFPY, resulting in T = 76.440 F.

At 1055 psig, representing the 32 EFPY Fermi Unit 2 hydrostatic pressure test, the T - RTNDT for the beltline region Curve A is 39.8 0F (see Section 4.3.2.2.2), and T = 116.8 0F (see Section 4.3.2.2.2).

Because the 32 EFPY beltline region hydrostatic pressure test temperature "T of 116.80 F is greater than the T = 108.440 F obtained with the FEA analysis results, the thickness discontinuity remains bounded by the beltline curve.

Similarly, the limiting beltline material RTNDT (ART) at the region of discontinuity at 24 EFPY is the beltline weld at 620 F (see Table 4-5), resulting in T = 93.44 0F. At 1055 psig, the 'T' for the 24 EFPY beltline region Curve A is 102°F. Because the 24 EFPY beltline region hydrostatic pressure test curve is greater than the "T"obtained by FEA, the thickness discontinuity remains bounded by the beltline curve.

G-9

GE Nuclear Energy NEDO -33133 Non-Proprietary Version Bottom Head to Lower Shell The maximum T - RTNDT calculated with the Finite Element Analysis results for the bottom head to lower shell region is 30.13 0F, as shown for Section 3 (see Figure G-1 for location of this section) in Table G-2. The maximum RTNDT for the bottom head to lower shell is -100F for the plates (see Table 4-1) and welds (see Table 4-3). Thus a value of T = 20.130 F is obtained from the linearized stresses obtained in the FEA analysis. From Tables B-1 and B-3, the bottom head T (appropriate for this location) used in the analysis is 86.10 F at 1055 psig. This value bounds the maximum value of T = 20.13 0F, obtained using the linearized stresses from the FEA analysis.

Bottom Head Lower Torus to Upper Torus The maximum T - RTNDT calculated with the Finite Element Analysis results for the bottom head lower torus to upper torus region is 7.540F, as shown for Sections 2 and 5 (see Figure G-1 for location of these sections) in Table G-2. The maximum RTNDT for the bottom head lower torus to upper torus is 300 F for the plates (see Table 4-1) and -500 F for the welds (see Table 4-3).

Thus a limiting value of T = 37.54DF is obtained from the linearized stresses obtained in the FEA analysis. From Tables B-1 and B-3, the bottom head T (appropriate for this location) used in the analysis is 86.1OF at 1055 psig. This value bounds the maximum value of T = 37.54 0F, obtained using the linearized stresses from the FEA analysis.

G-10

GE Nuclear Energy N EDO - 33133 Non-Proprietary Version Table G-3: Beltline Analysis and Results for Heatup and Cooldown at 1030 psig Primary Primary Secondary Secondary Membrane Bending Membrane Bending Mb = KKi 5 K, Total T-RTNDT Surface Mm, Case Pm Pb Sm Sb . 2/3 MmI (psi in') (psi in"n). (psi in"2) (F)

(psi) (psi) (psi) (psi) .

Heatup Inside 20462 -25 -163 -6469 2.29 1.53 46819 -10248 83391 44.20 Heatup Outside 20462 25 -163 6469 2.29 1.53 46895 9504 103294 60.90 Cooldown Inside 20462 -25 1 9785 2.29 1.53 46819 14940 108579 64.54 Cooldown Outside 20462 25 1 -9633 2.29 1.53 46895 -14705 79086 39.72 G-11

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version Table G-4: Bottom Head Analysis and Results for Heatup and Cooldown at 1030 psig Primary Primary Secondary Secondary Membrane Bending Membrane Bending Mb = Ki Ks K, Total T-RTNDT Case- Location Surface Pm PbS Sb Mm 2/3 Mm (psi in1 ) (psi in') in'2) 0

( F)

(psi) (psi)' (psi) (psi)

Heatup 2 Inside 18545 -2493 -5529 -4959 1.85 1.23 31234 -16345 46124 -23.63 Heatup 2 Outside 18990 2493 -5529 4959 1.85 1.23 38206 -4113 72300 31.72 Cooldown 2 Inside 18990 -2493 688 3255 1.85 1.23 32057 5287 69401 27.87 Cooldown 2 Outside 18990 2493 688 -3255 1.85 1.23 38206 -2742 73671 33.44 Heatup 3 Inside 23027 -3051 -461 -3568 1.85 1.23 38837 -5253 72422 31.87 Heatup 3 Outside 23027 3051 -461 3568 1.85 1.23 46363 3548 96274 55.63 Cooldown 3 Inside 23027 -3051 -8 9153 1.85 1.23 38837 11274 88949 49.45 Cooldown 3 Outside 23027 3051 -8 -9072 1.85 1.23 46363 -11203 81523 42.31 Heatup 5 Inside 5406 447 27020 -14010 1.85 1.23 10553 32708 53813 -0.29 Heatup 5 Outside 5406 -447 27020 14010 1.85 1.23 9450 67266 86165 46.89 Cooldown 5 Inside 5406 447 -49 -24810 1.85 1.23 10553 -30690 -9585 -

Cooldown 5 Outside 5406 -447 -49 . 24520 1.85 1.23 9450 30150 49049 -13.43 G-12

GE Nuclear Energy NEDO - 33133 Non-Proprietary Version Table G-5: Bottom Head Analysis and Results for (( )) at 1030 psig*

Primary Primary Secondary Secondary Membrane Bending Membrane Bending Mb KIP K, Total T-RTNDT Case* LocatPm Pb Sm Sb 2/3 Mm I 2) (psi in1'2) (psi in112) (F)

(psi) (psi) (psi) (psi) .

(( 2 Inside 18545 -2493 -410 -39260 1.85 1.23 31234 -49180 13289 -

(( l] 2 Outside 18990 2493 -410 38890 1.85 1.23 38206 47205 123617 73.63

(( )) 2 Inside 18990 -2493 -3241 12120 1.85 1.23 32057 8952 73066 32.69

(( )) 2 Outside 18990 2493 -3241 -12120 1.85 1.23 38206 -20944 55468 3.57

(( )) 3 Inside 23027 -3051 19 -36410 1.85 1.23 38837 -44870 32804 80.67

(( ii 3 Outside 23027 3051 19 36090 1.85 1.23 46363 44546 137273 58.15

(( )) 3 Inside 23027 -3051 -8 17740 1.85 1.23 38837 21865 99540 30.07

(( )) 3 Outside 23027 3051 -8 -17580 1.85 1.23 46363 -21696 71030 -

(( )) 5 Inside 5406 447 -23 -10350 1.85 1.23 10553 -12807 8298

(( 1] 5 Outside 5406 -447 -23 10210 1.85 1.23 9450 12551 31450 -

(( )) 5 Inside 5406 447 -55 -40660 1.85 1.23 10553 -50249 -29144 26.41

(( 1] 5 Outside 5406 -447 -55 40190 1.85 1.23 9450 49466 68365

• See Section 4.3.2.1.2 and Appendix H for more information regarding these transients.

((

))

G-13

GE Nuclear Energy NEDO-33133 Non-Proprietary Version G.4 Results and Conclusions for Transient Cases:

The results of the discontinuity analysis demonstrate that the linearized stresses in the bottom head to lower shell and bottom head torus, and beltline regions are bounded by the bottom head (CRD) Curve B, and the beltline Curve B (Figures 5-7 and 5-9, respectively).

Beltline The maximum Fermi Unit 2 plant-specific T - RTNDT calculated with the linearized stresses from the Finite Element Analysis (FEA) for the beltline thickness discontinuity is 64.540 F as shown in Table G-3. The limiting beltline weld material RTNDT (ART) at the region of the discontinuity is 770F (see Table 4-6) at 32 EFPY, resulting in T = 141.540 F.

The limiting beltline plate RTNDT (ART) at the region of the discontinuity is 450 F (see Table 4-6) at 32 EFPY, resulting in T = 109.540 F.

At 1030 psig, the 32 EFPY beltline Curve B temperature T = 143.80 F (see Table B-3).

Because the beltline region temperature 'T' of 143.80F is greater than the T = 109.540F obtained with the FEA analysis result, the thickness discontinuity remains bounded by the beltline curve.

Similarly, the limiting beltline material RTNDT (ART) at the region of discontinuity at 24 EFPY is the beltline weld at 620 F (see Table 4-5), resulting in T = 126.540 F. At 1030 psig, the "T" for the 24 EFPY beltline region Curve B is 1320F. Because the 24 EFPY beltline region Curve B is greater than the 'T" obtained by FEA, the-thickness discontinuity remains bounded by the beltline curve.

Bottom Head to Lower Shell The maximum Fermi Unit 2 plant-specific T - RTNDT for the thickness discontinuity in the bottom head to lower shell region at 1030 psig is 80.67°F as shown for Section 3 (see G-14

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Figure G-1 for location of this section) in Tables G-4 and G-5. The maximum RTNDT for this region is -100 F for the plates (see Table 4-1) and welds (see Table 4-3). This yields a maximum value of T = 70.670F.

From Tables B-1 and B-3, the bottom head T (appropriate for this location) used in the analysis for Curve B is 119.60F at 1030 psig. This value bounds the maximum value of T = 70.670F, obtained using the linearized stresses from the FEA analysis.

Bottom Head Lower Torus to Upper Torus The maximum Fermi Unit 2 plant-specific T - RTNDT for the thickness discontinuity in the bottom head lower to upper torus region at 1030 psig is 73.630F as shown for Sections 2 and 5 (see Figure G-1 for location of these sections) in Tables G-4 and G-5. The maximum RTNDT for this region is 300F for the plates (see Table 4-1) and -50'F for-the welds (see Table 4-3). This yields a maximum value of T= 103.630F.

From Tables B-1 and B-3, the bottom head T (appropriate for this location) used in the analysis is 119.60F at 1030 psig. This value bounds the maximum value of T = 103.63 0F, obtained using the linearized stresses from the FEA analysis.

It has been demonstrated in this analysis that Curve A for the bottom head (CRD) and beltline regions (Figures 5-1 and 5-4, respectively) bound the temperatures found for the hydrostatic test pressure temperatures from the FEA analysis. It has also been shown that Curve B for the bottom head (CRD) beltline regions (Figures 5-7 and 5-10, respectively) bound the temperatures found for the applicable transient pressures from the stresses obtained in the FEA analysis. Therefore, the transition discontinuity stresses in the beltline, bottom head to lower shell, and bottom head upper to lower torus are bounded by the P-T curves provided in Section 5 of this report.

G-15

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Appendix G

References:

1. Vessel Drawings a) Drawing Number E232-895, Rev. 4, "General Arrangement Elevation for 251" ID BWR", Combustion Engineering, Inc., Chattanooga, Tenn.

(VPF # 1976-077).

b) Drawing Number E232-901, Rev. 11, "Lower Vessel Shell Assembly -

Machining & Welding", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF# 1976-011).

c) Drawing Number E232-902, Rev. 16, "Upper Vessel Shell Assembly -

Machining & Welding", Combustion Engineering, Inc., Chattanooga, Tenn. (VPF # 1976-012).

2. Drawing Number E232-900, Rev. 5, "Bottom Head Machining and Welding",

Combustion Engineering, Chattanooga, Tenn., (VPF # 1976-014).

3. GE Drawing Number 761E246, "Reactor Vessel Thermal Cycles - Reactor Vessel", GE-APED, San Jose, CA, Revision 1 (GE Proprietary Information).

4. 'Reactor Vessel Purchase Specification Data Sheet", GE-APED, San Jose, CA, November 1971 (21A9242AC, Revision 3).

5. QA Records and RPV CMTRs for Fermi 2, GE PO Number 205-H0399, Contract Number 2667, Manufactured by Combustion Engineering, Inc.,

Chattanooga, Tenn.

6. "Fracture Toughness Criteria for Protection Against Failure", Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code, 1998 Edition with Addenda through 2000.

7. "Analysis of Flaws", Appendix A to Section Xl of the ASME Boiler and Pressure Vessel Code, 1998 Edition with Addenda through 2000.

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

Welding Research Council Bulletin 175, August 1972.

G-16

GE Nuclear Energy NEDO-33133 Non-Proprietary Version APPENDIX H CORE NOT CRITICAL CALCULATION FOR THE BOTTOM HEAD (CRD PENETRATION)

H-1

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version TABLE OF CONTENTS The following outline describes the contents of this Appendix:

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

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

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

H-2

GE Nuclear Energy .NEDO-331 33 Non-Proprietary Version H.2 Scope This Appendix describes the analytical methods used to determine the T-RTNDT value applicable for the Bottom Head Core Not Critical P-T curves. This evaluation uses new finite element fracture mechanics technology developed by the General Electric Company, which is used to augment the methods described in the ASME Boiler and Pressure Vessel Code [1]. This Appendix discusses the finite element analysis and the ASME Appendix G [1] calculations separately below.

H.3 Analysis Methods This section contains technical descriptions of the analytical methods used to perform the BWR Bottom Head fracture mechanics evaluation. The applicability of the current ASME Code, Section 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].

H.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 '/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 H-3

GE Nuclear Energy NEDO-331 33 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].

H.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 ((

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.

H-4

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

((:

• 1))

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-OF) units are:

H-5

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

))

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.

H-6

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

1. Pressure Loading -

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

31 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. ((

r 2 ))

Figure 2. Pressure Loads

2. (( ]l Thermal Transient

))

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 "F.

H-7

GE Nuclear Energy NEDO-33133-Non-Proprietary Version Figure 3. Regions To Which Thermal Loads Are Applied

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

b. Regions 2 and 3:

Time (min) h2 h3 T 0 496 413 (( 1]

[] 341 354 (( ]

496 413 (( ))

(( 496 413 (( ]

((.

Temperature Plot vs. Time (min.)

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

d. Region5:h=0.2,T=100 The peak thermal gradients Were used to compute the thermal stresses based on a uniform reference temperature of 700F.

H-8

GE Nuclear Energy NEDO-331 33 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, [l .]

4. Same as 2, (( ))

1]

The cracks considered for this analysis ((

H-9

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

[I:

11 H-10

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

((

H-11

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Stress Intensity Factor Computation

[I H-12

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Benchmarkinci I ]l Methodology H-13

GE Nuclear Energy NEDO-33133 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 H-14

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Benchmarking Of Pressure Loading Results Pressure Loading analyses ((

))

H-15

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

[I

))

H-16

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version H-17

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Thermal Transients Analysis Results For the thermal transient considered, the inner diameter of the vessel is hotter than the outer diameter; hence the I.D. cracks, [f )), 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.0. (( ))

crack.

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

[I ]1 Thermal Gradients f[ 1]

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

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version

• 1[

1]

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GE Nuclear Energy NEDO-331 33 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)

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.

((

))

H.3.3 ASME Code Appendix G Evaluation The peak stress intensities for the pressure and thermal load cases evaluated above are used as inputs 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:

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version K1 = SFp K p + SFT*Kit SF =2.0 p

SFt = 1.0

( KI 33.2 0.0 T - RTNDT 20.734 ) 0.02 Where: K! 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*in' 2 H.4 Results Review of the (( )) results above demonstrates that the OD (( ))

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

))

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Note that the pressure stress intensity has been adjusted by the factor (( )) to account for the vessel pressure at which the maximum thermal stress occurred. The finite element results summarized above were calculated using a vessel pressure ((

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

))

H.5 Conclusions For the ((f] 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 ((

H.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.

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version.

APPENDIX I FRACTURE MECHANICS EVALUATION FOR FLAW INDICATION 124 CONTAINED IN RPV LOWER-INTERMEDIATE SHELL VERTICAL WELD 15-308B 1-1

GE Nuclear Energy NEDO-33133 Non-Proprietary Version Summarv The purpose of this appendix is to provide a summary of the results of the fracture mechanics evaluation performed to qualify flaw indication number 124 on the RPV lower-intermediate shell in vertical weld 15-308B. Analyses are performed to demonstrate the impact on the beltline curves for both 24 and 32 EFPY.

The indication dimensions used in this evaluation were obtained from [1]. Based on the results of this evaluation, it is concluded that continued operation in the as-is condition is justified for the 40-year (32 EFPY) design life of the RPV, including Extended Power Uprate (EPU). As this indication occurs within the beltline region of the RPV, fluence as defined in [2] is considered in order to account for irradiation effects.

Background

Combustion Engineering Company (CE) used the Upjohn welding technique to fabricate the longitudinal and meridional seams in a number of RPVs, including the Fermi 2 RPV.

The Upjohn welding process allows fabrication of heavy plate sections without the use of positioning devices. With this welding technique, the major dimension of fabrication flaws extends through the weld thickness rather than parallel to the thickness as with other fabrication welding techniques. Other relevant details are:

• The initial inservice examination of the weld was performed in 1998. Fabrication RT was performed in 1970.

• The flaw being analyzed results from ASME Code required combination of two co-planar flaws, #123 and #124. In the 2003 ISI data, the flaw location is at 374.6 inches elevation above vessel '0', the flaw length (combined) is 2.0 inches, and the flaw depth (through-wall) is 4.24 inches.

• The initial (1998) inservice examination data was re-analyzed using the rules in effect in 2003. The area encompassed the location of the 2003 flaw. A flaw is present that corresponds with the 2003 flaw #124. The flaw elevation is 374.04 inches, the flaw length is 1.0 inch, and the flaw depth is 4.24 inches. The slight difference in elevation between the two examinations is attributable to a shift 1-2

GE Nuclear Energy NEDO-33133 Non-Proprietary Version in the weld '0' location. This is a low amplitude (<50% DAC) indication that did not require recording under the rules in effect in 1998. The indication would have been recorded and analyzed in 1998, had the current rules been in effect at that time.

• The flaw appears as an acceptable indication in the manufacturer's RT films. This confirms that it is a fabrication, rather than service-induced, flaw. Evaluation of preservice examination data is expected to provide further confirmation of this conclusion.

Indication Characterization Figure 1 shows the GERIS data sheet that gives the through-wall dimension of the subject indication, along with its width and surface proximity information. The indication length (through-wall) is 4.24 inches and the width is 2.0 inches. The indication is 0.7 inches from the ID surface. The wall thickness at this location is 6.75 inches. Both the surface distance and the thickness include a clad thickness of 0.3125 inches. For the purposes of the fracture'mechanics evaluation the indication was modeled as a planar indication with an elliptical shape and the plane of the indication perpendicular to the circumferential direction. Figure 2 shows the indication geometry used in the evaluation. The flaw characterization and surface proximity guidelines of Paragraph IWA-3300 of [3] were not used since the fracture mechanics evaluation results described later in this appendix modeled the exact flaw geometry and clearly showed that the indication is unlikely to become a surface flaw during future operation.

RPV Geometry. Material, and Loading Description The vessel inside diameter is 251 inches at the clad surface and the base material thickness is 6.75 inches minus 0.3125 inches, or 6.44 inches. The leak test/operating pressure was taken as 1030 psig [1] and the leak test temperature as 1200 F at this pressure as defined in Appendix B. The only other'loading is the thermal gradient stress resulting from 1000F/hour heatup/cooldown during plant startup/shutdown. Upset and Emergency/Faulted condition events were considered as not limiting based upon [4].

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GE Nuclear Energy NEDO-33133 Non-Proprietary Version The vessel material is SA-533, Gr B, Class 1. Since the indication is entirely located in the weld metal, the material fracture toughness properties of the weld were used in the fracture mechanics evaluation. The initial RTNDT of the weld metal was determined to be -50'F as seen in Section 4 of this report. The indication is located at the upper limit of the beltline region, and there is consequently some effect from the attenuated fluence level at this location. There is a shift in the RTNDT as a function of effective full power years (EFPY). The adjusted reference temperature (ART) at this location was determined to be -140 F based upon a fluence calculation for 32 EFPY that includes operation at EPU. Similarly, for 24 EFPY, the ART is -210 F.

Calculated Stresses Stresses are the key inputs in the fracture mechanics evaluation. The circumferential stress due to internal pressure was calculated using the strength of materials formula (a = PDi / 2t). With the internal pressure, P, as 1030 psig, vessel inside diameter, Di, as (251 + 2

• 0.3125) or 251.62 inches, and the thickness, t, as 6.44 inches, the circumferential stress, a, was calculated as 20.1 ksi.

The thermal gradient stress for the 100°F/hour cooldown rate was calculated using a finite element model and a high heat transfer coefficient value, as 9.2 ksi at the vessel ID surface and -4.3 ksi at the OD surface. This stress was resolved into a membrane and a bending stress over the crack length for the purpose of fracture mechanics evaluation.

It is conservative to consider only the cooldown rate since it produces a tensile stress at the ID surface.

A weld residual stress of 8 ksi with a cosine distribution through the thickness was assumed [5].

Fracture Mechanics Evaluation The stress intensity factors at the end of the long and short axes for an elliptical flaw contained in a plate of infinite dimension are given in standard fracture mechanics handbooks such as [6] and [7]. The formulas in [7] cover both the membrane stress and 1-4

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version bending stress loadings. However, due to the proximity of the indication to the ID surface, an increase in the calculated value of the stress intensity factor (K) is expected.

This is called the "edge effect" due to surface proximity. A finite element analysis was conducted to determine the edge effect. Figure 3 shows a general view of the finite element model in which the indication is modeled. Figure 4 shows a close-up of the area where the indication is modeled. Crack tip elements were used at the periphery of the elliptical geometry representing the indication. Based on a comparison of the calculated value of K at the end away from the edge and that near the edge, the edge effect was determined to be 1.1. This means that the theoretical values of K based on the infinite plate geometry need to be multiplied by 1.1 to account for the surface proximity.

The K values were calculated for both the pressure stress and the thermal gradient stress. The K due to weld residual stress was determined to be 5 ksi-in' 2 at the long end of the indication (location 'a' in Figure 2). K at the end of the short axis (location 'b' in Figure 2) was taken as zero since it is in the compressive region of the postulated residual stress.

The calculated values of K are shown in Table 1. A fatigue crack growth evaluation was also conducted using the largest value of calculated K (42.5 ksi-in12) and assumed 390 cycles (equivalent to 15 cycles per EFPY for the remaining 21 EFPY in the 40-year license period). The calculated value of fatigue crack growth was 0.04 inches. This projected fatigue crack growth has insignificant effect on the calculated value of K.

Therefore, the K values in Table 1 that were based on current indication dimensions were not recalculated.

Table 2 shows a comparison of calculated and allowable values of K for the leak test and the startup/shutdown conditions. It is seen that all of the calculated values of K (applied K) are less than the corresponding allowable values. The leak test condition is a more limiting case. Note that the allowable values of K are based upon location 'a' and are conservative for location 'b', which is farther from the ID surface and therefore sees a lower fluence.

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GE Nuclear Energy NEDO-33133 Non-Proprietary Version Since the values of applied K at location 'a' are less than the allowable values, it is also concluded that the subject indication will not propagate and become a surface flaw during future operation.

Summary of Results A fracture mechanics evaluation was conducted using techniques consistent with the philosophy of Section Xl of the ASME Code, The indication was modeled with length equal to 4.24 inches (along the weld thickness direction) and width equal to 2 inches.

The long end of the indication is located 0.7 inches from the ID surface, which includes a clad thickness of 0.3125 inches.

As demonstrated in Section 4.3.2.2.2 of this report, the hydrotest P-T curve is calculated, at 32 EFPY, to require a temperature of 116.80F at 1055 psig; the corresponding temperature at 1030 psig is 114.90F. For these conditions, the allowable value of K is 34.0 ksi-in" 2 and does not bound the calculated value of K, which is 35.7 ksi-in"1 .

Therefore, the beltline hydrotest P-T curve must be shifted to the right in order to increase the allowable value of K, resulting in a calculated value of Kthat is bounded by the allowable value of K. For the 32 EFPY curve, an adjustment of 50 F was applied such that Curve A requires a temperature of 119.90 F at 1030 psig; similarly for 24 EFPY, an adjustment of 130F was applied, resulting in a Curve A temperature of 112.90 F at 1030 psig.

Therefore, for the limiting leak test condition at both 24 and 32 EFPY, the highest calculated value of applied stress intensity factor, K, was 35.7 ksi-in' 2 and the allowable value of K was calculated to be 35.9 ksi-in" 2 . Note that the allowable value of K was conservatively calculated for location 'a', and compared to the calculated value of K, which is maximum at location 'b'.

Because the calculated values of K are less than the allowable values of K, the indication is not expected to become a surface indication during future operation.

Furthermore, the indication is acceptable in the as-is condition for operation over 32 EFPY/40 years including extended power uprate.

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GE Nuclear Energy NEDO-33133 Non-Proprietary Version Appendix I

References:

[1] Letter Number TMIS-04-0086/0801.26, MA Brooks (Detroit Edison) to K. Narayan (GE), "GE Design Input Request for Fermi RPV Flaw Handbook",

June 14, 2004.

[2] T. Wu', DTE Energy Fermi-2 Energy Center Neutron Flux Evaluation", GENE, San Jose, CA, (GE-NE-0000-0031-6254-Rl, Revision 1, February 2005 (GE Proprietary Information).

(3] "Rules for Inservice Inspection of Nuclear Power Plants", Section Xl of the ASME Boiler & Pressure Vessel Code, 1989 Edition.

[4] H.S. Mehta, et al, "10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels", GENE, San Jose, CA, (NEDO-32205A), February 1994 (GE Proprietary Information).

[5] White Paper on Reactor Vessel Integrity Requirements for Level A and B.

Conditions", EPRI Report No. TR-100251, January 1993.

[6] H. Tada, et al, "The Stress Analysis of Cracks Handbook", Third Edition, 2000.

[7i DR Rooke and DJ Cartwright, "Compendium of Stress Intensity Factors", HMS Office (1976).

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GE Nuclear Energy NEDO-33133 Non-Proprietary Version Table I Applied Stress Intensity Factor (K) Values K at Location 'a' K at Location 'b' Plant (ksi-in1 2) (ksi-in" 2)

Pressure Pressure Condition Weld Weld

+ Total + Total Residual Residual Thermal Thermal Leak Test* 22.6 5.0 27.6 35.7 0.0 35.7 Startup/Shutdown 30.3 5.0 35.3 41.5 0.0 41.5

• No thermal gradient stress for leak test Table 2 Comparison with Allowable Values Highest K, ART at Highst

, AR at Temperature Allowable K*

Plant Condition Applied 32/24 EFPY Tmeaue AlwbeK (ksi-in1 2) (0 F) (OF) (ksi-in'/ 2)

Leak Test 35.7 -14 /-21 120 35.9 Normal Startup/Shutdown 41.5 -14 /-21 Operating 63.3 Temperature

• Based on a safety factor of 410 1-8

NEDO-33i3 GE Nuclear Energy Non-Proprietary Version Figure 1 GERIS 2000 Indication Evaluation Data Sheet for Weld 15-308B GERIS 200 ndication EvaluationData Sheet

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Figure 2 Idealized Flaw Geometry Used in the Analysis (Schematic)

Or) ID 1-10

GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Figure 3 General View of Finite Element Model my wn c-DT:rLL CIA ZY A Used.

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GE Nuclear Energy NEDO-331 33 Non-Proprietary Version Figure 4 Close-up View of Finite Element Model ri2:CJI o~r X9 A Go 1-12