Text

Request for Changes Related to TS Section 3.4.9, Reactor Coolant System Pressure & Temperature (P/T) Limits, Section 1.0 - Appendix G
	ML030660091
Person / Time
Site:	Dresden
Issue date:	02/27/2003
From:	Simpson P; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-03-039
	Download: ML030660091 (175)
	v • d • e

GE Nuclear Energy Engineering and Technology GE-NE-0000-0002-9600-01 a General Electric Company Revision 0 175 Curtner Avenue Class I San Jose, CA 95125 February 2003 Pressure-Temperature Curves For Exelon Dresden Unit 3 Prepared by:

UL..Tilly. SAnir Tglreer Structural Mechanics and Materials Verified by: 06 Zý B.D. Frew, Principal Engineer Structural Mechanics and Materials Approved by. -Th v5'C D-B.J. Branlund. Principal Engineer Structural Mechanics and Materials

GE Nuclear Energy GE-NE-0000-0002-9600-01a

- iii -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1.0 DISCLAIMER OF RESPONSIBILITY This document was prepared by the General Electric Company (GE) and is furnished solely for the purpose or purposes stated in the transmittal letter.

No other use, direct or indirect, of the document or the information it contains is authorized. Neither GE nor any of the contributors to this document:

" Makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of any information may not infringe privately owned rights; or

"* Assumes any responsibility for liability or damage of any kind that may result from any use of such information.

Copyright, General Electric Company, 2003

- iv -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a EXECUTIVE

SUMMARY

This report provides the pressure-temperature curves (P-T curves) developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]; the P-T curves in this report represent 32 and 54 EFPY as determined for a 40- and 60-year life. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Cases N-640 and N-588, and

2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kic of Figure A-4200-1 of Appendix A in lieu of Figure G-221 0-1 in Appendix G to determine T-RTNDT. ASME Code Case N-588 allows the use of an alternative procedure for calculating the applied stress intensity factors of Appendix G for axial and circumferential welds. This report incorporates a fluence [14a, 14b] calculated in accordance with the GE Licensing Topical Report NEDC-32983P, which has been approved by the NRC in a SER [14c], and is in compliance with Regulatory Guide 1.190.

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

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

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B & C)

GE Nuclear Energy GE-NE-0000-0002-9600-01 a For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times.

The P-T curves apply for both heatup and cooldown and for both the 1/4T and 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, KIr, at 1/4T to be less than that at 3/4T for a given metal temperature.

Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 32 and 54 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beltline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beltline (at 32 and 54 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

- vi -

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE OF CONTENTS

1.0 INTRODUCTION

1 2.0 SCOPE OF THE ANALYSIS 3 3.0 ANALYSIS ASSUMPTIONS 5 4.0 ANALYSIS 6 4.1 INITIAL REFERENCE TEMPERATURE 6 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BELTLINE 13 4.3 PRESSURE-TEMPERATURE CURVE METHODOLOGY 19

5.0 CONCLUSION

S AND RECOMMENDATIONS 52

6.0 REFERENCES

69

- vii -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE OF APPENDICES APPENDIX A DESCRIPTION OF DISCONTINUITIES APPENDIX B PRESSURE-TEMPERATURE CURVE DATA TABULATION APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS APPENDIX D GE SIL 430 APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

APPENDIX G BOUNDING P-T CURVES FOR DRESDEN UNITS 2 & 3

- viii -

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE OF FIGURES FIGURE 4-1: SCHEMATIC OF THE DRESDEN UNIT 3 RPV SHOWING ARRANGEMENT OF VESSEL PLATES AND WELDS 10 FIGURE 4-2: CRD PENETRATION FRACTURE TOUGHNESS LIMITING TRANSIENTS 31 FIGURE 4-3: FEEDWATER NOZZLE FRACTURE TOUGHNESS LIMITING TRANSIENT 37 FIGURE 5-1: BOTTOM HEAD P-T CURVE FOR PRESSURE TEST [CURVE A] [20°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 55 FIGURE 5-2: UPPER VESSEL P-T CURVE FOR PRESSURE TEST [CURVE A] [2 0 °F/HR OR LESS COOLANT HEATUP/COOLDOWN] 56 FIGURE 5-3: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 32 EFPY [20 °F/HR OR LESS COOLANT HEATUP/COOLDOWN] 57 FIGURE 5-4: BELTLINE P-T CURVE FOR PRESSURE TEST [CURVE A] UP TO 54 EFPY [20 "F/HR OR LESS COOLANT HEATUP/COOLDOWN] 58 FIGURE 5-5: BOTTOM HEAD P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [1000 F/HR OR LESS COOLANT HEATUP/COOLDOWN] 59 FIGURE 5-6: UPPER VESSEL P-T CURVE FOR CORE NOT CRITICAL [CURVE B] [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 60 FIGURE 5-7: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 32 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 61 FIGURE 5-8: BELTLINE P-T CURVE FOR CORE NOT CRITICAL [CURVE B] UP TO 54 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 62 FIGURE 5-9: CORE CRITICAL P-T CURVES [CURVE C] UP TO 32 EFPY [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 63 FIGURE 5-10: CORE CRITICAL P-T CURVES [CURVE C] UP TO 54 EFPY [100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 64 FIGURE 5-11: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 32 EFPY [20°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 65 FIGURE 5-12: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 32 EFPY

[100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 66 0

FIGURE 5-13: COMPOSITE PRESSURE TEST P-T CURVES [CURVE A] UP TO 54 EFPY [2 °F/HR OR LESS COOLANT HEATUP/COOLDOWN] 67 FIGURE 5-14: COMPOSITE CORE NOT CRITICAL P-T CURVES [CURVE B] UP TO 54 EFPY

[(100°F/HR OR LESS COOLANT HEATUP/COOLDOWN] 68

- ix -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE OF TABLES TABLE 4-1: RTNTDT VALUES FOR DRESDEN UNIT 3 VESSEL MATERIALS 11 TABLE 4-2: RTNDT VALUES FOR DRESDEN UNIT 3 NOZZLE AND WELD MATERIALS 12 TABLE 4-3: DRESDEN UNIT 3 BELTLINE ART VALUES (32 EFPY) 17 TABLE 4-4: DRESDEN UNIT 3 BELTLINE ART VALUES (54 EFPY) 18 TABLE 4-5:

SUMMARY

OF THE IOCFR50 APPENDIX G REQUIREMENTS 21 TABLE 4-6: APPLICABLE BWR/3 DISCONTINUITY COMPONENTS FOR USE WITH FW (UPPER VESSEL) CURVES A & B 23 TABLE 4-7: APPLICABLE BWR/3 DISCONTINUITY COMPONENTS FOR USE WITH CRD (BOTTOM HEAD) CURVES A&B 23 TABLE 5-1: COMPOSITE AND INDIVIDUAL CURVES USED TO CONSTRUCT COMPOSITE P-T CURVES 54

GE Nuclear Energy GE-NE-0000-0002-9600-01 a

1.0 INTRODUCTION

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

Complete P-T curves were developed for 32 and 54 effective full power years (EFPY).

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

The methodology used to generate the P-T curves in this report is presented in Section 4.3 and is similar to the methodology used to generate the P-T curves in 2000 [1]. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Cases N-640 [4] and N-588 [5], and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 [6] for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kic of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. ASME Code Case N-588 allows the use of an alternative procedure for calculating the applied stress intensity factors of Appendix G for axial and circumferential welds. P-T curves are developed using geometry of the RPV shells and discontinuities, the initial RTNDT of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.

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

Adjusted Reference Temperature (ART) is the reference temperature when including irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 [7] provides the methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 32 GE Nuclear Energy GE-NE-0000-0002-9600-01 a and 54 EFPY are included in Section 4.2. The peak IDfluence values of 3.3 x 1017 n/cm 2 (32 EFPY) and 5.7 x 1017 n/cm 2 (54 EFPY) used in this report are discussed in Section 4.2.1.2. Beltline chemistry values are discussed in Section 4.2.1.1.

Comprehensive documentation of the RPV discontinuities that are considered in this report is included in Appendix A. This appendix also includes a table that documents which non-beltline discontinuity curves are used to protect the discontinuities.

Guidelines and requirements for operating and temperature monitoring are included in Appendix C. GE SIL 430, a GE service information letter regarding Reactor Pressure Vessel Temperature Monitoring is included in Appendix D. Appendix E demonstrates that all reactor vessel nozzles are outside the beltline region. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Finally, Appendix G provides a set of P-T curves that bound all requirements for both Dresden Unit 2 and Dresden Unit 3.

GE Nuclear Energy GE-NE-0000-0002-9600-01a 2.0 SCOPE OF THE ANALYSIS The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in 2000 [1]. A detailed description of the P-T curve bases is included in Section 4.3. The P-T curve methodology includes the following: 1) the incorporation of ASME Code Cases N-640 and N-588, and 2) the use of the Mm calculation in the ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of Kic of Figure A-4200-1 of Appendix A in lieu of Figure G-2210-1 in Appendix G to determine T-RTNDT. ASME Code Case N-588 allows the use of an alternative procedure for calculating the applied stress intensity factors to consider attenuation to reference flaw orientation of Appendix G for circumferential welds. This Code Case also provides an alternative procedure for calculating the applied stress intensity factor for axial welds.

Other features presented are:

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

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

The pressure-temperature (P-T) curves are established to the requirements of 1 OCFR50, Appendix G [8] to assure that brittle fracture of the reactor vessel is prevented. Part of the analysis involved in developing the P-T curves is to account for irradiation embrittlement effects in the core region, or beltline. The method used to account for irradiation embrittlement is described in Regulatory Guide 1.99, Rev. 2 [7].

In addition to beltline considerations, there are non-beltline discontinuity limits such as nozzles, penetrations, and flanges that influence the construction of P-T curves. The non-beltline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable Dresden Unit 3 vessel components. The non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [8].

Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize GE Nuclear Energy GE-NE-0000-0002-9600-01 a heating requirements prior to pressure testing. Operating and temperature monitoring requirements are found in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Appendix E demonstrates that all reactor vessel nozzles are outside the beltline region. Appendix F provides the calculation for equivalent margin analysis (EMA) for upper shelf energy (USE). Finally, Appendix G provides a set of P-T curves that bound all requirements for both Dresden Unit 2 and Dresden Unit 3.

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 3.0 ANALYSIS ASSUMPTIONS The following assumptions are made for this analysis:

For end-of-license (54 EFPY) fluence a mixed capacity factor is used to determine the EFPY for a 60-year plant life. An 80% capacity factor is based on the objective to have BWR's available for full power production 80% of the year (refueling outages, etc. -20%

of the year).

For a 60-year license an 80% capacity factor is assumed for up to 19.8 EFPY and to consider recent improvements in plant operation, a 97.5% capacity factor is used beginning at 19.8 EFPY; hence 54 EFPY is assumed to represent 60 years of operation.

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

The shutdown margin, provided in the Dresden Unit 3 Technical Specification, is calculated for a water temperature of 68°F.

The flux is calculated using a pre-EPU [14a] and a post-EPU [14b] flux, both calculated in accordance with Regulatory Guide 1.190. The pre-EPU flux is applied for 19.8 EFPY and the post-EPU flux is applied for 12.2 EFPY and 34.2 EFPY for 32 and 54 EFPY, respectively.

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 4.0 ANALYSIS 4.1 INITIAL REFERENCE TEMPERATURE 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 and are summarized as follows:

a. Test specimens shall be longitudinally oriented CVN specimens.

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

c. Pressure tests shall be conducted at a temperature at least 60OF above the qualification test temperature for the vessel materials.

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

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

b. RTNDT is defined as the higher of the dropweight NDT or 60°F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion is met.

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

10CFR50 Appendix G [8] states that for vessels constructed to a version of the ASME Code prior to the Summer 1972 Addendum, fracture toughness data and data analyses must be supplemented in an approved manner. GE developed methods for analytically GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 Dresden Unit 3 vessel per the current requirements, calculations were performed in accordance with the GE method for determining RTNDT. Example RTNDT calculations for vessel plate, weld, HAZ, and forging, and for bolting material LST are summarized in the remainder of this section.

The RTNDT values for the vessel weld materials were not calculated; these values were obtained from other sources (see Section 4.2, Tables 4-3 and 4-4).

For vessel plate material, the first step in calculating RTNDT is to establish the 50 ft-lb transverse test temperature from longitudinal test specimen data (obtained from certified material test reports, CMTRs [12]). For Dresden Unit 3 CMTRs, typically six energy values were listed at a given test temperature, corresponding to two sets of Charpy 0

tests. The lowest energy Charpy value is adjusted by adding 2 F per ft-lb energy difference from 50 ft-lb.

For example, for the Dresden Unit 3 beltline plate heat C1 290-2 in the lower intermediate shell course, the lowest Charpy energy and test temperature from the CMTRs is 45 ft-lb at 1 0°F. The estimated 50 ft-lb longitudinal test temperature is:

T50L = 10°F + [(50 - 45) ft-lb. 2°F/ft-lb] = 20'F The transition from longitudinal data to transverse data is made by adding 301F to the 50 ft-lb longitudinal test temperature; thus, for this case above, TsoT = 201F + 301F = 50°F GE Nuclear Energy GE-NE-0000-0002-9600-01 a The initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T5oT- 60 0F).

Dropweight testing to establish NDT for plate material is listed in the CMTR; the NDT for the case above is 100F. Thus, the initial RTNDT for plate heat C1290-2 is 101F.

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

For vessel forging material, such as nozzles and closure flanges, the method for establishing RTNDT is the same as for vessel plate material. For the feedwater nozzle at Dresden Unit 3 (Heat ZT2885-6), the NDT is 30°F and the lowest CVN data is 32 ft-lb at 40°F (transverse Charpy data). The corresponding value of (T50T - 60 0F) is:

(T50T - 60°F) = {[40 + (50 - 32) ft-lb" 30 F/ft-lb]) - 601F = 34 0F.

Therefore, the initial RTNDT is the greater of nil-ductility transition temperature (NDT) or (T50T- 60 0 F), which is 341F.

In the bottom head region of the vessel, the vessel plate method is applied for estimating RTNDT. For the bottom head center heat of Dresden Unit 3 (Heat CI1 73-2), the NDT is 40°F and the lowest CVN data was 33 ft-lb at 40 0 F. The corresponding value of (T5OT - 600 F) was:

(T5oT - 60 0F) = {[40 + (50 - 33) ft-lb. 2 0 F/ft-lb] + 30 0F} - 60°F = 44 0 F.

Therefore, the initial RTNDT was 44 0F.

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. Ifthe required Charpy results are not met, or are not reported, but the CVN energy reported is above 30 ft-lb, the requirements of the ASME Code Section III, Subsection NB-2300 at construction are applied, namely that the 30 ft-lb test temperature plus 60°F is the LST for the bolting materials. Charpy data for the Dresden Unit 3 closure studs did not all meet the 45 ft-lb, 25 MLE requirements at 10°F. Therefore, the LST for the bolting material is 701F. The highest RTNDT in the closure flange region is 23.1°F, for the vertical electroslag weld material.

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Thus, the higher of the LST and the RTNDT +60°F is 83.1 OF, the boltup limit in the closure flange region.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TOP HEAD TOP HEAD FLANGE SHELL FLANGE

S 7 SHELL#4

,F, SHELL *3 TOP OF

. /, SHELL #2 ACTIVE FUEL (TAF) 30.3 AIAL WELDS (ES")

GIRTH WELD)

BOTTOM OF SHELL #1 ACTIVE FUEL (BAAF) 216V-

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

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

Figure 4-1: Schematic of the Dresden Unit 3 RPV Showing Arrangement of Vessel Plates and Welds GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table 4-1: RTNDT Values for Dresden Unit 3 Vessel Materials COMPONENT HEAT TEST I CHARPY ENERGY (Tro60) DROP RTNDT TEMP. (FT-LB) (°F) WEIGHT (°F)

()°FI NDT _

PLATES & FORGINGS:

Top Head & Flange Dollar Plate C1 177-4 40 73 72 74 10 40 40 MK201 Top Head Torus A0458-2 10 54 60 73 -20 10 10 MK 202 C1 173-4 10 70 51 74 -20 10 10 C1177-3 10 54 69 70 -20 10 10 Top Head Flange MK209 5Pl127 10 43 7504 71.92 -6 10 10 MK48 5P1114 10 57 108 106 -20 -10 -10 Shell Courses Upper Shell C1191-1 10 50 43 55 -6 10 10 MK60 0C1191-2 10 40 49 52 0 10 10 B5144-1 10 64 51 62 -20 10 10 Upper Int. Shell B5144-2 10 65 66 40 0 10 10 MK59 C1516-1 10 39 43 49 2 10 10 B5159-1 10 83 57 65 -20 10 10 Low-Int. Shell C1290-2 10 45 60 62 -10 10 10 MK58 A0237-1 10 71 70 59 -20 10 10 B5118-1 10 66 67 66 -20 10 10 Lower Shell C1256-2 10 75 70 90 -20 -10 -10 MK57 C1182-2 10 70 61 64 -20 10 10 B5159-2 10 55 50 65 -20 0 0 Bottom Head Dollar Plate A0284-2 40 54 65 60 10 40 40 MK1 Btm Head Torus, Btm Head A0237-2 40 92 91 109 10 40 40 MK2 C1177-1 40 49 62 74 12 40 40 C1177-2 40 66 64 83 10 40 40 C1485-1* 40 Bottom Center, Btm Head C1173-2 40 49 33 47 44 40 44 MK4 C1 173-1 40 41 45 74 28 40 40

CMTR not available: 40°F RTNOT assumed per purchase specification 21AI109 NOTE: These are minimum Charpy values.

- 11 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table 4-2: RTNDT Values for Dresden Unit 3 Nozzle and Weld Materials COMPONENT HEAT TEST CHARPY ENERGY (TfoT-60) DROP RTNDT TEMP. (FT-LB) (°F) WEIGHT (°F)

(OF) NDT NOZZLES Recirc. Outlet Nozzle ZT2405-1 40 65 74.5 74.5 -20 40 40 MK8 (Transverse data)

Recirc Inlet Nozzle ZT2405-4 40 72.5 65 84 -20 40 40 MK7 (Transverse data) ZT2405-3 40 78.5 64 60 5 -20 40 40 (Longitudinal) ZT2869 40 31 37.5 39 48 30 48 Steam Outlet Nozzle ZT2405-2 40 52.5 42 49 4 40 40 MK14 (Transverse data)

Feedwater Nozzle ZT2405-5 40 52.5 61.5 70.5 -20 40 40 MK10 (Transverse data) ZT2885-6 40 32 34 36.5 34 30 34 Core Spray Nozzle ZT2869-5 40 39 36 44.5 38 30 38 MKII ZT2782 40 43 42 34 42 30 42 6"Instrumentation, Vent & ZT3043 40 102 130 117 10 40 40 CRD HSR Nozzle MK206 & 204 & 13 Jet Pump Nozzle EV8446 40 68.5 64 49 12 40 40 MK1 9 Core Diff. Press &Liq. Con C1173-2 40 49 33 47 44 40 44 Nozzle Penetration Drain Nozzle 212918 40 238 239 237 10 N/A 10 CRD Penetration A0284-2 40 54 65 60 10 40 40 Isolation Cond. Nozzle ZT2405-3 40 67 47.5 53.5 -12.5 40 40 MK15 (Transverse data)

WELDS:

Vertical Welds ESW 23.1 Girth Welds 299L44/ -5 8650 STUDS: LST Studs 6720372 10 53 35 58 70 MK61 NOTE: These are minimum Charpy values.

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 4.2 ADJUSTED REFERENCE TEMPERA TURE FOR BEL TLINE The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (Rev 2) provides the methods for determining the ART. The Rev 2 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section. An evaluation of ART for all beltline plates and several beltline welds was performed and summarized in Tables 4-3 and 4-4 for 32 and 54 EFPY, respectively.

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

SHIFT = ARTNDT + Margin where, ARTNDT = [CF]*f (0.28-0 10 log f)

Margin = 2(a?2 + Ga)0.5 CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = 1/T fluence / 1019 2 05 Margin = 2(0I2 + oA )

c = standard deviation on initial RTNDT, which is taken to be 0°F (13*F for electroslag welds and 20°F for SAW girth welds).

Ga = standard deviation on ARTNDT, 28°F for welds and 17°F for base material, except that 0 A need not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT The margin term a, has constant values in Rev 2 of 171F for plate and 281F for weld.

However, cA need not be greater than 0.5

ARTNDT. Since the GE/BWROG method of GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 01 is taken to be 0°F for the vessel plate and most weld materials, except that a, is 131F for the beltline electroslag weld materials [13d] and 20°F for the beltline SAW girth weld materials [13e].

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

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

Submerged Arc Welds: Copper and nickel from separate evaluation [1 3b & 13c]

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

The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of Rev 2, to determine a chemistry factor (CF) per Paragraph 1.1 of Rev 2 for welds and plates, respectively. Best estimate results are used for the beltline electroslag [1 3d] and submerged arc weld [1 3e] materials for the initial RTNDT; therefore, the standard deviation (a,) is specified.

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

The bounding peak fast flux for the RPV inner surface from Reference 14 is 3.12e8 n/cm 2-s for pre-EPU and 3.46e8 n/cm 2-s for EPU conditions. The calculated fast flux at the representative 350 capsule center is 3.36e8 n/cm 2-s [14a], including the same bias adjustment as that applied to the RPV. The flux wire measurement for the Dresden Unit 3 capsule removed during the 1982 refueling outage at 5.98 EFPY is 3.76e8 n/cm 2-s [22], resulting in a calculation-to-measurement ratio of 0.89.

32 EFPY Fluence GE Nuclear Energy GE-NE-0000-0002-9600-01 a Dresden Unit 3 began EPU operation at 19.8 EFPY, thereby operating for 12.2 EFPY at EPU conditions for 32 EFPY. The RPV ID surface fluence for 32 EFPY is calculated as follows:

3.12e8 n/cm2 -s*1.01e9 s*(19.8/32) + 3.46e8 n/cm 2 -s*l1.01e9 s*(12.2/32)= 3.3e17 n/cm 2.

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

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

using the Dresden Unit 3 plant specific fluence and vessel thickness of 6.125". The 32 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

3.3e17 n/cm 2

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

The 32 EFPY 114T fluence for the lower shell plate and axial welds and the lower to lower intermediate girth weld is:

2.4e17 n/cm 2

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

54 EFPY Fluence As stated above, Dresden Unit 3 began EPU operation at 19.8 EFPY, thereby operating for 34.2 EFPY at EPU conditions for 54 EFPY. The RPV ID surface fluence for 54 EFPY is calculated as follows:

3.12e8 n/cm 2 -s*l.7e9 s*(19.8/54) + 3.46e8 n/cm 2 -s*l.7e9 s*(34.2/54)= 5.7e17 n/cm 2.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01a The fluence at 1/4T is calculated per Equation 3 of Regulatory Guide 1.99, Revision 2 [7]

using the Dresden Unit 3 plant specific fluence and vessel thickness of 6.125". The 54 EFPY 1/4T fluence for the lower-intermediate shell plate and axial welds is:

5.7e17 n/cm 2

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

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

4.1e17 n/cm 2

exp(-0.24 * (6.125/4)) = 2.9e17 n/cm2.

4.2.2 Limiting Beltline Material The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, Rev 2 was applied to compute ART.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table 4-3: Dresden Unit 3 Beltline ART Values (32 EFPY)

Lower-Intermedlate Plate and Vertical Welds Thickness- 613 inches 32 EFPY Peak I D 1l1en-e - 3 3E+17 n/tm^2 32 EFPY Peak 1/4 T Ilenee - 2.30+17 o/cori2 32 EFPY Peak 1/4 T tl'eoce - 2.3E+ 17 /c=n^2 Lower Plate and Vertical Welds and Chili Weld 32 EFPY Peak I D flueice 24E+17 1 n/(-12 Thickness - 6 13 inches 32 EFPY Peak 1/4Tlec 1 6E+17 n/=m^2 32 EFPY Peak /4 T fluenc 16E 17 n/=m^2 Initial 1/4T 32 EFPY Oj CFA 32 EFPY 32 EFPY COMPONENT HEAT OR HEAT/LOT */.C. '/%N3 CF RTndt -wleoce A RTndt Margin Shift ART

F n/cf,2 -F IF 'F -F PLATES Lower 6.111-2 C1256-2 oIl 050 73 -10 16E+1? 11 0 6 i1 22 12 6-111-6 B5159-2 0 24 047 153 0 16E+17 23 0 12 23 46 46 C1182-2 022 050 148 10 16E+ 17 22 0 I1 22 45 55 6.111-7 Loweo-Iotermedlate 6-111-3 A0237-1 0.23 049 151 10 2.31+17 28 0 14 28 56 66 6-111-10 B5118-1 0.22 049 146 10 2.3E+17 27 0 14 27 54 64 C1290-2 015 049 104 10 2 3E+17 19 0 10 19 39 49 6-111-11 WELDS Lower-intermediate 0.24 037 141 23 2 3E017 26 13 13 37 63 96 ES*

Lower 024 037 141 23 16E+17 21 13 II 34 S5 78 ES*

Girth Lower to Lower-Intcrmedsl SAW-- 299L44/8630 034 068 221 -5 1613+17 34 20 17 52 86 81 SLbhenstry values arebased eo datmfror BAW-2258. dated January 1996,but odjusted.Values of initil RTnmtand ol areobtained from the utmedocument.

' iChenstry values or based oo data from BAW-2325, dated May 1998 and lIntial RTndt and ol ameobtained from the BAW-1803-1, dated May 1991 GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table 4-4: Dresden Unit 3 Beltline ART Values (54 EFPY)

L.wer-at ermediate Plate and Vertical Welds Thickness- 613 inches 54 EFPY Peak LD fluce - 5 7E+17 u/cm^2 54 EFPY Peak 1/4 T fluence - 3 9E+17 =cm^2 54 EFPY Peak 1/4 T floenace - 3 9E+17 n/cm^2 Lower Plate and Vertical Welds and Girth Weld Thicknes- 613 inches 54 EFPY PeakI D flucnce- 41 E+17 ntcm^2 54 EFPY Peak 1/4 T fluence 2 9E+I7 vcm^2 54 EFPY Peak 14 T flucnce- 2 9E+17 n/cm^2 Initial 1/4T 54 EFPY Cl CFa 54 EFPY 54 EFPY COMPONENT HEAT OR HEAT/LOT -/Cu %N, CF RTndt Flacice A RTndi Margin Shift ART "F /cnM^2 tF F -F -F PLATES Lower 6-111-2 C1256-2 01i 050 73 -10 2.9E+17 16 0 8 16 31 21 6-111-6 B5159-2 0.24 047 153 0 2.9E+17 33 0 16 33 65 65 6-111-7 Cl1182-2 0.22 0.50 140 10 2.9C+17 32 0 16 32 63 73 Lower.Ittiintedi.ate 6-111-3 A0237-1 0.23 049 151 10 3 9E+17 39 0 17 34 73 83 6-111-10 B5118-1 0.22 049 146 10 3 9E+17 37 0 17 34 71 81 6-111-11 C1290-2 015 049 104 10 3 9E+17 27 0 13 27 53 63 WELDS LAwer-laterciedlate ES* 0.24 037 141 23 3 9E+17 36 13 I0 45 81 104 Lawe ES* 024 037 141 23 29r+17 30 13 15 40 70 93 Girth Lower to Lowe-Intermeddiate SAW-* 299L440650 034 068 221 -5 29E+17 47 20 24 62 109 104 Lhenistry values are based on data front BAW-2258, dated Jamuary1996,but adjusted. Values of InitialRIndt and ol aceobtained fron the same document.

Chcimnsiiy values are based on data from BAW-2325, dated May 1998 and Initial RTndt andol areobtained ftenethe BAW-1803-1,dated May 1991 GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 of the ASME Code [6]) forms the basis for the requirements of 10CFR50 Appendix G. The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

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

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B &C)

The closure flange region includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the neutron fluence is sufficient to cause a significant shift of RTNDT. The remaining portions of the vessel (i.e., upper vessel, lower vessel) include shells, components like the nozzles, the support skirt, and stabilizer brackets; these regions will also be called the non-beltline region.

For the core not critical and the core critical curves, the P-T curves specify a coolant 0 F/hr or less for which the curves are heatup and cooldown temperature rate of IOO applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the GE Nuclear Energy GE-NE-0000-0002-9600-01 a nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the 1/4T and 314T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 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, Kin, at 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

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

-20 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table 4-5: Summary of the 10CFR50 Appendix G Requirements Oprtng Conditionand P*ressureriio<i: Minimum Temperature Rlequirement:,l 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 + 60°F*

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 + 60°F*

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

b. Core critical - Curve C

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

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

60°F adder is included by GE as an additional conservatism as discussed in Section 4.3.2.3 There are four vessel region§ 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 GE-NE-0000-0002-9600-01 a GE PROPRIETARY INFORMATION DELETED 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0E17 n/cm 2 ) to cause any significant shift of RTNDT. Non-beltline components include nozzles (see Appendix E),

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

Detailed stress analyses of the non-beltline components were performed for the BWRJ6 specifically for the purpose of fracture toughness analysis. The BWRP6 stress analysis bounds for BWR/2 through BWR/5 designs, as will be demonstrated in the following evaluation. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include 100°F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling injections. Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [6] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T RTNDT). Plots were developed for the limiting BWRI6 components: the feedwater nozzle (FW) and the CRD penetration (bottom head). All other components in the non-beltline GE Nuclear Energy GE-NE-0000-0002-9600-01 a regions are categorized under one of these two components as described in Tables 4-6 and 4-7.

Table 4-6: Applicable BWRP3 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification FW Nozzle CRD HYD System Return Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Main Closure Flange Support Skirt Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle Steam Water Interface Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head (B only)

Top Head Nozzles (B only)

Recirculation Outlet Nozzle (B only)

Table 4-7: Applicable BWR/3 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B Discontinuity Identification CRD and Bottom Head Top Head Nozzles Recirculation Outlet Nozzle Shell"*

Support Skirt**

Shroud Support Attachments**

Core AP and Liquid Control Nozzle**

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 Dresden Unit 3 as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, as determined in Section 4.3.2.1.1 through Section 4.3.2.1.4. The generic value was adapted to the conditions at Dresden Unit 3 by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline.

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

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

In a finite element analysis [ ], the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K1. The evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section X1 Appendix G [6] and shown below. The results of that computation were K,= 143.6 ksi-in'2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 84°F.

-24 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a I

0 The limit for the coolant temperature change rate is 2 OFlhr 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, t'/2 = 2.83. The resulting value obtained was:

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

Kim = Mm cpm = ksi-in" 2 KIb = (2/3) Mm"-Opb = ksi-in" 2 The total K4 is therefore:

K, = 1.5 (Kim+ KIb) + Mm" (crsm + (2/3) "Cysb) = 143.6 ksi-inl/2 GE Nuclear Energy GE-NE-0000-0002-9600-01 a This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNDT) for a specific K,is based on the Kjc the equation of Paragraph A-4200 in ASME Appendix A [17]:

(T - RTNDT) = In [(K, - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(144 - 33.2)/20.734]/0.02 (T - RTNDT) = 84°F 2

The generic curve was generated by scaling 143.6 ksi-inl/ by the nominal pressures and calculating the associated (T - RTNDT):

Pressure Test CRD Penetration K, and (T - RTNDT) as a Function Of Pressure Nominal Pressure',,,- K, T - RTNDT S... (psig) (ksi-in'r2) (OF) 1563 144 84 1400 129 77 1200 111 66 1000 92 52 800 74 33 600 55 3 400 37 -88 The highest RTNDT for the bottom head plates and welds is 44°F, as shown in Tables 4-1 and 4-2.

-26 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Second, the P-T curve is dependent on the calculated K,value, and the K,value is proportional to the stress and the crack depth as shown below:

K, oc a (na)" (4-1)

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

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

The Dresden Unit 3 specific bottom head dimensions are R = 125.7 inches and t =8 inches minimum [19], resulting in:

Dresden Unit 3 specific: R / (t)r2 = 125.7 / (8)"2 = 44 inch1'2 (4-3)

-27 -

GE Nuclear Energy GE-NE-0000-0002-9600-0 la Since the generic value of R/(t)" 2 is larger, the generic P-T curve is conservative when applied to the Dresden Unit 3 bottom head.

4.3.2.1.2 Core Not CriticalHeatup/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.

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' .

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the KI, equation of Paragraph A-4200 in ASME Appendix A [17] for the core not critical curve:

(T - RTNDT) = In [(KI - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 102°F The generic curve was generated by scaling 192 ksi-inl/ 2 by the nominal pressures and calculating the associated (T - RTNDT):

Core Not Critical CRD Penetration K, and (T - RTNDT) as a Function of Pressure "Nominal Pressure K, T RTNDT (psig) (ksi-in!"2 )' (OF) 1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49 -14 The highest RTNDT for the bottom head plates and welds is 440F, as shown in Tables 4-1 and 4-2.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a GE Nuclear Energy GE-NE-0000-0002-9600-01 a Figure 4-2.

GE Nuclear Energy GE-NE-0000-0002-9600-01a 4.3.2.1.3 Pressure Test - Non-Beltline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, K1, for the feedwater nozzle was computed using the methods from WRC 175 [15] together with the nozzle dimension for a generic 251-inch BWR/6 feedwater nozzle. The result of that computation was K, = 200 ksi-in 12/ for an applied pressure of 1563 psig preservice hydrotest pressure.

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

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

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

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

a= 1 (tn 2 + tv 2)1/2 =2.36 inches t, = thickness of nozzle = 7.125 inches t, = thickness of vessel = 6.1875 inches rn = apparent radius of nozzle = r, + 0.29 r,=7.09 inches r, = actual inner radius of nozzle = 6.0 inches r, = nozzle radius (nozzle corner radius) = 3.75 inches Thus, air, = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an airm of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K,, is 1.5 c (na)112 - F(a/rn):

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Nominal K, = 1.5 34.97" (2r" 2.36)"2 - 1.4 = 200 ksi-in"2 The method to solve for (T - RTNDT) for a specific K,is based on the K1c equation of Paragraph A-4200 in ASME Appendix A [17] for the pressure test condition:

(T - RTNDT) = In [(KI - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(200 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 104.2°F The generic pressure test P-T curve was generated by scaling 200 ksi-in'1 2 by the nominal pressures and calculating the associated (T - RTNDT),

I GE Nuclear Energy GE-NE-0000-0002-9600-01 a

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

However, the RTNDT was increased to 44 0F to consider the stresses in the top head nozzle together with the initial RTNDT as described below. The generic pressure test P-T curve is applied to the Dresden Unit 3 feedwater nozzle curve by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 44 0F.

-34 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Second, the P-T curve is dependent on the K, value calculated. The Dresden Unit 3 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [19] and K, are shown below:

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

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

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

a y1/4 (tn 2 + tv 2 ) 1/2 =2.35 inches t, = thickness of nozzle = 7.15 inches t, = thickness of vessel = 6.125 inches rn = apparent radius of nozzle = r, + 0.29 r6=6.9 inches r, = actual inner radius of nozzle = 6.0 inches r, = nozzle radius (nozzle corner radius) = 3.0 inches Thus, airm = 2.35 / 6.9 = 0.34. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an air, of 0.34, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (7ra)2 --F(a/rn):

Nominal K, = 1.5 - 35.04 * (.:- 2.35)1/2" 1.4 = 200 ksi-inlr2 GE Nuclear Energy GE-NE-0000-0002-9600-01 a 4.3.2.1.4 Core Not CriticalHeatup/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 relatively cold feedwater flow in hotter vessel coolant.

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

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

Bulletin 175 [15].

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

Kip = SF a (na)2

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Figure 4-3.

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

The stresses used in Equation 4-4 were taken from design stress reports for the feedwater nozzle. The stresses considered are primary membrane, apm, and primary bending, apb. Secondary membrane, srm, and secondary bending, asb, stresses are included in the total K, by using ASME Appendix G [6] methods for secondary portion, Kis = Mm (asm + (2/3) Cysb) (4-5)

GE Nuclear Energy GE-NE-0000-0002-9600-01 a In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [15].

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

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

The method to solve for (T - RTNDT) for a specific K,is based on the K1c equation of Paragraph A-4200 in ASME Appendix A [17]. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

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

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

However, a thickness of 7.5 inches is not conservative for the pressure stress evaluation. Therefore, the pressure stress (apm) was adjusted for the actual vessel thickness of 6.1875 inches (i.e., 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 asm = 16.19 ksi ays = 45.0 ksi t, 6.1875 inches apb = 0.22 ksi asb = 19.04 ksi a = 2.36 inches rn = 7.09 inches t, = 7.125 inches In this case the total stress, 60.29 ksi, exceeds the yield stress, ays, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to GE Nuclear Energy GE-NE-0000-0002-9600-01 a the following equation based on the assumptions and recommendation of WRC Bulletin 175 [15]. (The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the temperature assumed for the crack root is the inside surface temperature.)

R = [cas - opm + ((atota, - oa) / 30)] / (atotal - Opm) (4-7)

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

apm = 24.84 ksi asm = 9.44 ksi Opb = 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, t1r 2 = 3.072. The resulting value obtained was:

Mm = 1.85 for -.

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

Kip = 2.0. (24.84 + 0.13). (n- 2.36)"2 1.4 Kip = 190.4 ksi-in1/2 K1s is calculated from Equation 4-5:

KIS = 2.845. (9.44 + 2/3. 11.10)

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Klý = 47.9 ksi-in" 2 The total K, is, therefore, 238.3 ksi-in" 2 .

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

(T - RTNDT) = In [(238.3- 33.2) / 20.734] / 0.02 (T - RTNDT) = 115*F The curve was generated by scaling the stresses used to determine the K1; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40°F water injected into the hot reactor vessel nozzle. In the base case that yielded a K, value of 238 ksi-in" 2, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4°F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tsaturation - 40) / (551.4 - 40). From K, the associated (T - RTNDT) can be calculated:

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Core Not Critical Feedwater Nozzle K,and (T - RTNDT) as a Function of Pressure Nominal Pressure (psig) Saturation

(°F)- Temp.

R *,i KI*

(ksi-in'2)* (T (OF)

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

Note: For each change in stress for each pressure and saturation temperature condition, there is a corresponding change to R that influences the determination of K1.

The highest non-beltline RTNDT for the feedwater nozzle at Dresden Unit 3 is 40'F as shown in Table 4-2. However, the RTNDT was increased to 44°F to consider the stresses in the top head nozzle as previously discussed. The generic curve is applied to the Dresden Unit 3 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of 44°F as discussed in Section 4.3.2.1.3.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a The stress intensity factors (KI), calculated for the beltline region according to ASME Code Appendix G procedures [6], were based on a combination of pressure and thermal stresses for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate; values were calculated for 100°F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits.

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

am = PR / tm*n (4-8)

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

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

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

for the pressure test condition:

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

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

-42 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a GE's current practice for the pressure test curve is to add a stress intensity factor, Kit, for a coolant heatup/cooldown rate of 2 0 °F/hr to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatup/cooldown rate of 1 00°F/hr. The Kit calculation for a coolant heatup/cooldown rate of 1OO0 F/hr is described in Section 4.3.2.2.3 below.

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

Adjusted RTNDT = Initial RTNDT + Shift A = 23 + 63 = 86°F (Based on ART values in Table 4-3)

Vessel Height H = 823 inches Bottom of Active Fuel Height B = 216 inches Vessel Radius (to inside of clad) R = 125.7 inches Minimum Vessel Thickness (without clad) t = 6.125 inches Pressure is calculated to include hydrostatic pressure for a full vessel:

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

= 1105 + (823 - 216) 0.0361 = 1127 psig Pressure stress:

- = PR/t (4-11)

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01a Mm = 1.85 for Ift_<2 Mm = 0.926 4t for 2<t1<3.464 = 2.29 Mm = 3.21 for -I >3.464 The stress intensity factor for the pressure stress is Kim = Mm," a. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20°F/hr instead of 100°F/hr.

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

Using the Kic equation of Paragraph A-4200 in ASME Appendix A [17], KIrm = 52.96, and Kit = 2.29 for a 20°F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

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

= ln[(1.5 52.96 + 2.29 - 33.2) / 20.734] / 0.02

= 42.5°F T can be calculated by adding the adjusted RTNDT:

T = 42.5 + 86 = 128.5°F for P = 1105 psig at 32 EFPY For Dresden Unit 3, the beltline axial weld is the limiting material at 32 EFPY. However, at 54 EFPY the beltline girth weld becomes limiting by <1'F. However, because the calculated value of Kim is reduced for a girth weld due to implementation of Code Case N-588 (circumferentially oriented defect for circumferential welds), the axial weld bounds the P-T curve beltline region requirements. To demonstrate that by using Code Case N-588, the axial weld has the most limiting temperature for the P-T curves in the beltline region, the stress intensity calculations for both the axial and girth welds at 54 EFPY are presented.

Axial Weld Calculation:

GE Nuclear Energy GE-NE-0000-0002-9600-01 a The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

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

hence, t"f2 = 2.47. The resulting value obtained was:

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

Equation 4-9 can be rearranged, and 1.5 Kim substituted for K1c, to solve for (T - RTNDT).

Using the KIc equation of Paragraph A-4200 in ASME Appendix A [17], KIm = 52.96, and Ki= 2.29 for a 2 0 °F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

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

= ln[(1.5- 52.96 + 2.29 - 33.2) / 20.734] / 0.02

= 42.5°F T can be calculated by adding the adjusted RTNDT:

T = 42.5 + 104 = 146.5°F for P = 1105 psig at 54 EFPY Girth Weld Calculation:

The value of Mm for an inside circumferential postulated surface flaw from Paragraph G-2214.1 [6] was based on a thickness of 6.125 inches (the minimum thickness without cladding); hence, t12 = 2.47. The resulting value obtained was:

Mm = 1.85 for rtS2 Mm =0.926 .F" for 2<-f<3.464 1.10 GE Nuclear Energy GE-NE-0000-0002-9600-01a Mm = 3.21 for rt'>3.464 The stress intensity factor for the pressure stress is KIr = Mm

a. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20 °F/hr instead of 100°F/hr.

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

Using the K1, equation of Paragraph A-4200 in ASME Appendix A [17], Kim = 25.4, and Kit= 2.28 for a 2 0 *F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

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

= ln[(1.5 25.4 + 2.28 - 33.2) / 20.734] / 0.02

= -53OF T can be calculated by adding the adjusted RTNDT:

T = -53 + 104 = 51'F for P = 1105 psig at54 EFPY As stated above, based on the applied pressure and temperature stress intensity factors, the axial weld flaw bounds the P-T curve in the beltline region for 54 EFPY.

4.3.2.2.3 Beltline Region - Core Not CriticalHeatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section XI Appendix G [6]:

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a The pressure stress intensity factor KIm is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.

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

a 2T(x,t) / a x 2 = 1 I p (aT(x,t) / at) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and P3is the thermal diffusivity.

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

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

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

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 Mt relationships were derived in the Welding Research Council (WRC)

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

4.3.2.2.4 Calculationsfor the Beltline Region Core Not Critical Heatup/Cooldown This sample calculation is for a pressure of 1105 psig for 32 EFPY. The core not critical heatup/cooldown curve at 1105 psig uses the same Kim as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. In addition, there is a Kit term for the thermal stress.

The additional inputs used to calculate Kit are:

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

Thermal diffusivity at 550°F (most conservative value) 13= 0.354 ft2/ hr [21]

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a AT = GC 2 / 2P (4-16)

= 100- (0.526)2/ (2- 0.354) = 39°F The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.2914) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, K1t = Mt AT = 11.39, can be calculated. Klm has the same value as that calculated in Section 4.3.2.2.2.

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

(T - RTNDT) = ln[((2 - Klm + Kl) -33.2)/20.734]/0.2 (4-17)

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

= 70 OF T can be calculated by adding the adjusted RTNDT:

T=70+86=156°F forP= 1105 psig at 32 EFPY 4.3.2.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT. Similar to the evaluations performed for the bottom head and upper vessel, a BWRJ6 finite element analysis [18] was used to model the flange region. The local stresses were computed for determination of the stress intensity factor, K1. Using a 1/4T flaw size and the K1c formulation to determine T - RTNDT, for pressures above 312 psig the P-T limits for all flange regions are bounded by the 10 CFR50 Appendix G requirement of RTNDT+ 90°F (the largest T-RTNDT for the flange at 1563 psig is 73°F).

For pressures below 312 psig, the flange curve is bounded by RTNDT + 60 (the largest T RTNDT for the flange at 312 psig is 540F), therefore, instead of determining a T GE Nuclear Energy GE-NE-0000-0002-9600-01 a (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 Dresden Unit 3 at low pressures.

The approach used for Dresden Unit 3 for the bolt-up temperature was based on the conservative value of (RTNDT + 60), or the LST of the bolting materials, whichever is greater. The 60'F adder is included by GE for two reasons: 1)the pre-1971 requirements of the ASME Code Section III, Subsection NA, Appendix G included the 60°F adder, and 2) inclusion of the additional 60°F requirement above the RTNDT provides the additional assurance that a 1/4T flaw size is acceptable. As shown in Tables 4-1 and 4-2, the limiting initial RTNDT for the closure flange region is represented by the electroslag weld materials in the upper shell at 23.1 OF, and the LST of the closure studs is 70°F; therefore, the bolt-up temperature value used is 83°F. This conservatism is appropriate because bolt-up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture.

10CFR50 Appendix G, paragraph IV.A.2 [8] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 90°F) 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 68 0 F for the reason discussed below.

-50 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a The shutdown margin, provided in the Dresden Unit 3 Technical Specification, is calculated for a water temperature of 68 0 F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68°F limit, further extensive calculations would be required to justify a lower temperature. The 83°F limit for the upper vessel and beltline region and the 68°F limit for the bottom head curve apply when the head is on and tensioned and when the head is off while fuel is in the vessel. When the head is not tensioned and fuel is not in the vessel, the requirements of 10CFR50 Appendix G [8] do not apply, and there are no limits on the vessel temperatures.

4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFR50 Appendix G [8], Table 1. Table 1 of [8] requires that core critical P-T limits be 40°F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40°F for pressures above 312 psig.

Table 1 of 10CFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 600F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 83°F, based on an RTNDT of 23.1°F. In addition, above 312 psig the Curve C temperature must be at least the greater of RTNDT of the closure region + 160°F or the temperature required for the hydrostatic pressure test (Curve A at 1105 psig). The requirement of closure region RTNDT + 160°F causes a temperature shift in Curve C at 312 psig.

-51 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a

5.0 CONCLUSION

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

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

Closure flange region (Region A)

Core beltline region (Region B)

Upper vessel (Regions A & B)

Lower vessel (Regions B & C)

For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times.

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

-52 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a The following P-T curves were generated for Dresden Unit 3.

"* Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 and 54 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel and closure assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition.

"* Separate P-T curves were developed for the upper vessel, beltline (at 32 and 54 EFPY), and bottom head for the Pressure Test and Core Not Critical conditions.

"* A composite P-T curve was also generated for the Core Critical condition at 32 and 54 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.

Using the flux from Reference 14, the P-T curves are beltline limited above 1220 psig for Curve A and above 1290 psig for Curve B at 32 EFPY. At 54 EFPY, the P-T curves become beltline limited above 760 psig for Curve A and above 730 psig for Curve B.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table 5-1: Composite and Individual Curves Used To Construct Composite P-T Curves

-Figure Table Num~bers Curve Curve Description 'Numbers for for Presentation of Presentation of

______________________ te"P-T C urves 'the P-T Curves Curve A Bottom Head Limits (CRD Nozzle) Figure 5-1 Table B-1 & 3 Upper Vessel Limits (FW Nozzle) Figure 5-2 Table B-1 & 3 Beltline Limits for 32 EFPY Figure 5-3 Table B-1 Beltline Limits for 54 EFPY Figure 5-4 Table B-3 Curve B Bottom Head Limits (CRD Nozzle) Figure 5-5 Table B-1 & 3 Upper Vessel Limits (FW Nozzle) Figure 5-6 Table B-1 & 3 Beltline Limits for 32 EFPY Figure 5-7 Table B-1 Beltline Limits for 54 EFPY Figure 5-8 Table B-3 Curve C Composite Curve for 32 EFPY** Figure 5-9 Table B-2 Composite Curve for 54 EFPY** Figure 5-10 Table B-4 A &B Composite Curves for 32 EFPY Bottom Head and Composite Curve A Figure 5-11 Table B-2 for 32 EFPY*

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

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

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100

"" 1000 hINITIAL RTndt VALUE IS

.900 0 49°F FOR BOTTOM HEAD S800 ILl HEATUP/COOLDOWN RATE OF COOLANT

< 20°F/HR 0

w 600 Z

U) j W., 500 IL 3O60 o 400 w

200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (7F)

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

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

- 55 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100 U)

C.

1000 S900 0 F INITIAL RTndt VALUE IS 44°F FOR UPPER VESSEL

' 800 ILl o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 20°F/HR z

. 500 w

m 400 LU 300 f PSIGN-312 _O_

200 I FLNGE REGION 83-F -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 (0F)

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 1000 IL 900 0

uJ BELTLINE CURVE U) 800 -_ ADJUSTED AS SHOWN:

w EFPY SHIFT (°F) 32 63 o 700 L

'U HEATUPICOOLDOWN

- RATE OF COOLANT Z

< 20°F/HR

'U IL 300 200 -BELTLINE LIMITS 100 0

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

Figure 5-3: Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 0.

1000 S900 0

BELTLINE CURVE ADJUSTED AS SHOWN:

S800 EFPY SHIFT (=F)

LU 54 81 o 700 HEATUP/COOLDOWN w

S600 z

RATE OF COOLANT

< 20°F/HR S500 S400 w

300 200 -BELTLINE LIMITS 100 0

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

Figure 5-4: Beltline P-T Curve for Pressure Test [Curve A] up to 54 EFPY

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100 C.

"" 1000 w

INITIAL RTndt VALUE IS IL 900 0 49*F FOR BOTTOM HEAD w

o 700 HEATUP/COOLDOWN RATE OF COOLANT w < 100°FIHR j 500 LU n 400 LU 300 200 100 0

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

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

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

- 59 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100 1000 C. 900 INITIAL RTndt VALUE IS 0 44*F FOR UPPER VESSELI U) 800 o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 100°F/HR 600 Z

w=i 500 w

w it 300 200 - UPPER VESSEL LIMITS (Including Flange and FW 100 Nozzle Limits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 1000 BELTLINE CURVE ADJUSTED AS SHOWN:

CL EFPY SHIFT (°F) 900 32 63 LU 800 I-.

0 700 HEATUPICOOLDOWN ILu RATE OF COOLANT

-I 600 < 100°F/HR ILu m

n 500 CD 400 o

300 200

-BELTLINE LIMITS 100 0

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 S

BELTLINE CURVE S1 000 ADJUSTED AS SHOWN:

0 EFPY SHIFT (°F) 54 81 LU x

w 1800 o 700 Lu HEATUPICOOLDOWN RATE OF COOLANT S600 < 100°F/HR z

500 Lu S400 300 200

-BELTLINE LIMITS 100 0

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 INITIAL RTndt VALUES ARE 1300 23.1°F FOR BELTLINE, 44°F FOR UPPER VESSEL, 1200 AND 49°F FOR BOTTOM HEAD 1100 BELTLINE CURVE 1000 ADJUSTED AS SHOWN:

w EFPY SHIFT (°F) 32 63 I. 900 0

I C 800 HEATUP/COOLDOWN C"

RATE OF COOLANT

< 100°FIHR 0 700

600 z

"f 500

,LU w

wU 400 300 200 I

I I- BELTLINEAND NON-BELTLINE 100 , - Minimum Cnticality LIMITS STemperature 83F 0

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01a 1400 INITIAL RTndt VALUES ARE 1300 23.10 F FOR BELTLINE, 44°F FOR UPPER VESSEL, 1200 AND 49°F FOR BOTTOM HEAD 1100

,1000 - BELTLINE CURVE ADJUSTED AS SHOWN:

"w EFPY SHIFT (°F)

m. S 900 -- 54 81 0I8 9800 - -- HEATUP/COOLDOWN

"> :RATE OF COOLANT

.O S70700 - -,

< 100°F/HR S600 z

S500 IU 400 300 300 - - - - - -- - -

- BELTLINE AND r *NON-BELTLINE 100 Minimum Criticality LIMITS Temperature 83°F 0 25 50 75 100 125 150 175 200 225 250 275 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

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

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

-64 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 INITIAL RTndt VALUES ARE 1200 23.1*F FOR BELTLINE, 44°F FOR UPPER VESSEL, AND 1100 49*F FOR BOTTOM HEAD U)

C.

"1000 BELTLINE CURVES W ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

IL 900 32 63 0

w w

o 700 HEATUPICOOLDOWN W RATE OF COOLANT

< 20°FIHR W 600 z

j 500 LU n 400 LU 300

-UPPER VESSEL

-AND BELTLINE 200 LIMITS

...... BOTTOM HEAD 100 - CURVE 0

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-O1a 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1°F FOR BELTLINE, 44°F FOR UPPER VESSEL, 1100 AND 49°F FOR BOTTOM HEAD

"-0 1000 BELTLINE CURVES w ADJUSTED AS SHOWN:

Z EFPY SHIFT (°F) 5900 32 63 w

'n 800 O 700 I,-

LU C HEATUP/COOLDOWN iu RATE OF COOLANT zto 3600 < 100°F/HR I-.

500 m 400 300

-UPPER VESSEL 200 AND BELTLINE LIMITS

--...... BOTTOM HEAD 100 CURVE 0

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

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

[1 OOF/hr or less coolant heatup/cooldown]

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 INITIAL RTndt VALUES ARE 1200 I 23.1°F FOR BELTLINE, 44°F FOR UPPER VESSEL, AND 1100 49°F FOR BOTTOM HEAD C.

1000 - BELTLINE CURVES 4 ADJUSTED AS SHOWN:

. 900-EFPY SHIFT (°F) 0I 900 54 81

,_1 cn 800 Cn w

o 700 t- HEATUP/COOLDOWN

< RATE OF COOLANT

600 -< 20°F/HR z

It=

i 500 BOTrOM w HEAD D: 68"F n2 400 Lu a

300 FGUPPER VESSEL 200 . N AND BELTLINE LIMITS

BOTTOM HEAD 100 CURVE 0-"

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-O1a 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1 F FOR BELTLINE, 44°F FOR UPPER VESSEL, 1100 AND 49*F FOR BOTTOM HEAD In C.

"- 1000 w

BELTLINE CURVES z ADJUSTED AS SHOWN:

0900 EFPY SHIFT (°F) 0 I. 54 81 L~u cn 800

.L o 700 I-t HEATUP/COOLDOWN ILl RATE OF COOLANT w- 600 < 100°F/HR Z

500 LU n 400 w

300

-UPPER VESSEL 200 AND BELTLINE LIMITS

...... BOTTOM HEAD 100 CURVE I0 0

0 25 50 75 100 125 150 175 200 225 25(

MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-14: Composite Core Not Critical P-T Curves [Curve B] up to 54 EFPY 0 F/hr or less coolant heatup/cooldown]

[1 OO

-68 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a

6.0 REFERENCES

1. R. G. Carey, "Pressure-Temperature Curves for ComEd Dresden Unit 3," GE-NE, San Jose, CA, May 2000 (GE-NE-B13-02057-00-03-R1, Revision 1) (GE Proprietary).

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

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

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

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

Pressure Vessel Code, Approval Date December 12, 1997.

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

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

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

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

GE Nuclear Energy GE-NE-0000-0002-9600-01 a

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

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

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

12. QA Records & RPV CMTR's:

Dresden 3 - (QA Records & RPV CMTR's Dresden Unit 3 GE PO# 205-55579, Manufactured by B&W), "General Electric Company Atomic Power Equipment Department (APED) Quality Control - Procured Equipment, RPV QC", Barberton, Ohio, Mt. Vernon, Indiana, and Madison, Indiana.

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

b) Howell Letter to NRC dated May 28, 1998, transmitting B&WOG Report, "Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity", BAW-2325, May 1998.

c) Letter from R.M. Krich to the NRC, "Response to Request for Additional Information Regarding Reactor Pressure Vessel Integrity - Dresden Nuclear Power Station, Units 2 and 3 Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 - LaSalle County Nuclear Power Station, Units I and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos.

50-373 and 50-374 - Quad Cities Nuclear Power Station, Units I and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265", Commonwealth Edison Company, Downers Grove, IL, July 30,1998.

d) "Evaluation of RTNDT, USE and Chemical Composition of Core Region Electroslag Welds for Dresden Units 2 and 3", BAW-2258, January 1996.

e) "Correlations for Predicting the Effects of Nuclear Reactors on Linde 80 Submerged Arc Welds", BAW-1 803, Revision 1, May 1991.

-70 -

GE Nuclear Energy GE-NE-0000-0002-9600-01 a

14. a) S. Sitaraman, "Dresden and Quad Cities Neutron Flux Evaluation," GE-NE, San Jose, CA, December 2002 (GE-NE-0000-001 1-0531-RO, Revision 0)(GE Proprietary Information).

b) S.S. Wang, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T0313: RPV Flux Evaluation", GE-NE, San Jose, CA, October 2000 (GE-NE-A22-00103-29-01, Rev. 0)(GE Proprietary Information).

c) 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 XI of the ASME Boiler & Pressure Vessel Code, 1995 Edition with Addenda through 1996.

18.

19. Bottom Head and Feedwater Nozzle Dimensions: "Final Design Report for General Electric - NED Dresden IIl", Babcock & Wilcox Co., Mt. Vernon, Indiana, August 1970 (GE VPF 2252-181-1).

20.

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

GE Nuclear Energy GE-NE-0000-0002-9600-01a

22. E.B. Norris, "Dresden Nuclear Power Station Unit 3 Reactor Vessel Irradiation Surveillance Program Analysis of Capsule No. 18", Southwest Research Institute, San Antonio, TX, February 1984 (Project No. 06-7484-003).

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX A DESCRIPTION OF DISCONTINUITIES A-1

GE Nuclear Energy GE-NE-0000-0002-9600-01 a A-2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table A Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Per ASME Code Appendix G, Section G2223 (c), fracture toughness analysis to demonstrate protection against non-ductile failure is not required for portions of nozzles and appurtenances having a thickness of 2.5" or less provided the lowest service temperature is not lower than RTNDT plus 60°F. Also Inconel discontinuities require no fracture toughness evaluations. The RPV penetrations of the nozzles listed in Table A-1 and bound the RPV penetration for the nozzles listed below, therefore, no further fracture toughness evaluation is performed for these nozzles. Nozzles and appurtenances < 2.5" or made from Inconel are not included in Table A-1 and are listed below. The Top Head Lifting Lugs are also not included in Table A-1 because the loads only occur on these components when the reactor is shutdown during an outage.

Components not requiring a fracture toughness evaluation are listed below:

Nozzle or Nozzle or Appurtenance Material Reference RTNDT LST Appurtenance (OF) (OF)

Identification MK 12 2" Instrumentation < 2.5" Alloy 600 N/A N/A 1,2&7 Penetration in RPV Shell I MK 22 Drain- Penetration < 2.5" - Bottom SA105-GR 2 40 100 Head 1, 2 & 7 MK 51 - 54 Shroud Support Attachment to RPV Alloy 600 N/A N/A Wall 1, 2 & 7 Attachment to Bottom Head I _ I MK 74, 75 & Insulation Brackets - Shells and Carbon Steel N/A N/A 77-84 Bottom Head Attachment to RPV 1, 2 & 7 Shells Attachment to Dollar Plate and RPV Shells MK 101-127 Control Rod Drive Stub Tubes - Alloy 600 N/A N/A Bottom Head 2&7 Penetration in Dollar Plate Mk 139, 141 & High and Low Pressure Seal Leak Carbon Steel 142 Detection- 1 &7 Penetration - 1" * - Flange Not a pressure boundary component; therefore requires no fracture toughness evaluation.

Mk 210 Top Head Lifting Lugs (only loads at outage) 1, 2 & 7 Attachment to Torus Not a pressure boundary component; therefore requires no fracture toughness evaluation.

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

- NIA - Not applicable for this material type.

A-3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX A

REFERENCES:

1. RPV Outline orAs-Built:

"* Babcock & Wilcox Co. Drawing # 26903F, Revision 2, "General Outline",

Babcock & Wilcox Co, Mt. Vernon, Indiana, (GE-NE VPF# 2252-139-4)

Dresden Unit 3.

"* Babcock & Wilcox Co. Drawing # 26904F, Revision 3, "Outline Sections",

Babcock & Wilcox Co, Mt. Vernon, Indiana, (GE-NE VPF# 2252-140-3)

Dresden Unit 3.

2. Certified Stress Report:

"Certified Design Document for Dresden Unit 3" B&W contract No. 610-0111, GE Order No. 205-55579", Babcock & Wilcox Co, Mt. Vernon, Indiana, August, 1970, (GE-NE VPF# 2252-181-1) - Dresden Unit 3.

3. Babcock & Wilcox Co. Drawing #15181 OE, Revision 2, "Support Skirt Assy &

Details", Babcock & Wilcox Co, Mt. Vernon, Indiana, (GE-NE VPF# 2252-133-4)

Dresden Unit 3.

4. GE Drawing #104R861, Revision 5, "Reactor Assembly, Nuclear Boiler," GE-NED, San Jose, CA - Dresden Units 2 & 3.

5. Fax Transmittal of NDIT No. SEC-DB-99-163 from Bob Geier to Ray Carey, "Pressure - Temperature (P-T) Curve Limit Re-evaluation for Dresden Units 2 and 3", Commonwealth Edison Company - Dresden Nuclear Station, Morris, IL, 11/2/99.

6. Babcock & Wilcox Co. Drawing #151808E, Revision 1, "Shroud Support", Babcock

& Wilcox Co, Mt. Vernon, Indiana, (GE-NE VPF# 2252-131-03) - Dresden Unit 3.

7. QA Records & RPV CMTR's:

Dresden 3 - (QA Records & RPV CMTR's Dresden Unit 3 GE PO# 205-55579, Mfg by B&W)"General Electric Company Atomic Power Equipment Department (APED)

Quality Control - Procured Equipment, RPV QC" Project: Dresden 3, Purchase Order: 205-55579, Vendor: Babcock & Wilcox, Location: Mt. Vernon, Indiana.

A-4

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX B PRESSURE TEMPERATURE CURVE DATA TABULATION B-1

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-I. Dresden Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-3,5-5, 5-6, & 5-7 BOTTOM UPPER 32 EFPY' BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (OF) (°F) (OF) (OF) (OF) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 83.1 83.1 140 68.0 83.1 83.1 68.0 83.1 83.1 150 68.0 83.1 83.1 68.0 84.2 83.1 160 68.0 83.1 83.1 68.0 86.9 83.1 170 68.0 83.1 83.1 68.0 89.5 83.1 180 68.0 83.1 83.1 68.0 91.9 83.1 190 68.0 83.1 83.1 68.0 94.2 83.1 200 68.0 83.1 83.1 68.0 96.3 83.1 210 68.0 83.1 83.1 68.0 98.3 83.1 220 68.0 83.1 83.1 68.0 100.3 83.1 230 68.0 83.1 83.1 68.0 102.1 83.1 B-2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-1. Dresden Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-3,5-5, 5-6, & 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE CURVE A CURVE A CURVE A B CURVE B CURVE B PRESSURE (PSIG) (OF) (°F) (OF) (OF) (°F) (OF) 240 68.0 83.1 83.1 68.0 103.9 83.1 250 68.0 83.1 83.1 68.0 105.6 83.1 260 68.0 83.1 83.1 68.0 107.2 83.1 270 68.0 83.1 83.1 68.0 108.8 83.1 280 68.0 83.1 83.1 68.0 110.3 83.1 290 68.0 83.1 83.1 68.0 111.8 83.1 300 68.0 83.1 83.1 68.0 113.2 83.1 310 68.0 83.1 83.1 68.0 114.5 83.1 312.5 68.0 83.1 83.1 68.0 114.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.0 143.1 143.1 430 68.0 113.1 113.1 68.0 143.1 143.1 440 68.0 113.1 113.1 68.0 143.1 143.1 450 68.0 113.1 113.1 68.0 143.1 143.1 460 68.0 113.1 113.1 68.0 143.1 143.1 470 68.0 113.1 113.1 68.0 143.1 143.1 B-3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-1. Dresden Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-3,5-5, 5-6, & 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 480 68.0 113.1 113.1 68.0 143.1 143.1 490 68.0 113.1 113.1 68.0 143.1 143.1 500 68.0 113.1 113.1 68.0 143.1 143.1 510 68.0 113.1 113.1 68.0 143.1 143.1 520 68.0 113.1 113.1 68.2 143.1 143.1 530 68.0 113.1 113.1 70.2 143.1 143.1 540 68.0 113.1 113.1 72.1 143.1 143.1 550 68.0 113.1 113.1 73.9 143.1 143.1 560 68.0 113.1 113.1 75.7 143.1 143.1 570 68.0 113.1 113.1 77.4 143.1 143.1 580 68.0 113.1 113.1 79.0 143.1 143.1 590 68.0 113.1 113.1 80.6 143.1 143.1 600 68.0 113.1 113.1 82.2 143.1 143.1 610 68.0 113.1 113.1 83.7 143.1 143.1 620 68.0 113.1 113.1 85.1 143.1 143.1 630 68.0 113.1 113.1 86.5 143.4 143.1 640 68.0 113.1 113.1 87.9 143.8 143.1 650 68.0 113.1 113.1 89.2 144.2 143.1 660 68.0 113.1 113.1 90.5 144.7 143.1 670 68.0 113.1 113.1 91.8 145.1 143.1 680 68.0 113.1 113.1 93.1 145.5 143.1 690 68.0 113.1 113.1 94.3 145.9 143.1 700 69.2 113.1 113.1 95.4 146.3 143.1 710 70.7 113.1 113.1 96.6 146.7 143.1 720 72.1 113.1 113.1 97.7 147.1 143.1 730 73.5 113.1 113.1 98.8 147.5 143.1 B-4

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE B-1. Dresden Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-3,5-5, 5-6, & 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (°F) (OF) (OF) (OF) (OF) 740 74.8 113.1 113.1 99.9 147.9 143.1 750 76.1 113.1 113.1 101.0 148.2 143.1 760 77.4 113.1 113.1 102.0 148.6 143.1 770 78.6 113.1 113.1 103.0 149.0 143.1 780 79.8 113.1 113.1 104.0 149.4 143.1 790 81.0 113.1 113.1 105.0 149.8 143.1 800 82.2 113.1 113.1 105.9 150.1 143.1 810 83.3 113.1 113.1 106.9 150.5 143.1 820 84.4 113.4 113.1 107.8 150.9 143.1 830 85.5 114.1 113.1 108.7 151.2 143.1 840 86.5 114.8 113.1 109.6 151.6 143.1 850 87.6 115.5 113.1 110.4 151.9 143.1 860 88.6 116.2 113.1 111.3 152.3 143.1 870 89.6 116.9 113.1 112.1 152.6 143.1 880 90.5 117.6 113.1 113.0 153.0 143.1 890 91.5 118.3 113.1 113.8 153.3 143.1 900 92.4 118.9 113.1 114.6 153.7 143.2 910 93.4 119.6 113.1 115.4 154.0 143.9 920 94.3 120.2 113.1 116.1 154.4 144.6 930 95.1 120.9 114.0 116.9 154.7 145.3 940 96.0 121.5 115.0 117.7 155.0 146.0 950 96.9 122.1 115.9 118.4 155.4 146.7 960 97.7 122.7 116.8 119.1 155.7 147.3 970 98.6 123.3 117.8 119.9 156.0 148.0 980 99.4 123.9 118.7 120.6 156.4 148.6 990 100.2 124.5 119.5 121.3 156.7 149.3 B-5

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-1. Dresden Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-3,5-5, 5-6, & 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE CURVE A CURVE A CURVE A B CURVE B CURVE B PRESSURE

(°F) (OF) (OF) (OF) (OF)

(PSIG) (OF) 1000 101.0 125.1 120.4 122.0 157.0 149.9 101.7 125.7 121.2 122.6 157.3 150.6 1010 102.5 126.2 122.1 123.3 157.6 151.2 1020 103.3 126.8 122.9 124.0 158.0 151.8 1030 104.0 127.4 123.7 124.6 158.3 152.4 1040 104.7 127.9 124.5 125.3 158.6 153.0 1050 128.5 125.3 125.9 158.9 153.6 1060 105.4 129.0 126.1 126.5 159.2 154.2 1070 106.2 106.9 129.5 126.8 127.2 159.5 154.8 1080 130.1 127.6 127.8 159.8 155.3 1090 107.6 130.6 128.3 128.4 160.1 155.9 1100 108.2 130.8 128.7 128.7 160.3 156.2 1105 108.6 131.1 129.0 129.0 160.4 156.4 1110 108.9 109.6 131.6 129.7 129.6 160.7 157.0 1120 110.2 132.1 130.5 130.2 161.0 157.5 1130 110.9 132.6 131.2 130.7 161.3 158.1 1140 111.5 133.1 131.8 131.3 161.6 158.6 1150 133.6 132.5 131.9 161.9 159.1 1160 112.1 134.1 133.2 132.4 162.2 159.7 1170 112.8 134.6 133.8 133.0 162.5 160.2 1180 113.4 114.0 135.1 134.5 133.5 162.7 160.7 1190 135.5 135.1 134.1 163.0 161.2 1200 114.6 115.2 136.0 135.8 134.6 163.3 161.7 1210 136.5 136.4 135.2 163.6 162.2 1220 115.8 116.3 136.9 137.0 135.7 163.9 162.7 1230 137.4 137.6 136.2 164.2 163.2 1240 116.9 B-6

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-1. Dresden Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-3,5-5, 5-6, & 5-7 BOTTOM UPPER 32 EFPY IBOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (°F) (OF) (°F) (OF) (OF) 1250 117.5 137.8 138.2 136.7 164.4 163.7 1260 118.0 138.3 138.8 137.2 164.7 164.2 1270 118.6 138.7 139.4 137.7 165.0 164.6 1280 119.1 139.2 140.0 138.2 165.2 165.1 1290 119.7 139.6 140.6 138.7 165.5 165.6 1300 120.2 140.0 141.1 139.2 165.8 166.0 1310 120.7 140.5 141.7 139.7 166.1 166.5 1320 121.3 140.9 142.3 140.2 166.3 166.9 1330 121.8 141.3 142.8 140.6 166.6 167.4 1340 122.3 141.7 143.4 141.1 166.8 167.8 1350 122.8 142.1 143.9 141.6 167.1 168.3 1360 123.3 142.6 144.4 142.0 167.4 168.7 1370 123.8 143.0 145.0 142.5 167.6 169.1 1380 124.3 143.4 145.5 142.9 167.9 169.6 1390 124.8 143.8 146.0 143.4 168.1 170.0 1400 125.3 144.2 146.5 143.8 168.4 170.4 B-7

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-2. Dresden Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-9, 5-11 and 5-12 UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY PRESSURE CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (°F) 0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 84.0 70 68.0 83.1 68.0 83.1 91.2 80 68.0 83.1 68.0 83.1 97.2 90 68.0 83.1 68.0 83.1 102.3 100 68.0 83.1 68.0 83.1 1068 110 68.0 83.1 68.0 83.1 110.9 120 68.0 83.1 68.0 83.1 114.7 130 68.0 83.1 68.0 83.1 118.2 140 68.0 83.1 68.0 83.1 121.4 150 68.0 83.1 68.0 84.2 124.2 160 68.0 83.1 68.0 86.9 126.9 170 68.0 83.1 68.0 89.5 129.5 180 68.0 83.1 68.0 91.9 131.9 190 68.0 83.1 68.0 94.2 134.2 200 68.0 83.1 68.0 96.3 136.3 210 68.0 83.1 68.0 98.3 138.3 220 68.0 83.1 68.0 100.3 140.3 B-8

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-2. Dresden Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-9, 5-11 and 5-12 UPPER RPV & BOTTOM"' UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY PRESSURE CURVE A CURVE B CURVE B CURVE C (OF) (°F) (OF) (OF) (OF)

(PSIG) 230 68.0 83.1 68.0 102.1 142.1 240 68.0 83.1 68.0 103.9 143.9 250 68.0 83.1 68.0 105.6 145.6 260 68.0 83.1 68.0 107.2 147.2 270 68.0 83.1 68.0 108.8 148.8 280 68.0 83.1 68.0 110.3 150.3 290 68.0 83.1 68.0 111.8 151.8 300 68.0 83.1 68.0 113.2 153.2 310 68.0 83.1 68.0 114.5 154.5 312.5 68.0 83.1 68.0 114.9 154.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.0 143.1 183.1 430 68.0 113.1 68.0 143.1 183.1 440 68.0 113.1 68.0 143.1 183.1 450 68.0 113.1 68.0 143.1 183.1 B-9

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-2. Dresden Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-9, 5-11 and 5-12 UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY PRESSURE CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 460 68.0 113.1 68.0 143.1 183.1 470 68.0 113.1 68.0 143.1 183.1 480 68.0 113.1 68.0 143.1 183.1 490 68.0 113.1 68.0 143.1 183.1 500 68.0 113.1 68.0 143.1 183.1 510 68.0 113.1 68.0 143.1 183.1 520 68.0 113.1 68.2 143.1 183.1 530 68.0 113.1 70.2 143.1 183.1 540 68.0 113.1 72.1 143.1 183.1 550 68.0 113.1 73.9 143.1 183.1 560 68.0 113.1 75.7 143.1 183.1 570 68.0 113.1 77.4 143.1 183.1 580 68.0 113.1 79.0 143.1 183.1 590 68.0 113.1 80.6 143.1 183.1 600 68.0 113.1 82.2 143.1 183.1 610 68.0 113.1 83.7 143.1 183.1 620 68.0 113.1 85.1 143.1 183.1 630 68.0 113.1 86.5 143.4 183.4 640 68.0 113.1 87.9 143.8 183.8 650 68.0 113.1 89.2 144.2 184.2 660 68.0 113.1 90.5 144.7 184.7 670 68.0 113.1 91.8 145.1 185.1 680 68.0 113.1 93.1 145.5 185.5 690 68.0 113.1 94.3 145.9 185.9 700 69.2 113.1 95.4 146.3 186.3 B-I0

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE B-2. Dresden Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-9, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 710 70.7 113.1 96.6 146.7 186.7 720 72.1 113.1 97.7 147.1 187.1 730 73.5 113.1 98.8 147.5 187.5 740 74.8 113.1 99.9 147.9 187.9 750 76.1 113.1 101.0 148.2 188.2 760 77.4 113.1 102.0 148.6 188.6 770 78.6 113.1 103.0 149.0 189.0 780 79.8 113.1 104.0 149.4 189.4 790 81.0 113.1 105.0 149.8 189.8 800 82.2 113.1 105.9 150.1 190.1 810 83.3 113.1 106.9 150.5 190.5 820 84.4 113.4 107.8 150.9 190.9 830 85.5 114.1 108.7 151.2 191.2 840 86.5 114.8 109.6 151.6 191.6 850 87.6 115.5 110.4 151.9 191.9 860 88.6 116.2 111.3 152.3 192.3 870 89.6 116.9 112.1 152.6 192.6 880 90.5 117.6 113.0 153.0 193.0 890 91.5 118.3 113.8 153.3 193.3 900 92.4 118.9 114.6 153.7 193.7 910 93.4 119.6 115.4 154.0 194.0 920 94.3 120.2 116.1 154.4 194.4 930 95.1 120.9 116.9 154.7 194.7 940 96.0 121.5 117.7 155.0 195.0 950 96.9 122.1 118.4 155.4 195.4 B-11

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-2. Dresden Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-9, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (OF) (OF) (OF)

(PSIG) (OF) (OF) 960 97.7 122.7 119.1 155.7 195.7 970 98.6 123.3 119.9 156.0 196.0 980 99.4 123.9 120.6 156.4 196.4 990 100.2 124.5 121.3 156.7 196.7 1000 101.0 125.1 122.0 157.0 197.0 1010 101.7 125.7 122.6 157.3 197.3 1020 102.5 126.2 123.3 157.6 197.6 1030 103.3 126.8 124.0 158.0 198.0 1040 104.0 127.4 124.6 158.3 198.3 1050 104.7 127.9 125.3 158.6 198.6 1060 105.4 128.5 125.9 158.9 198.9 1070 106.2 129.0 126.5 159.2 199.2 1080 106.9 129.5 127.2 159.5 199.5 1090 107.6 130.1 127.8 159.8 199.8 1100 108.2 130.6 128.4 160.1 200.1 1105 108.6 130.8 128.7 160.3 200.3 1110 108.9 131.1 129.0 160.4 200.4 1120 109.6 131.6 129.6 160.7 200.7 1130 110.2 132.1 130.2 161.0 201.0 1140 110.9 132.6 130.7 161.3 201.3 1150 111.5 133.1 131.3 161.6 201.6 1160 112.1 133.6 131.9 161.9 201.9 1170 112.8 134.1 132.4 162.2 202.2 1180 113.4 134.6 133.0 162.5 202.5 1190 114.0 135.1 133.5 162.7 202.7 B-1 2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-2. Dresden Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-9, 5-11 and 5-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY CURVE A CURVE B CURVE B CURVE C PRESSURE CURVE A (PSIG) (OF) (°F) (OF) (OF) (OF) 1200 114.6 135.5 134.1 163.0 203.0 1210 115.2 136.0 134.6 163.3 203.3 1220 115.8 136.5 135.2 163.6 203.6 1230 116.3 137.0 135.7 163.9 203.9 1240 116.9 137.6 136.2 164.2 204.2 1250 117.5 138.2 136.7 164.4 204.4 1260 118.0 138.8 137.2 164.7 204.7 118.6 139.4 137.7 165.0 205.0 1270 1280 119.1 140.0 138.2 165.2 205.2 1290 119.7 140.6 138.7 165.6 205.6 1300 120.2 141.1 139.2 166.0 206.0 1310 120.7 141.7 139.7 166.5 206.5 1320 121.3 142.3 140.2 166.9 206.9 1330 121.8 142.8 140.6 167.4 207.4 1340 122.3 143.4 141.1 167.8 207.8 1350 122.8 143.9 141.6 168.3 208.3 1360 123.3 144.4 142.0 168.7 208.7 123.8 145.0 142.5 169.1 209.1 1370 1380 124.3 145.5 142.9 169.6 209.6 1390 124.8 146.0 143.4 170.0 210.0 1400 125.3 146.5 143.8 170.4 210.4 B-I 3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-3. Dresden Unit 3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) ( 0F) (OF) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 83.1 83.1 140 68.0 83.1 83.1 68.0 83.1 83.1 150 68.0 83.1 83.1 68.0 84.2 83.1 160 68.0 83.1 83.1 68.0 86.9 83.1 170 68.0 83.1 83.1 68.0 89.5 83.1 180 68.0 83.1 83.1 68.0 91.9 83.1 190 68.0 83.1 83.1 68.0 94.2 83.1 200 68.0 83.1 83.1 68.0 96.3 83.1 210 68.0 83.1 83.1 68.0 98.3 83.1 220 68.0 83.1 83.1 68.0 100.3 83.1 230 68.0 83.1 83.1 68.0 102.1 83.1 B-14

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-3. Dresden Unit 3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (°F) (OF) (OF) (OF) (OF) 240 68.0 83.1 83.1 68.0 103.9 83.1 250 68.0 83.1 83.1 68.0 105.6 83.1 260 68.0 83.1 83.1 68.0 107.2 83.1 270 68.0 83.1 83.1 68.0 108.8 83.1 280 68.0 83.1 83.1 68.0 110.3 83.1 290 68.0 83.1 83.1 68.0 111.8 83.1 300 68.0 83.1 83.1 68.0 113.2 83.1 310 68.0 83.1 83.1 68.0 114.5 83.1 312.5 68.0 83.1 83.1 68.0 114.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.0 143.1 143.1 430 68.0 113.1 113.1 68.0 143.1 143.1 440 68.0 113.1 113.1 68.0 143.1 143.1 450 68.0 113.1 113.1 68.0 143.1 143.1 460 68.0 113.1 113.1 68.0 143.1 143.1 470 68.0 113.1 113.1 68.0 143.1 143.1 B-1 5

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-3. Dresden Unit 3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 54 EFPY BOTTOM "UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (°F) (OF) 480 68.0 113.1 113.1 68.0 143.1 143.1 490 68.0 113.1 113.1 68.0 143.1 143.1 500 68.0 113.1 113.1 68.0 143.1 143.1 510 68.0 113.1 113.1 68.0 143.1 143.1 520 68.0 113.1 113.1 68.2 143.1 143.1 530 68.0 113.1 113.1 70.2 143.1 143.1 540 68.0 113.1 113.1 72.1 143.1 143.1 550 68.0 113.1 113.1 73.9 143.1 143.1 560 68.0 113.1 113.1 75.7 143.1 143.1 570 68.0 113.1 113.1 77.4 143.1 143.1 580 68.0 113.1 113.1 79.0 143.1 143.1 590 68.0 113.1 113.1 80.6 143.1 143.1 600 68.0 113.1 113.1 82.2 143.1 143.1 610 68.0 113.1 113.1 83.7 143.1 143.1 620 68.0 113.1 113.1 85.1 143.1 143.1 630 68.0 113.1 113.1 86.5 143.4 143.1 640 68.0 113.1 113.1 87.9 143.8 143.1 650 68.0 113.1 113.1 89.2 144.2 143.1 660 68.0 113.1 113.1 90.5 144.7 143.1 670 68.0 113.1 113.1 91.8 145.1 143.1 680 68.0 113.1 113.1 93.1 145.5 143.1 690 68.0 113.1 113.1 94.3 145.9 143.1 700 69.2 113.1 113.1 95.4 146.3 144.0 710 70.7 113.1 113.1 96.6 146.7 145.1 720 72.1 113.1 113.1 97.7 147.1 146.0 730 73.5 113.1 113.1 98.8 147.5 147.0 B-1 6

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-3. Dresden Unit 3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER' 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEA CURVEA CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (°F) (OF) 740 74.8 113.1 113.1 99.9 147.9 148.0 750 76.1 113.1 113.1 101.0 148.2 148.9 760 77.4 113.1 113.1 102.0 148.6 149.8 770 78.6 113.1 113.3 103.0 149.0 150.7 780 79.8 113.1 114.7 104.0 149.4 151.6 790 81.0 113.1 116.1 105.0 149.8 152.5 800 82.2 113.1 117.4 105.9 150.1 153.3 810 83.3 113.1 118.7 106.9 150.5 154.2 820 84.4 113.4 120.0 107.8 150.9 155.0 830 85.5 114.1 121.2 108.7 151.2 155.8 840 86.5 114.8 122.4 109.6 151.6 156.6 850 87.6 115.5 123.5 110.4 151.9 157.4 860 88.6 116.2 124.7 111.3 152.3 158.2 870 89.6 116.9 125.8 112.1 152.6 158.9 880 90.5 117.6 126.9 113.0 153.0 159.7 890 91.5 118.3 128.0 113.8 153.3 160.4 900 92.4 118.9 129.0 114.6 153.7 161.2 910 93.4 119.6 130.0 115.4 154.0 161.9 920 94.3 120.2 131.0 116.1 154.4 162.6 930 95.1 120.9 132.0 116.9 154.7 163.3 940 96.0 121.5 133.0 117.7 155.0 164.0 950 96.9 122.1 133.9 118.4 155.4 164.7 960 97.7 122.7 134.8 119.1 155.7 165.3 970 98.6 123.3 135.8 119.9 156.0 166.0 980 99.4 123.9 136.7 120.6 156.4 166.6 990 100.2 124.5 137.5 121.3 156.7 167.3 B-1 7

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-3. Dresden Unit 3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 1000 101.0 125.1 138.4 122.0 157.0 167.9 1010 101.7 125.7 139.2 122.6 157.3 168.6 1020 102.5 126.2 140.1 123.3 157.6 169.2 1030 103.3 126.8 140.9 124.0 158.0 169.8 1040 104.0 127.4 141.7 124.6 158.3 170.4 1050 104.7 127.9 142.5 125.3 158.6 171.0 1060 105.4 128.5 143.3 125.9 158.9 171.6 1070 106.2 129.0 144.1 126.5 159.2 172.2 1080 106.9 129.5 144.8 127.2 159.5 172.8 1090 107.6 130.1 145.6 127.8 159.8 173.3 1100 108.2 130.6 146.3 128.4 160.1 173.9 1105 108.6 130.8 146.7 128.7 160.3 174.2 1110 108.9 131.1 147.0 129.0 160.4 174.4 1120 109.6 131.6 147.7 129.6 160.7 175.0 1130 110.2 132.1 148.5 130.2 161.0 175.5 1140 110.9 132.6 149.2 130.7 161.3 176.1 1150 111.5 133.1 149.8 131.3 161.6 176.6 1160 112.1 133.6 150.5 131.9 161.9 177.1 1170 112.8 134.1 151.2 132.4 162.2 177.7 1180 113.4 134.6 151.8 133.0 162.5 178.2 1190 114.0 135.1 152.5 133.5 162.7 178.7 1200 114.6 135.5 153.1 134.1 163.0 179.2 1210 115.2 136.0 153.8 134.6 163.3 179.7 1220 115.8 136.5 154.4 135.2 163.6 180.2 1230 116.3 136.9 155.0 135.7 163.9 180.7 1240 116.9 137.4 155.6 136.2 164.2 181.2 B-1 8

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-3. Dresden Unit 3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-1, 5-2, 5-4, 5-5, 5-6, & 5-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (°F) (OF) (OF) (OF) (OF) (OF) 1250 117.5 137.8 156.2 136.7 164.4 181.7 1260 118.0 138.3 156.8 137.2 164.7 182.2 1270 118.6 138.7 157.4 137.7 165.0 182.6 1280 119.1 139.2 158.0 138.2 165.2 183.1 1290 119.7 139.6 158.6 138.7 165.5 183.6 1300 120.2 140.0 159.1 139.2 165.8 184.0 1310 120.7 140.5 159.7 139.7 166.1 184.5 1320 121.3 140.9 160.3 140.2 166.3 184.9 1330 121.8 141.3 160.8 140.6 166.6 185.4 1340 122.3 141.7 161.4 141.1 166.8 185.8 1350 122.8 142.1 161.9 141.6 167.1 186.3 1360 123.3 142.6 162.4 142.0 167.4 186.7 1370 123.8 143.0 163.0 142.5 167.6 187.1 1380 124.3 143.4 163.5 142.9 167.9 187.6 1390 124.8 143.8 164.0 143.4 168.1 188.0 1400 125.3 144.2 164.5 143.8 168.4 188.4 B-I 9

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-4. Dresden Unit 3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B &C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 84.0 70 68.0 83.1 68.0 83.1 91.2 80 68.0 83.1 68.0 83.1 97.2 90 68.0 83.1 68.0 83.1 102.3 100 68.0 83.1 68.0 83.1 106.8 110 68.0 83.1 68.0 83.1 110.9 120 68.0 83.1 68.0 83.1 114.7 130 68.0 83.1 68.0 83.1 118.2 140 68.0 83.1 68.0 83.1 121.4 150 68.0 83.1 68.0 84.2 124.2 160 68.0 83.1 68.0 86.9 126.9 170 68.0 83.1 68.0 89.5 129.5 180 68.0 83.1 68.0 91.9 131.9 190 68.0 83.1 68.0 94.2 134.2 200 68.0 83.1 68.0 96.3 136.3 210 68.0 83.1 68.0 98.3 138.3 220 68.0 83.1 68.0 100.3 140.3 B-20

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE B-4. Dresden Unit 3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVEB CURVE C (PSIG) (OF) (OF) (*F) (OF) (OF) 230 68.0 83.1 68.0 102.1 142.1 240 68.0 83.1 68.0 103.9 143.9 250 68.0 83.1 68.0 105.6 145.6 260 68.0 83.1 68.0 107.2 147.2 270 68.0 83.1 68.0 108.8 148.8 280 68.0 83.1 68.0 110.3 150.3 290 68.0 83.1 68.0 111.8 151.8 300 68.0 83.1 68.0 113.2 153.2 310 68.0 83.1 68.0 114.5 154.5 312.5 68.0 83.1 68.0 114.9 154.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.0 143.1 183.1 430 68.0 113.1 68.0 143.1 183.1 440 68.0 113.1 68.0 143.1 183.1 450 68.0 113.1 68.0 143.1 183.1 B-21

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-4. Dresden Unit 3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV &' BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (°F) (OF) 460 68.0 113.1 68.0 143.1 183.1 470 68.0 113.1 68.0 143.1 183.1 480 68.0 113.1 68.0 143.1 183.1 490 68.0 113.1 68.0 143.1 183.1 500 68.0 113.1 68.0 143.1 183.1 510 68.0 113.1 68.0 143.1 183.1 520 68.0 113.1 68.2 143.1 183.1 530 68.0 113.1 70.2 143.1 183.1 540 68.0 113.1 72.1 143.1 183.1 550 68.0 113.1 73.9 143.1 183.1 560 68.0 113.1 75.7 143.1 183.1 570 68.0 113.1 77.4 143.1 183.1 580 68.0 113.1 79.0 143.1 183.1 590 68.0 113.1 80.6 143.1 183.1 600 68.0 113.1 82.2 143.1 183.1 610 68.0 113.1 83.7 143.1 183.1 620 68.0 113.1 85.1 143.1 183.1 630 68.0 113.1 86.5 143.4 183.4 640 68.0 113.1 87.9 143.8 183.8 650 68.0 113.1 89.2 144.2 184.2 660 68.0 113.1 90.5 144.7 184.7 670 68.0 113.1 91.8 145.1 185.1 680 68.0 113.1 93.1 145.5 185.5 690 68.0 113.1 94.3 145.9 185.9 700 69.2 113.1 95.4 146.3 186.3 B-22

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-4. Dresden Unit 3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 710 70.7 113.1 96.6 146.7 186.7 720 72.1 113.1 97.7 147.1 187.1 730 73.5 113.1 98.8 147.5 187.5 740 74.8 113.1 99.9 148.0 188.0 750 76.1 113.1 101.0 148.9 188.9 760 77.4 113.1 102.0 149.8 189.8 770 78.6 113.3 103.0 150.7 190.7 780 79.8 114.7 104.0 151.6 191.6 790 81.0 116.1 105.0 152.5 192.5 800 82.2 117.4 105.9 153.3 193.3 810 83.3 118.7 106.9 154.2 194.2 820 84.4 120.0 107.8 155.0 195.0 830 85.5 121.2 108.7 155.8 195.8 840 86.5 122.4 109.6 156.6 196.6 850 87.6 123.5 110.4 157.4 197.4 860 88.6 124.7 111.3 158.2 198.2 870 89.6 125.8 112.1 158.9 198.9 880 90.5 126.9 113.0 159.7 199.7 890 91.5 128.0 113.8 160.4 200.4 900 92.4 129.0 114.6 161.2 201.2 910 93.4 130.0 115.4 161.9 201.9 920 94.3 131.0 116.1 162.6 202.6 930 95.1 132.0 116.9 163.3 203.3 940 96.0 133.0 117.7 164.0 204.0 950 96.9 133.9 118.4 164.7 204.7 B-23

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE B-4. Dresden Unit 3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT LIMITING 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 960 97.7 134.8 119.1 165.3 205.3 970 98.6 135.8 119.9 166.0 206.0 980 99.4 136.7 120.6 166.6 206.6 990 100.2 137.5 121.3 167.3 207.3 1000 101.0 138.4 122.0 167.9 207.9 1010 101.7 139.2 122.6 168.6 208.6 1020 102.5 140.1 123.3 169.2 209.2 1030 103.3 140.9 124.0 169.8 209.8 1040 104.0 141.7 124.6 170.4 210.4 1050 104.7 142.5 125.3 171.0 211.0 1060 105.4 143.3 125.9 171.6 211.6 1070 106.2 144.1 126.5 172.2 212.2 1080 106.9 144.8 127.2 172.8 212.8 1090 107.6 145.6 127.8 173.3 213.3 1100 108.2 146.3 128.4 173.9 213.9 1105 108.6 146.7 128.7 174.2 214.2 1110 108.9 147.0 129.0 174.4 214.4 1120 109.6 147.7 129.6 175.0 215.0 1130 110.2 148.5 130.2 175.5 215.5 1140 110.9 149.2 130.7 176.1 216.1 1150 111.5 149.8 131.3 176.6 216.6 1160 112.1 150.5 131.9 177.1 217.1 1170 112.8 151.2 132.4 177.7 217.7 1180 113.4 151.8 133.0 178.2 218.2 1190 114.0 152.5 133.5 178.7 218.7 B-24

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE B-4. Dresden Unit 3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures 5-10, 5-13 and 5-14 "UPPERRPV & BOTTOM UPPER RPV &

HEAD HEAD BELTLINE AT LIMITING 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE C (PSIG) (°F) (°F) (OF) (OF) 1200 114.6 153.1 134.1 179.2 219.2 1210 115.2 153.8 134.6 179.7 219.7 1220 115.8 154.4 135.2 180.2 220.2 1230 116.3 155.0 135.7 180.7 220.7 1240 116.9 155.6 136.2 181.2 221.2 1250 117.5 156.2 136.7 181.7 221.7 1260 118.0 156.8 137.2 182.2 222.2 1270 118.6 157.4 137.7 182.6 222.6 1280 119.1 158.0 138.2 183.1 223.1 1290 119.7 158.6 138.7 183.6 223.6 1300 120.2 159.1 139.2 184.0 224.0 1310 120.7 159.7 139.7 184.5 224.5 1320 121.3 160.3 140.2 184.9 224.9 1330 121.8 160.8 140.6 185.4 225.4 1340 122.3 161.4 141.1 185.8 225.8 1350 122.8 161.9 141.6 186.3 226.3 1360 123.3 162.4 142.0 186.7 226.7 1370 123.8 163.0 142.5 187.1 227.1 1380 124.3 163.5 142.9 187.6 227.6 1390 124.8 164.0 143.4 188.0 228.0 1400 125.3 164.5 143.8 188.4 228.4 B-25

GE Nuclear Energy GE-NE-0000-0002-9600-O la APPENDIX C OPERATING AND TEMPERATURE MONITORING REQUIREMENTS C-1

GE Nuclear Energy GE-NE-0000-0002-9600-01 a C.1 NON-BELTLINE MONITORING DURING PRESSURE TESTS It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

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

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

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

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

C-2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by _20°F per hour. Ifthe coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.

C.2.2 Curve B: Non-Nuclear Heatup/Cooldown Curve B should be used whenever Curve A or Curve C do not apply. In other words, the operator must follow this curve during times when the coolant is heating or cooling faster than 20°F per hour during a hydrotest and when the core is not critical.

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

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

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

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

C-3

GE Nuclear Energy GE-NE-0000-0002-9600-01a In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:

"* Head flange boltup

"* Leakage test (Curve A compliance)

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

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

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

C-4

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX C

REFERENCES:

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

C-5

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX D GE SIL 430 D-1

GE Nuclear Energy GE-NE-0000-0002-9600-01a September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions concerning primary and alternate reactor pressure vessel (RPV) temperature monitoring measurements for complying with RPV brittle fracture and thermal stress requirements.

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

TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)

Measurement Use Limitations Steam dome saturation Primary measurement Must convert saturated temperature as determined above 212°F for Tech steam pressure to from main steam instrument Spec IOOOF/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

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

and cooldown rate.

Alternate measurement When above 212°F need to above 2120F. allow for temperature variations (up to 10-15 0 F lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

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

RHR heat exchanger Alternate measurement Must have previously inlet coolant for Tech Spec 1OOOF/hr correlated RHR inlet temperature cooldown rate when in coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow. Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta T's will be indicated coolant temperature. Delta T limit is 1000 F for BWR/6s and 145 0F for earlier BWRs.

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

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

outside metal surface Should have drain line flow.

temperatures.

D-3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations S.................................

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

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

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

temperature limit for head boltup.

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

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

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

Top head outside Information only. Very slow to respond to RPV surface T/Cs coolant changes. Not avail able on BWR/6s.

D-4

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 GE-NE-0000-0002-9600-01a Product

Reference:

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

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

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

D-6

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX E DETERMINATION OF BELTLINE REGION AND IMPACT ON FRACTURE TOUGHNESS E-1

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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/cm 2 . Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal 1.0e17 n/cm 2 , then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel, and do not need to be considered in the P-T curve evaluation.

The following dimensions are obtained from the referenced drawings:

Shell # 2 - Top of Active Fuel (TAF): 360.3" (from vessel 0) [1]

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

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

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

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

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

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

From [2], it is obvious that the recirculation inlet and outlet nozzles are closest to the beltline region (the top of the recirculation inlet nozzle is -23" from BAF and the top of the recirculation outlet nozzle is -28" from BAF), and no other nozzles are within the BAF-TAF region of the reactor vessel. The girth weld between Shell Rings #2 and#3 is

-31" above TAF. Therefore, if it can be shown that the peak fluence at these locations is less than 1.0e17 n/cm 2 , it can be safely concluded that all nozzles and welds, other E-2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a than those included in Tables 4-3 and 4-4, are outside the beltline region of the reactor vessel.

Based on the bounding 32 and 54 EFPY axial flux profile for pre- and post-EPU [5, 6],

the RPV flux level dropped to less than 1e17 n/cm2 at the same radius at -1" below the BAF and at -6" above TAF. The beltline region considered in the development of the P-T curves is adjusted to include the additional 6" above the active fuel region and the additional 1", below the active fuel region. This adjusted beltline region extends from 215.3" to 366.3" above reactor vessel "0" for both 32 and 54 EFPY.

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

E-3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX E

REFERENCES:

1. Dresden/Quad Cities LR PT Curves - Data Input Request, Robert Stachniak (Exelon), 4126102.

2. Babcock & Wilcox Co. (B&W) Drawing #151803E, Revision 1, "Recirculation Nozzles", (GE-NE VPF# 2252-130-3), Dresden Unit 3.

3. Babcock & Wilcox Co. (B&W) Drawing #26903F, Revision 2, "General Outline", (GE-NE VPF# 2252-139-4), Dresden Unit 3.

4. Babcock & Wilcox Co. (B&W) Drawing # 151797E, Revision 1, "Shell Segment Assembly", (GE-NE VPF# 2252-126-3), Dresden Unit 3.

5. S. Sitaraman, "Dresden and Quad Cities Neutron Flux Evaluation," GE-NE, San Jose, CA, December 2002, (GE-NE-0000-001 1-0531-RO, Revision 0)(GE Proprietary Information).

6. S.S. Wang, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T0313: RPV Flux Evaluation", GE-NE, San Jose, CA, (GE-NE A22-00103-29-01, Revision 0), October 2000, (GE Proprietary Information).

E-4

GE Nuclear Energy. GE-NE-0000-0002-9600-01 a APPENDIX F EQUIVALENT MARGIN ANALYSIS (EMA) FOR UPPER SHELF ENERGY (USE)

F-1

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 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 and 54 EFPY. Calculations of 32 and 54 EFPY USE, using Regulatory Guide 1.99, Revision 2 [2] methods and BWROG Equivalent Margin Analyses [3, 4, 5] methods are summarized in Tables F-1 through F-4.

Unirradiated upper shelf data was not available for all of the material heats in the Dresden Unit 3 beltline region. Therefore, Dresden Unit 3 is evaluated to verify that the BWROG EMA is applicable. The USE decrease prediction values from Regulatory Guide 1.99, Revision 2 are used for the beltline components as shown in Tables F-1 through F-4. These calculations are based upon the 32 and 54 EFPY peak 1/4T fluence as provided in Tables 4-3 and 4-4.

Based on the results presented in Tables F-1 through F-4, the USE EMA values for the Dresden Unit 3 reactor vessel beltline materials remain within the limits of Regulatory Guide 1.99, Revision 2 and I OCFR50 Appendix G for both 32 and 54 EFPY of operation.

F-2

GE Nuclear Energy GE-NE-0000-0002-9600-01a Table F-1 Equivalent Margin Analysis Plant Applicability Verification Form for Dresden Unit 3 For 32 EFPY (including Extended Power Uprate)

BWRI3-6 PLATE Surveillance Plate USE-

%Cu = 0.13 1 st Capsule Fluence = 9.3 x 1015 n/cm 2

= 2 2nd Capsule Fluence 2.9 x 1016 n/cm

= 2 3 rd Capsule Fluence 7.1 x 1016 n/cm 1 st Capsule Measured % Decrease = 0 (Charpy Curves)

Capsule Measured % Decrease = -4 (increase) (Charpy Curves) 2 nd 3rd Capsule Measured % Decrease = -11 (increase) (Charpy Curves)

Capsule R.G. 1.99 Predicted % Decrease = 4 (R.G. 1.99, Figure 2) 1 st Capsule R G. 1.99 Predicted % Decrease = 6 (R G 1.99, Figure 2) 2 nd Td 3 Capsule R G. 1.99 Predicted % Decrease = 7 (R.G. 1.99, Figure 2)

Limiting Beltline Plate (Heat A0237-1) USE-

%Cu = 0.23 2

32 EFPY 114T Fluence = 2.3 x 1017 n/cm R.G. 1.99 Predicted % Decrease = 14 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 14%:5 21%, so vessel plates are bounded by equivalent margin analysis F-3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table F-2 Equivalent Margin Analysis Plant Applicability Verification Form for Dresden Unit 3 For 32 EFPY (including Extended Power Uprate)

BWR/2-6 WELD Surveillance Weld USE-

%Cu = 0.20 Is' Capsule Fluence = 9.3 x 1015 n/cm 2 2nd Capsule Fluence = 2.9 x 1016 n/cm2 2d Capsule Fluence = 2 7.1 x 106 n/cm 1 St Capsule Measured % Decrease = -7 (increase) (Charpy Curves) 2nd Capsule Measured % Decrease = -51 (increase) (Charpy Curves) 3rd Capsule Measured % Decrease

= 0 (Charpy Curves)

Ist Capsule R.G. 1.99 Predicted % Decrease = 7 (R.G. 1.99, Figure 2) 2 nd Capsule R.G. 1.99 Predicted % Decrease = 9 (R.G. 1.99, Figure 2) 3rd Capsule R.G. 1.99 Predicted % Decrease = 11 (R.G. 1.99, Figure 2)

Limiting Beltline Weld (299L44) USE:

%Cu = 034 32 EFPY 1/4T Fluence = 2.3 x 1017 n/cm 2 R.G. 1.99 Predicted % Decrease = 20.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = NIA (R.G. 1.99, Position 2.2) 20.5%:5 34%, so vessel welds are bounded by equivalent margin analysis F-4

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table F-3 Equivalent Margin Analysis Plant Applicability Verification Form for Dresden Unit 3 For 54 EFPY (including Extended Power Uprate)

BWR/3-6 PLATE Surveillance Plate USE,

%Cu = 0.13 1 st Capsule Fluence = 9.3 x Q10s n/cm 2 2

2nd Capsule Fluence = 2.9 x 1016 n/cm Capsule Fluence = 2 3 rd 7.1 x 1016 n/cm 1 st Capsule Measured % Decrease =0 (Charpy Curves)

= -4 (increase) (Charpy Curves) 2 nd Capsule Measured % Decrease

= -11 (increase) (Charpy Curves) 3d Capsule Measured % Decrease

=4 (R.G. 1.99, Figure 2)

Ist Capsule R.G. 1.99 Predicted % Decrease

=6 (R.G. 1.99, Figure 2) 2 nd Capsule R.G. 1.99 Predicted % Decrease

=7 (R.G. 1.99, Figure 2) 3'd Capsule R.G. 1.99 Predicted % Decrease Limiting Beltline Plate (Heat A0237-11 USE:

%Cu = 0.23 2

54 EFPY 1/4T Fluence = 3.9 x 1017 n/cm R.G. 1.99 Predicted % Decrease = 15.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 15.5% < 23.5%, so vessel plates are bounded by equivalent margin analysis F-5

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table F-4 Equivalent Margin Analysis Plant Applicability Verification Form for Dresden Unit 3 For 54 EFPY (including Extended Power Uprate)

BWRI2-6 WELD Surveillance Weld USE:

%Cu = 0.20 2

1 st Capsule Fluence = 9.3 x 1015 n/cm 2 nd Capsule Fluence = 2.9 x 1016 n/cm2 2

3'd Capsule Fluence = 7.1 x 1016 n/cm 1 st Capsule Measured % Decrease = -7 (increase) (Charpy Curves) 2"d Capsule Measured % Decrease = -51 (increase) (Charpy Curves) 3rd Capsule Measured % Decrease = 0 (Charpy Curves) 1st Capsule R.G. 1.99 Predicted % Decrease = 7 (R.G. 1.99, Figure 2)

Capsule R.G. 1.99 Predicted % Decrease = 9 (R.G. 1.99, Figure 2) 2 nd 3rd Capsule R.G. 1.99 Predicted % Decrease = 11 (R.G. 1.99, Figure 2)

Limiting Beltline Weld (299L44) USE-

%Cu = 0.34 2

54 EFPY 114T Fluence = 2.9 x 1017 n/cm R G. 1.99 Predicted % Decrease = 21.5 (R.G. 1.99, Figure 2)

Adjusted % Decrease = N/A (R.G. 1.99, Position 2.2) 21.5%:5 39%, so vessel welds are bounded by equivalent margin analysis F-6

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX F

REFERENCES:

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

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

3. J.T. Wiggins (NRC) to L.A. England (Gulf States Utilities Co.), "Acceptance for Referencing of Topical Report NEDO-32205, Revision 1, '10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels"', December 8, 1993.

4. L.A. England (BWR Owners' Group) to Daniel G. McDonald (USNRC), "BWR Owners' Group Topical Report on Upper Shelf Energy Equivalent Margin Analysis - Approved Version", BWROG-94037, March 21, 1994.

5. C.I. Grimes (NRC) to Carl Terry (Niagara Mohawk Power Company),

"Acceptance For Referencing Of EPRI Proprietary Report TR-1 13596, "BWR Vessel And Internals Project, BWR Reactor Pressure Vessel Inspection And Flaw Evaluation Guidelines (BWRVIP-74)" And Appendix A, "Demonstration Of Compliance With The Technical Information Requirements Of The License Renewal Rule (10 CFR 54.21)", October 18, 2001.

F-7

GE Nuclear Energy GE-NE-0000-0002-9600-01 a APPENDIX G BOUNDING P-T CURVES FOR DRESDEN UNITS 2 & 3 G-1

GE Nuclear Energy GE-NE-0000-0002-9600-01 a This appendix contains P-T curves that bound the limiting material characteristics of both Dresden Unit 2 and Dresden Unit 3. Composite and individual curves are presented for both 32 and 54 EFPY similar to those provided within the main body of this report. Table G-1 provides the figure numbers and the corresponding tabulation for each P-T curve presented in this appendix.

G-2

GE Nuclear Energy GE-NE-0000-0002-9600-01 a Table G-1: Composite and Individual Curves Used To Construct Composite P-T Curves Figure Table Numbers Curve, Curve Description Numbers for for Presentation of' Presentation of

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

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

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

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

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

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

G-3

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100 CL "1000 LUi x

IL 900 0 IINITIAL 49 F FORRTndt BOTTOMVALUE ISI HEAD w

Un 800 U)

O 700 HEATUP/COOLDOWN RATE OF COOLANT

< 2 0 °F/HR S600 Z

j 500 Lu gLU 400 a-300 200 100 0

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

Figure G-1: Bounding Dresden 2&3 Bottom Head P-T Curve for Pressure Test [Curve A]

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

G4

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100 1000 S900 INITIAL RTndt VALUE IS 0

II-3 51 F FOR UPPER VESSELI (n800 o 700 w

HEATUP/COOLDOWN RATE OF COOLANT S600 < 20°F/HR 2

S500 g) 400 w

0:

F312 PSIG 300 200 FLANGE REGION 83*F -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 G-2: Bounding Dresden 2&3 Upper Vessel P-T Curve for Pressure Test [Curve A]

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

G-5

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100

"- 1000 0- 900 0

BELTLINE CURVE S800 _. . .. . ADJUSTED AS SHOWN:

w EFPY SHIFT (°F) 32 63 o 700 HEATUP/COOLDOWN z* 600 RATE OF COOLANT

< 20 °F/HR

. 500 w

) 400 w

300 200 105- BELTLINE LIMITS 100 0

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

(°F)

Figure G-3: Bounding Dresden 2&3 Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY [20 °F/hr or less coolant heatup/cooldown]

G-6

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 CL "1000 w

IL 900 0

BELTLINE CURVE ADJUSTED AS SHOWN:

S800 EFPY SHIFT (°F) w 54 81 o 700 HEATUP/COOLDOWN S600 RATE OF COOLANT z < 20°F/HR S500 n 400 U) 0.

300 200 -BELTLINE LIMITS 100 0

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

Figure G-4: Bounding Dresden 2&3 Beltline P-T Curve for Pressure Test [Curve A] up to 54 EFPY [20°F/hr or less coolant heatup/cooldown]

G-7

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100

"-1000 INITIAL RTndt VALUE IS I.

w 900 49-F FOR BOTTOM HEAD 0

'0 800 U) o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 100"F/HR w

W- 600 Z

j500 LuI 400 IL 300 200 100 0

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

Figure G-5: Bounding Dresden 2&3 Bottom Head P-T Curve for Core Not Critical [Curve B]

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

G-8

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 1100 0.

"- 1000 w

S900 INITIAL RTndt VALUE IS I 0

51°F FOR UPPER VESSELI S800 LU o 700 HEATUP/COOLDOWN RATE OF COOLANT Lu < 100°F/HR k 600 i 500 Lu nV 400 Lu Jr.

300 200

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

J1FLANGE REGION 83FO 0

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

Figure G-6: Bounding Dresden 2&3 Upper Vessel P-T Curve for Core Not Critical [Curve B]

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

G-9

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.10F FOR BELTLINE 1100 C.

1000 BELTLINE CURVE ADJUSTED AS SHOWN:

0 EFPY SHIFT (°F) 900 32 63 0 800 a

700 HEATUP/COOLDOWN a,, RATE OF COOLANT a, 600 < 100°F/HR It 500 co 400 uI a,.

r312-PýSIG 300 200 10CFR50 - BELTLINE LIMITS BOLTUP 100 0

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

Figure G-7: Bounding Dresden 2&3 Beltline P-T Curve for Core Not Critical [Curve B] up to 32 EFPY 0 F/hr or less coolant heatup/cooldown]

[1 OO G-10

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUE IS 23.1°F FOR BELTLINE 1100 U.

S1000 BELTLINE CURVE ADJUSTED AS SHOWN:

EFPY SHIFT (°F)

CL 900 54 81 0

800 LU 0 700 HEATUP/COOLDOWN RATE OF COOLANT w 600 < 100°F/HR z

500 w

u 400 3O LU 312 PSIG 300 200 10CFRSO -BELTLINE LIMITS BOLTUP 100 - 83°F 0

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

(°F)

Figure G-8: Bounding Dresden 2&3 Beltline P-T Curve for Core Not Critical [Curve B] up to 54 EFPY

[1 O0°F/hr or less coolant heatup/cooldown]

G-11

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 INITIAL RTndt VALUES ARE 1300 23.1°F FOR BELTLINE, 51°F FOR UPPER VESSEL, 1200 AND 49°F FOR BOTTOM HEAD 1100 U)

BELTLINE CURVE "1000 ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 32 63 IL 900 0

S800 HEATUP/COOLDOWN LU RATE OF COOLANT

< 10 0 °F/HR o 700 600 Z

f 500 LU

) 400 IL 300 200 AND 0000-BELTLINE NON-BELTLINE 100 S. . .. Minimum Cnticality LIMITS mperature 83°F 0

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

Figure G-9: Bounding Dresden 2&3 Core Critical P-T Curves [Curve C] up to 32 EFPY

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

G-12

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 INITIAL RTndt VALUES ARE 23.1°F FOR BELTLINE, 1300 51°F FOR UPPER VESSEL, AND 1200 49°F FOR BOTTOM HEAD 1100 U)

C. BELTLINE CURVE "1000 ADJUSTED AS SHOWN:

w EFPY SHIFT (OF) 54 81

.900 0

co

( 800 w

o 700 z

3 500 w

S400 LU 0:

300 200

- BELTLINE AND SMinimum Cnticahty NON-BELTLINE 100 LIMITS

Temperature 83°F 0

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

(°F)

Figure G-10: Bounding Dresden 2&3 Core Critical P-T Curves [Curve C] up to 54 EFPY

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

G-13

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 INITIAL RTndt VALUES ARE 1200 23.1°F FOR BELTLINE, 51°F FOR UPPER VESSEL, AND 1100 49°F FOR BOTTOM HEAD a.

1000 BELTLINE CURVES IL ADJUSTED AS SHOWN:

EFPY SHIFT (°F) 900 32 63 "I"l o

U/)

U) 800 i-C.

0uJ 700 z HEATUP/COOLDOWN RATE OF COOLANT Lu, < 20=FIHR 600 o

U) 500 i:3

,,Ix 400 uIL 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 G-1 1: Bounding Dresden 2&3 Composite Pressure Test P-T Curves [Curve A]

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

G-14

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUES ARE 23.10 F FOR BELTLINE, 51 F FOR UPPER VESSEL, 1100 AND 49*F FOR BOTTOM HEAD C.

S1000 a BELTLINE CURVES ADJUSTED AS SHOWN:

IL 900 EFPY SHIFT (°F) 0 32 63 (n 800 o 700 HEATUPICOOLDOWN RATE OF COOLANT S600 < 100°FIHR 3 500 n 400 Lu 300

-UPPER VESSEL 200 AND BELTLINE LIMITS

-......BOTTOM HEAD 100 CURVE 0

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

(°F)

Figure G-12: Bounding Dresden 2&3 Composite Core Not Critical P-T Curves [Curve B]

up to 32 EFPY [1 00°F/hr or less coolant heatup/cooldown]

G-15

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 INITIAL RTndt VALUES ARE 1200 23.1*F FOR BELTLINE, 51*F FOR UPPER VESSEL, AND 1100 49*F FOR BOTTOM HEAD 0.

"1000 BELTLINE CURVES w ADJUSTED AS SHOWN:

EFPY SHIFT (OF)

IL 900 54 81 0

U0 800 C')

w o 700 w

=

z 600 S500 U) 400 w

c.

300

- UPPER VESSEL 200 AND BELTLINE LIMITS

.BOTTOM HEAD 100 CURVE 0

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

Figure G-13: Bounding Dresden 2&3 Composite Pressure Test P-T Curves [Curve A]

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

G-16

GE Nuclear Energy GE-NE-0000-0002-9600-01 a 1400 1300 1200 INITIAL RTndt VALUES ARE 23.1°F FOR BELTLINE, 51°F FOR UPPER VESSEL, 1100 AND 49°F FOR BOTTOM HEAD C.

w 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

CL 900 EFPY SHIFT (°F) 0 54 81 III w

cn 800 IL, o 700 HEATUP/COOLDOWN w RATE OF COOLANT S600 < 100°F/HR Z

S500 S400 w

300

-UPPER VESSEL 200 AND BELTLINE LIMITS

-......BOTTOM HEAD 100 CURVE 0

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

(°F)

Figure G-14: Bounding Dresden 2&3 Composite Core Not Critical P-T Curves [Curve B]

for up to 54 EFPY [1OO0 F/hr or less coolant heatup/cooldown]

G-17

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-2. Bounding Dresden 2&3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6, & G-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (°F) (OF) (OF) (°F) (OF) (°F) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 85.2 83.1 140 68.0 83.1 83.1 68.0 88.4 83.1 150 68.0 83.1 83.1 68.0 91.2 83.1 160 68.0 83.1 83.1 68.0 93.9 83.1 170 68.0 83.1 83.1 68.0 96.5 83.1 180 68.0 83.1 83.1 68.0 98.9 83.1 190 68.0 83.1 83.1 68.0 101.2 83.1 200 68.0 83.1 83.1 68.0 103.3 83.1 210 68.0 83.1 83.1 68.0 105.3 83.1 220 68.0 83.1 83.1 68.0 107.3 83.1 230 68.0 83.1 83.1 68.0 109.1 83.1 240 68.0 83.1 83.1 68.0 110.9 83.1 G-18

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-2. Bounding Dresden 2&3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6, & G-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE CURVE A CURVE A CURVE A SB CURVE B CURVE B PRESSURE (OF) (OF) (OF) (OF) (°F) (OF)

(PSIG) 68.0 83.1 83.1 68.0 112.6 83.1 250 68.0 83.1 83.1 68.0 114.2 83.1 260 68.0 83.1 83.1 68.0 115.8 83.1 270 68.0 83.1 83.1 68.0 117.3 83.1 280 68.0 83.1 83.1 68.0 118.8 83.1 290 68.0 83.1 83.1 68.0 120.2 83.1 300 68.0 83.1 83.1 68.0 121.5 83.1 310 68.0 83.1 83.1 68.0 121.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 68.0 1131 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.0 143.1 143.1 430 68.0 113.1 113.1 68.0 143.1 143.1 440 68.0 113.1 113.1 68.0 143.1 143.1 450 68.0 113.1 113.1 68.0 143.1 143.1 460 68.0 113.1 113.1 68.0 143.1 143.1 470 68.0 113.1 113.1 68.0 143.1 143.1 480 G-19

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE G-2. Bounding Dresden 2&3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6, & G-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (°F) 490 68.0 113.1 113.1 68.0 143.1 143.1 500 68.0 113.1 113.1 68.0 143.1 143.1 510 68.0 113.1 113.1 68.0 143.1 143.1 520 68.0 113.1 113.1 68.2 143.2 143.1 530 68.0 113.1 113.1 70.2 144.0 143.1 540 68.0 113.1 113.1 72.1 144.8 143.1 550 68.0 113.1 113.1 73.9 145.6 143.1 560 68.0 113.1 113.1 75.7 146.4 143.1 570 68.0 113.1 113.1 77.4 147.1 143.1 580 68.0 113.1 113.1 79.0 147.9 143.1 590 68.0 113.1 113.1 80.6 148.6 143.1 600 68.0 113.1 113.1 82.2 149.1 143.1 610 68.0 113.1 113.1 83.7 149.6 143.1 620 68.0 113.1 113.1 85.1 150.0 143.1 630 68.0 113.1 113.1 86.5 150.4 143.1 640 68.0 113.1 113.1 87.9 150.8 143.1 650 68.0 113.1 113.1 89.2 151.2 143.1 660 68.0 113.1 113.1 90.5 151.7 143.1 670 68.0 113.1 113.1 91.8 152.1 143.1 680 68.0 113.1 113.1 93.1 152.5 143.1 690 68.0 113.1 113.1 94.3 152.9 143.1 700 69.2 113.1 113.1 95.4 153.3 143.1 710 70.7 113.1 113.1 96.6 153.7 143.1 720 72.1 113.1 113.1 97.7 154.1 143.1 730 73.5 113.3 113.1 98.8 154.5 143.1 740 74.8 114.1 113.1 99.9 154.9 143.1 G-20

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-2. Bounding Dresden 2&3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6, & G-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE CURVE A CURVE A CURVE A B CURVE B CURVE B PRESSURE (PSIG) (OF) (OF) (OF) (0F) (OF) (OF) 76.1 115.0 113.1 101.0 155.2 143.1 750 77.4 115.8 113.1 102.0 155.6 143.1 760 770 78.6 116.6 113.1 103.0 156.0 143.1 79.8 117.3 113.1 104.0 156.4 143.1 780 81.0 118.1 113.1 105.0 156.8 143.1 790 82.2 118.9 113.1 105.9 157.1 143.1 800 83.3 119.6 113.1 106.9 157.5 143.1 810 84.4 120.4 113.1 107.8 157.9 143.1 820 85 5 121.1 113.1 108.7 158.2 143.1 830 86.5 121.8 113.1 109.6 158.6 143.1 840 87.6 122.5 113.1 110.4 158.9 143.1 850 88.6 123.2 113.1 111.3 159.3 143.1 860 89.6 123.9 113.1 112.1 159.6 143.1 870 90.5 124.6 113.1 113.0 160.0 143.1 880 91.5 125.3 113.1 113.8 160.3 143.1 890 92.4 125.9 113.1 114.6 160.7 143.2 900 93.4 126.6 113.1 115.4 161.0 143.9 910 94.3 127.2 113.1 116.1 161.4 144.6 920 95.1 127.9 114.0 116.9 161.7 145.3 930 96.0 128.5 115.0 117.7 162.0 146.0 940 96.9 129.1 115.9 118.4 162.4 146.7 950 97.7 129.7 116.8 119.1 162.7 147.3 960 98.6 130.3 117.8 119.9 163.0 148.0 970 99.4 130.9 118.7 120.6 163.4 148.6 980 100.2 131.5 119.5 121.3 163.7 149.3 990 101.0 132.1 120.4 122.0 164.0 149.9 1000 G-21

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-2. Bounding Dresden 2&3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6, & G-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (°F) (OF) 1010 101.7 132.7 121.2 122.6 164.3 150.6 1020 102.5 133.2 122.1 123.3 164.6 151.2 1030 103.3 133.8 122.9 124.0 165.0 151.8 1040 104.0 134.4 123.7 124.6 165.3 152.4 1050 104.7 134.9 124.5 125.3 165.6 153.0 1060 105.4 135.5 125.3 125.9 165.9 153.6 1070 106.2 136.0 126.1 126.5 166.2 154.2 1080 106.9 136.5 126.8 127.2 166.5 154.8 1090 107.6 137.1 127.6 127.8 166.8 155.3 1100 108.2 137.6 128.3 128.4 167.1 155.9 1105 108.6 137.8 128.7 128.7 167.3 156.2 1110 108.9 138.1 129.0 129.0 167.4 156.4 1120 109.6 138.6 129.7 129.6 167.7 157.0 1130 110.2 139.1 130.5 130.2 168.0 157.5 1140 110.9 139.6 131.2 130.7 168.3 158.1 1150 111.5 140.1 131.8 131.3 168.6 158.6 1160 112.1 140.6 132.5 131.9 168.9 159.1 1170 112.8 141.1 133.2 132.4 169.2 159.7 1180 113.4 141.6 133.8 133.0 169.5 160.2 1190 114.0 142.1 134.5 133.5 169.7 160.7 1200 114.6 142.5 135.1 134.1 170.0 161.2 1210 115.2 143.0 135.8 134.6 170.3 161.7 1220 115.8 143.5 136.4 135.2 170.6 162.2 1230 116.3 143.9 137.0 135.7 170.9 162.7 1240 116.9 144.4 137.6 136.2 171.2 163.2 1250 117.5 144.8 138.2 136.7 171.4 163.7 G-22

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-2. Bounding Dresden 2&3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-3, G-5, G-6, & G-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (OF) (OF) (OF) (°F) (°F)

(PSIG) (OF) 1260 118.0 145.3 138.8 137.2 171.7 164.2 1270 118.6 145.7 139.4 137.7 172.0 164.6 1280 119.1 146.2 140.0 138.2 172.2 165.1 1290 119.7 146.6 140.6 138.7 172.5 165.6 1300 120.2 147.0 141.1 139.2 172.8 166.0 1310 120.7 147.5 141.7 139.7 173.1 166.5 1320 121.3 147.9 142.3 140.2 173.3 166.9 1330 121.8 148.3 142.8 140.6 173.6 167.4 1340 122.3 148.7 143.4 141.1 173.8 167.8 1350 122.8 149.1 143.9 141.6 174.1 168.3 1360 123.3 149.6 144.4 142.0 174.4 168.7 1370 123.8 150.0 145.0 142.5 174.6 169.1 1380 124.3 150.4 145.5 142.9 174.9 169.6 1390 124.8 150.8 146.0 143.4 175.1 170.0 1400 125.3 151.2 146.5 143.8 175.4 170.4 G-23

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-11 and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY , 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (°F) (OF) 0 68.0 83.1 68.0 83.1 83.1 10 68.0 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 83.1 60 68.0 83.1 68.0 83.1 91.0 70 68.0 83.1 68.0 83.1 98.2 80 68.0 83.1 68.0 83.1 104.2 90 68.0 83.1 68.0 83.1 109.3 100 68.0 83.1 68.0 83.1 113.8 110 68.0 83.1 68.0 83.1 117.9 120 68.0 83.1 68.0 83.1 121.7 130 68.0 83.1 68.0 85.2 125.2 140 68.0 83.1 68.0 88.4 128.4 150 68.0 83.1 68.0 91.2 131.2 160 68.0 83.1 68.0 93.9 133.9 170 68.0 83.1 68.0 96.5 136.5 180 68.0 83.1 68.0 98.9 138.9 190 68.0 83.1 68.0 101.2 141.2 200 68.0 83.1 68.0 103.3 143.3 210 68.0 83.1 68.0 105.3 145.3 G-24

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0'F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-11 and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (°F) (°F) (OF),

220 68.0 83.1 68.0 107.3 147.3 230 68.0 83.1 68.0 109.1 149.1 240 68.0 83.1 68.0 110.9 150.9 250 68.0 83.1 68.0 112.6 152.6 260 68.0 83.1 68.0 114.2 154.2 270 68.0 83.1 68.0 115.8 155.8 280 68.0 83.1 68.0 117.3 157.3 290 68.0 83.1 68.0 118.8 158.8 300 68.0 83.1 68.0 120.2 160.2 310 68.0 83.1 68.0 121.5 161.5 312.5 68.0 83.1 68.0 121.9 161.9 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 68.0 113.1 68.0 143.1 183.1 380 68.0 113.1 68.0 143.1 183.1 390 68.0 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 113.1 68.0 143.1 183.1 420 68.0 113.1 68.0 143.1 183.1 G-25

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-1 1 and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) ("F) (OF) 430 68.0 113.1 68.0 143.1 183.1 440 68.0 113.1 68.0 143.1 183.1 450 68.0 113.1 68.0 143.1 183.1 460 68.0 113.1 68.0 143.1 183.1 470 68.0 113.1 68.0 143.1 183.1 480 68.0 113.1 68.0 143.1 183.1 490 68.0 113.1 68.0 143.1 183.1 500 68.0 113.1 68.0 143.1 183.1 510 68.0 113.1 68.0 143.1 183.1 520 68.0 113.1 68.2 143.2 183.2 530 68.0 113.1 70.2 144.0 184.0 540 68.0 113.1 72.1 144.8 184.8 550 68.0 113.1 73.9 145.6 185.6 560 68.0 113.1 75.7 146.4 186.4 570 68.0 113.1 77.4 147.1 187.1 580 68.0 113.1 79.0 147.9 187.9 590 68.0 113.1 80.6 148.6 188.6 600 68.0 113.1 82.2 149.1 189.1 610 68.0 113.1 83.7 149.6 189.6 620 68.0 113.1 85.1 150.0 190.0 630 68.0 113.1 86.5 150.4 190.4 640 68.0 113.1 87.9 150.8 190.8 650 68.0 113.1 89.2 151.2 191.2 G-26

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-1 1 and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) ("F) (OF) 660 68.0 113.1 90.5 151.7 191.7 670 68.0 113.1 91.8 152.1 192.1 680 68.0 113.1 93.1 152.5 192.5 690 68.0 113.1 94.3 152.9 192.9 700 69.2 113.1 95.4 153.3 193.3 710 70.7 113.1 96.6 153.7 193.7 720 72.1 113.1 97.7 154.1 194.1 730 73.5 113.3 98.8 154.5 194.5 740 74.8 114.1 99.9 154.9 194.9 750 76.1 115.0 101.0 155.2 195.2 760 77.4 115.8 102.0 155.6 195.6 770 78.6 116.6 103.0 156.0 196.0 780 79.8 117.3 104.0 156.4 196.4 790 81.0 118.1 105.0 156.8 196.8 800 82.2 118.9 105.9 157.1 197.1 810 83.3 119.6 106.9 157.5 197.5 820 84.4 120.4 107.8 157.9 197.9 830 85.5 121.1 108.7 158.2 198.2 840 86.5 121.8 109.6 158.6 198.6 850 87.6 122.5 110.4 158.9 198.9 860 88.6 123.2 111.3 159.3 199.3 870 89.6 123.9 112.1 159.6 199.6 880 90.5 124.6 113.0 160.0 200.0 G-27

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-11 and G-12

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

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (0F) (OF) (OF) (°F) 890 91.5 125.3 113.8 160.3 200.3 900 92.4 125.9 114.6 160.7 200.7 910 93.4 126.6 115.4 161.0 201.0 920 94.3 127.2 116.1 161.4 201.4 930 95.1 127.9 116.9 161.7 201.7 940 96.0 128.5 117.7 162.0 202.0 950 96.9 129.1 118.4 162.4 202.4 960 97.7 129.7 119.1 162.7 202.7 970 98.6 130.3 119.9 163.0 203.0 980 99.4 130.9 120.6 163.4 203.4 990 100.2 131.5 121.3 163.7 203.7 1000 101.0 132.1 122.0 164.0 204.0 1010 101.7 132.7 122.6 164.3 204.3 1020 102.5 133.2 123.3 164.6 204.6 1030 103.3 133.8 124.0 165.0 205.0 1040 104.0 134.4 124.6 165.3 205.3 1050 104.7 134.9 125.3 165.6 205.6 1060 105.4 135.5 125.9 165.9 205.9 1070 106.2 136.0 126.5 166.2 206.2 1080 106.9 136.5 127.2 166.5 206.5 1090 107.6 137.1 127.8 166.8 206.8 1100 108.2 137.6 128.4 167.1 207.1 1105 108.6 137.8 128.7 167.3 207.3 G-28

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-11 and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

, HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (°F) (OF) 1110 108.9 138.1 129.0 167.4 207.4 1120 109.6 138.6 129.6 167.7 207.7 1130 110.2 139.1 130.2 168.0 208.0 1140 110.9 139.6 130.7 168.3 208.3 1150 111.5 140.1 131.3 168.6 208.6 1160 112.1 140.6 131.9 168.9 208.9 1170 112.8 141.1 132.4 169.2 209.2 1180 113.4 141.6 133.0 169.5 209.5 1190 114.0 142.1 133.5 169.7 209.7 1200 114.6 142.5 134.1 170.0 210.0 1210 115.2 143.0 134.6 170.3 210.3 1220 115.8 143.5 135.2 170.6 210.6 1230 116.3 143.9 135.7 170.9 210.9 1240 116.9 144.4 136.2 171.2 211.2 1250 117.5 144.8 136.7 171.4 211.4 1260 118.0 145.3 137.2 171.7 211.7 1270 118.6 145.7 137.7 172.0 212.0 1280 119.1 146.2 138.2 172.2 212.2 1290 119.7 146.6 138.7 172.5 212.5 1300 120.2 147.0 139.2 172.8 212.8 1310 120.7 147.5 139.7 173.1 213.1 1320 121.3 147.9 140.2 173.3 213.3 1330 121.8 148.3 140.6 173.6 213.6 G-29

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-3. Bounding Dresden 2&3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-9, G-1 1 and G-12 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY ,32 EFPY 32 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 1340 122.3 148.7 141.1 173.8 213.8 1350 122.8 149.1 141.6 174.1 214.1 1360 123.3 149.6 142.0 174.4 214.4 1370 123.8 150.0 142.5 174.6 214.6 1380 124.3 150.4 142.9 174.9 214.9 1390 124.8 150.8 143.4 175.1 215.1 1400 125.3 151.2 143.8 175.4 215.4 G-30

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-4. Bounding Dresden 2&3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-l, G-2, G-4, G-5, G-6, & G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE CURVE A CURVE A CURVE A B CURVE B CURVE B (OF) (OF) (OF) (OF) (°F) (°F) 0 68.0 83.1 83.1 68.0 83.1 83.1 10 68.0 83.1 83.1 68.0 83.1 83.1 20 68.0 83.1 83.1 68.0 83.1 83.1 30 68.0 83.1 83.1 68.0 83.1 83.1 40 68.0 83.1 83.1 68.0 83.1 83.1 50 68.0 83.1 83.1 68.0 83.1 83.1 60 68.0 83.1 83.1 68.0 83.1 83.1 70 68.0 83.1 83.1 68.0 83.1 83.1 80 68.0 83.1 83.1 68.0 83.1 83.1 90 68.0 83.1 83.1 68.0 83.1 83.1 100 68.0 83.1 83.1 68.0 83.1 83.1 110 68.0 83.1 83.1 68.0 83.1 83.1 120 68.0 83.1 83.1 68.0 83.1 83.1 130 68.0 83.1 83.1 68.0 85.2 83.1 140 68.0 83.1 83.1 68.0 88.4 83.1 150 68.0 83.1 83.1 68.0 91.2 83.1 160 68.0 83.1 83.1 68.0 93.9 83.1 170 68.0 83.1 83.1 68.0 96.5 83.1 180 68.0 83.1 83.1 68.0 98.9 83.1 190 68.0 83.1 83.1 68.0 101.2 83.1 200 68.0 83.1 83.1 68.0 103.3 83.1 210 68.0 83.1 83.1 68.0 105.3 83.1 220 68.0 83.1 83.1 68.0 107.3 83.1 230 68.0 83.1 83.1 68.0 109.1 83.1 240 68.0 83.1 83.1 68.0 110.9 83.1 G-31

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G4. Bounding Dresden 2&3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6, & G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE CURVE A CURVE A CURVE A B CURVE B CURVE B PRESSURE (OF) (OF) (OF)

(PSIG) (°F) (OF) (OF) 68.0 83.1 83.1 68.0 112.6 83.1 250 68.0 83.1 83.1 68.0 114.2 83.1 260 68.0 83.1 83.1 68.0 115.8 83.1 270 68.0 83.1 83.1 68.0 117.3 83.1 280 68.0 83.1 83.1 68.0 118.8 83.1 290 68.0 83.1 83.1 68.0 120.2 83.1 300 68.0 83.1 83.1 68.0 121.5 83.1 310 68.0 83.1 83.1 68.0 121.9 83.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 312.5 68.0 113.1 113.1 68.0 143.1 143.1 320 68.0 113.1 113.1 68.0 143.1 143.1 330 68.0 113.1 113.1 68.0 143.1 143.1 340 68.0 113.1 113.1 68.0 143.1 143.1 350 68.0 113.1 113.1 68.0 143.1 143.1 360 68.0 113.1 113.1 68.0 143.1 143.1 370 68.0 113.1 113.1 68.0 143.1 143.1 380 68.0 113.1 113.1 68.0 143.1 143.1 390 68.0 113.1 113.1 68.0 143.1 143.1 400 68.0 113.1 113.1 68.0 143.1 143.1 410 68.0 113.1 113.1 68.0 143.1 143.1 420 68.0 113.1 113.1 68.0 143.1 143.1 430 113.1 113.1 68.0 143.1 143.1 440 68.0 68.0 113.1 113.1 68.0 143.1 143.1 450 68.0 113.1 113.1 68.0 143.1 143.1 460 68.0 113.1 113.1 68.0 143.1 143.1 470 68.0 113.1 113.1 68.0 143.1 143.1 480 68.0 113.1 113.1 68.0 143.1 143.1 490 G-32

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G4. Bounding Dresden 2&3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6, & G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (OF) (OF) (OF) (OF)

(PSIG) (OF) (°F) 500 68.0 113.1 113.1 68.0 143.1 143.1 510 68.0 113.1 113.1 68.0 143.1 143.1 520 68.0 113.1 113.1 68.2 143.2 143.1 530 68.0 113.1 113.1 70.2 144.0 143.1 540 68.0 113.1 113.1 72.1 144.8 143.1 550 68.0 113.1 113.1 73.9 145.6 143.1 560 68.0 113.1 113.1 75.7 146.4 143.1 570 68.0 113.1 113.1 77.4 147.1 143.1 580 68.0 113.1 113.1 79.0 147.9 143.1 590 68.0 113.1 113.1 80.6 148.6 143.1 600 68.0 113.1 113.1 82.2 149.1 143.1 610 68.0 113.1 113.1 83.7 149.6 143.1 620 68.0 113.1 113.1 85.1 150.0 143.1 630 68.0 113.1 113.1 86.5 150.4 143.1 640 68.0 113.1 113.1 87.9 150.8 143.1 650 68.0 113.1 113.1 89.2 151.2 143.1 660 68.0 113.1 113.1 90.5 151.7 143.1 670 68.0 113.1 113.1 91.8 152.1 143.1 680 68.0 113.1 113.1 93.1 152.5 143.1 690 68.0 113.1 113.1 94.3 152.9 143.1 700 69.2 113.1 113.1 95.4 153.3 144.0 710 70.7 113.1 113.1 96.6 153.7 145.1 720 72.1 113.1 113.1 97.7 154.1 146.0 730 73.5 113.3 113.1 98.8 154.5 147.0 740 74.8 114.1 113.1 99.9 154.9 148.0 750 76.1 115.0 113.1 101.0 155.2 148.9 760 77.4 115.8 113.1 102.0 155.6 149.8 G-33

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-4. Bounding Dresden 2&3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6, & G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (°F) (OF) (OF) (°F) (OF) (OF) 770 78.6 116.6 113.3 103.0 156.0 150.7 780 79.8 117.3 114.7 104.0 156.4 151.6 790 81.0 118.1 116.1 105.0 156.8 152.5 800 82.2 118.9 117.4 105.9 157.1 153.3 810 83.3 119.6 118.7 106.9 157.5 154.2 820 84.4 120.4 120.0 107.8 157.9 155.0 830 85.5 121.1 121.2 108.7 158.2 155.8 840 86.5 121.8 122.4 109.6 158.6 156.6 850 87.6 122.5 123.5 110.4 158.9 157.4 860 88.6 123.2 124.7 111.3 159.3 158.2 870 89.6 123.9 125.8 112.1 159.6 158.9 880 90.5 124.6 126.9 113.0 160.0 159.7 890 91.5 125.3 128.0 113.8 160.3 160.4 900 92.4 125.9 129.0 114.6 160.7 161.2 910 93.4 126.6 130.0 115.4 161.0 161.9 920 94.3 127.2 131.0 116.1 161.4 162.6 930 95.1 127.9 132.0 116.9 161.7 163.3 940 96.0 128.5 133.0 117.7 162.0 164.0 950 96.9 129.1 133.9 118.4 162.4 164.7 960 97.7 129.7 134.8 119.1 162.7 165.3 970 98.6 130.3 135.8 119.9 163.0 166.0 980 99.4 130.9 136.7 120.6 163.4 166.6 990 100.2 131.5 137.5 121.3 163.7 167.3 1000 101.0 132.1 138.4 122.0 164.0 167.9 1010 101.7 132.7 139.2 122.6 164.3 168.6 1020 102.5 133.2 140.1 123.3 164.6 169.2 1030 103.3 133.8 140.9 124.0 165.0 169.8 G-34

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE G-4. Bounding Dresden 2&3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6, & G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY HEAD VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 1040 104.0 134.4 141.7 124.6 165.3 170.4 1050 104.7 134.9 142.5 125.3 165.6 171.0 1060 105.4 135.5 143.3 125.9 165.9 171.6 1070 106.2 136.0 144.1 126.5 166.2 172.2 1080 106.9 136.5 144.8 127.2 166.5 172.8 1090 107.6 137.1 145.6 127.8 166.8 173.3 1100 108.2 137.6 146.3 128.4 167.1 173.9 1105 108.6 137.8 146.7 128.7 167.3 174.2 1110 108.9 138.1 147.0 129.0 167.4 174.4 1120 109.6 138.6 147.7 129.6 167.7 175.0 1130 110.2 139.1 148.5 130.2 168.0 175.5 1140 110.9 139.6 149.2 130.7 168.3 176.1 1150 111.5 140.1 149.8 131.3 168.6 176.6 1160 112.1 140.6 150.5 131.9 168.9 177.1 1170 112.8 141.1 151.2 132.4 169.2 177.7 1180 113.4 141.6 151.8 133.0 169.5 178.2 1190 114.0 142.1 152.5 133.5 169.7 178.7 1200 114.6 142.5 153.1 134.1 170.0 179.2 1210 115.2 143.0 153.8 134.6 170.3 179.7 1220 115.8 143.5 154.4 135.2 170.6 180.2 1230 116.3 143.9 155.0 135.7 170.9 180.7 1240 116.9 144.4 155.6 136.2 171.2 181.2 1250 117.5 144.8 156.2 136.7 171.4 181.7 1260 118.0 145.3 156.8 137.2 171.7 182.2 1270 118.6 145.7 157.4 137.7 172.0 182.6 1280 119.1 146.2 158.0 138.2 172.2 183.1 1290 119.7 146.6 158.6 138.7 172.5 183.6 G-35

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-4. Bounding Dresden 2&3 P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-1, G-2, G-4, G-5, G-6, & G-8 BOTTOM UPPER 54 EFPY BOTTOM UPPER 54 EFPY VESSEL BELTLINE HEAD CURVE VESSEL BELTLINE HEAD CURVE A CURVE A B CURVE B CURVE B PRESSURE CURVE A (OF) (OF) (OF) (OF)

(PSIG) (OF) (OF) 172.8 184.0 1300 120.2 147.0 159.1 139.2 147.5 159.7 139.7 173.1 184.5 1310 120.7 147.9 160.3 140.2 173.3 184.9 1320 121.3 148.3 160.8 140.6 173.6 185.4 1330 121.8 148.7 161.4 141.1 173.8 185.8 1340 122.3 149.1 161.9 141.6 174.1 186.3 1350 122.8 149.6 162.4 142.0 174.4 186.7 1360 123.3 150.0 163.0 142.5 174.6 187.1 1370 123.8 150.4 163.5 142.9 174.9 187.6 1380 124.3 150.8 164.0 143.4 175.1 188.0 1390 124.8 151.2 164.5 143.8 175.4 188.4 1400 125.3 G-36

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY CURVE A CURVE B CURVE B CURVE C PRESSURE CURVE A (OF) (OF) (OF) (OF)

(PSIG) 83.1 68.0 83.1 83.1 0 68.0 68.0 83.1 68.0 83.1 83.1 10 83.1 68.0 83.1 83.1 20 68.0 83.1 68.0 83.1 83.1 30 68.0 83.1 68.0 83.1 83.1 40 68.0 83.1 68.0 83.1 83.1 50 68.0 83.1 68.0 83.1 91.0 60 68.0 83.1 68.0 83.1 98.2 70 68.0 83.1 68.0 83.1 104.2 80 68.0 83.1 68.0 83.1 109.3 90 68.0 83.1 68.0 83.1 113.8 100 68.0 83.1 68.0 83.1 117.9 110 68.0 83.1 68.0 83.1 121.7 120 68.0 83.1 68.0 85.2 125.2 130 68.0 83.1 68.0 88.4 128.4 140 68.0 83.1 68.0 91.2 131.2 150 68.0 83.1 68.0 93.9 133.9 160 68.0 83.1 68.0 96.5 136.5 170 68.0 83.1 68.0 98.9 138.9 180 68.0 83.1 68.0 101.2 141.2 190 68.0 83.1 68.0 103.3 143.3 200 68.0 83.1 68.0 105.3 145.3 210 68.0 G-37

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY CURVE A CURVE A CURVE B CURVE B CURVE C PRESSURE (OF) (OF) (°F)

(PSIG) (OF) (OF) 68.0 83.1 68.0 107.3 147.3 220 68.0 83.1 68.0 109.1 149.1 230 68.0 83.1 68.0 110.9 150.9 240 68.0 83.1 68.0 112.6 152.6 250 68.0 83.1 68.0 114.2 154.2 260 68.0 83.1 68.0 115.8 155.8 270 68.0 83.1 68.0 117.3 157.3 280 83.1 68.0 118.8 158.8 290 68.0 68.0 83.1 68.0 120.2 160.2 300 68.0 83.1 68.0 121.5 161.5 310 68.0 83.1 68.0 121.9 161.9 312.5 68.0 113.1 68.0 143.1 183.1 312.5 68.0 113.1 68.0 143.1 183.1 320 68.0 113.1 68.0 143.1 183.1 330 68.0 113.1 68.0 143.1 183.1 340 68.0 113.1 68.0 143.1 183.1 350 68.0 113.1 68.0 143.1 183.1 360 68.0 113.1 68.0 143.1 183.1 370 113.1 68.0 143.1 183.1 380 68.0 68.0 113.1 68.0 143.1 183.1 390 113.1 68.0 143.1 183.1 400 68.0 113.1 68.0 143.1 183.1 410 68.0 68.0 113.1 68.0 143.1 183.1 420 G-38

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (°F) (OF) (OF) (OF) (OF) 430 68.0 113.1 68.0 143.1 183.1 440 68.0 113.1 68.0 143.1 183.1 450 68.0 113.1 68.0 143.1 183.1 460 68.0 113.1 68.0 143.1 183.1 470 68.0 113.1 68.0 143.1 183.1 480 68.0 113.1 68.0 143.1 183.1 490 68.0 113.1 68.0 143.1 183.1 500 68.0 113.1 68.0 143.1 183.1 510 68.0 113.1 68.0 143.1 183.1 520 68.0 113.1 68.2 143.2 183.2 530 68.0 113.1 70.2 144.0 184.0 540 68.0 113.1 72.1 144.8 184.8 550 68.0 113.1 73.9 145.6 185.6 560 68.0 113.1 75.7 146.4 186.4 570 68.0 113.1 77.4 147.1 187.1 580 68.0 113.1 79.0 147.9 187.9 590 68.0 113.1 80.6 148.6 188.6 600 68.0 113.1 82.2 149.1 189.1 610 68.0 113.1 83.7 149.6 189.6 620 68.0 113.1 85.1 150.0 190.0 630 68.0 113.1 86.5 150.4 190.4 640 68.0 113.1 87.9 150.8 190.8 650 68.0 113.1 89.2 151.2 191.2 G-39

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV &' BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY 54 EFPY CURVE A CURVE A CURVE B CURVE B CURVE C PRESSURE (OF) (OF)

(PSIG) (°F) (OF) (°F) 660 68.0 113.1 90.5 151.7 191.7 670 68.0 113.1 91.8 152.1 192.1 680 68.0 113.1 93.1 152.5 192.5 690 68.0 113.1 94.3 152.9 192.9 700 69.2 113.1 95.4 153.3 193.3 710 70.7 113.1 96.6 153.7 193.7 720 72.1 113.1 97.7 154.1 194.1 730 73.5 113.3 98.8 154.5 194.5 740 74.8 114.1 99.9 154.9 194.9 750 76.1 115.0 101.0 155.2 195.2 760 77.4 115.8 102.0 155.6 195.6 770 78.6 116.6 103.0 156.0 196.0 780 79.8 117.3 104.0 156.4 196.4 790 81.0 118.1 105.0 156.8 196.8 82.2 118.9 105.9 157.1 197.1 800 810 83.3 119.6 106.9 157.5 197.5 84.4 120.4 107.8 157.9 197.9 820 85.5 121.2 108.7 158.2 198.2 830 840 86.5 122.4 109.6 158.6 198.6 850 87.6 123.5 110.4 158.9 198.9 860 88.6 124.7 111.3 159.3 199.3 870 89.6 125.8 112.1 159.6 199.6 880 90.5 126.9 113.0 160.0 200.0 G-40

GE Nuclear Energy GE-NE-0000-0002-9600-01a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY 54 EFPY PRESSURE CURVE A CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 890 91.5 1280 113.8 160.4 200.4 900 92.4 129.0 114.6 161.2 201.2 910 93.4 130.0 115.4 161.9 201.9 920 94.3 131.0 116.1 162.6 202.6 930 95.1 132.0 116.9 163.3 203.3 940 96.0 133.0 117.7 164.0 204.0 950 96.9 133.9 118.4 164.7 204.7 960 97.7 134.8 119.1 165.3 205.3 970 98.6 135.8 119.9 166.0 206.0 980 99.4 136.7 120.6 166.6 206.6 990 100.2 137.5 121.3 167.3 207.3 1000 101.0 138.4 122.0 167.9 207.9 1010 101.7 139.2 122.6 168.6 208.6 102.5 140.1 123.3 169.2 209.2 1020 1030 103.3 140.9 124.0 169.8 209.8 1040 104.0 141.7 124.6 170.4 210.4 1050 104.7 142.5 125.3 171.0 211.0 1060 105.4 143.3 125.9 171.6 211.6 1070 106.2 144.1 126.5 172.2 212.2 1080 106.9 144.8 127.2 172.8 212.8 1090 107.6 145.6 127.8 173.3 213.3 1100 108.2 146.3 128.4 173.9 213.9 1105 108.6 146.7 128.7 174.2 214.2 G-41

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY 54 EFPY CURVE A CURVE A CURVE B CURVE B CURVE C PRESSURE (OF) (OF) (°F) (OF) (OF)

(PSIG) 108.9 147.0 129.0 174.4 214.4 1110 109.6 147.7 129.6 175.0 215.0 1120 110.2 148.5 130.2 175.5 215.5 1130 110.9 149.2 130.7 176.1 216.1 1140 111.5 149.8 131.3 176.6 216.6 1150 112.1 150.5 131.9 177.1 217.1 1160 112.8 151.2 132.4 177.7 217.7 1170 113.4 151.8 133.0 178.2 218.2 1180 114.0 152.5 133.5 178.7 218.7 1190 114.6 153.1 134.1 179.2 219.2 1200 115.2 153.8 134.6 179.7 219.7 1210 115.8 154.4 135.2 180.2 220.2 1220 116.3 155.0 135.7 180.7 220.7 1230 116.9 155.6 136.2 181.2 221.2 1240 117.5 156.2 136.7 181.7 221.7 1250 118.0 156.8 137.2 182.2 222.2 1260 118.6 157.4 137.7 182.6 222.6 1270 119.1 158.0 138.2 183.1 223.1 1280 119.7 158.6 138.7 183.6 223.6 1290 120.2 159.1 139.2 184.0 224.0 1300 120.7 159.7 139.7 184.5 224.5 1310 121.3 160.3 140.2 184.9 224.9 1320 121.8 160.8 140.6 185.4 225.4 1330 G-42

GE Nuclear Energy GE-NE-0000-0002-9600-01 a TABLE G-5. Bounding Dresden 2&3 Composite P-T Curve Values for 54 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A for Figures G-10, G-13 and G-14 BOTTOM UPPER RPV & BOTTOM UPPER RPV & UPPER RPV &

HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 54 EFPY 54 EFPY 54 EFPY CURVE A CURVE A CURVE B CURVE B CURVE C PRESSURE (OF)

(PSIG) (OF) (°F) (OF) (OF) 1340 122.3 161.4 141.1 185.8 225.8 122.8 161.9 141.6 186.3 226.3 1350 123.3 162.4 142.0 186.7 226.7 1360 123.8 163.0 142.5 187.1 227.1 1370 124.3 163.5 142.9 187.6 227.6 1380 124.8 164.0 143.4 188.0 228.0 1390 125.3 164.5 143.8 188.4 228.4 1400 G-43

RS-03-039, Request for Changes Related to TS Section 3.4.9, Reactor Coolant System Pressure & Temperature (P/T) Limits, Section 1.0 - Appendix G

Contents

Text

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

REFERENCES:

REFERENCES:

Reference:

REFERENCES:

REFERENCES:

Navigation menu

RS-03-039, Request for Changes Related to TS Section 3.4.9, Reactor Coolant System Pressure & Temperature (P/T) Limits, Section 1.0 - Appendix G

Text

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

REFERENCES:

REFERENCES:

Reference:

REFERENCES:

REFERENCES:

Navigation menu

Search