ML060760392

Response to Request for Additional Information - License Amendment Request, Proposed Changes to Extend the Use of Pressure-Temperature Limits Specified in Tech Spec Figures 3.4.9-1, 3.4.9-2 and 3.4.9-3.

ML060760392
Person / Time
Site:	Peach Bottom
Issue date:	03/15/2006
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC7519
Download: ML060760392 (59)

Text

Exelon Vucleat www exeloncorp corn T zoo Exelori Way KeiineIt Square, PA 19348 10 CFR 50.90 March 15, 2006 U S . Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Unit 3 Renewed Facility Operating License No. DPR-56 NRC Docket No. 50-278

Subject:

Response to Request for Additional Information - License Amendment Request, "Proposed Changes to Extend the Use of Pressure-Temperature Limits Specified in Technical Specifications Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3" (TAC No.

MC7519)

References:

(1) Letter from P. B. Cowan, Exelon Generation Company, LLC, to U. S. Nuclear Regulatory Commission, License Amendment Request, "Proposed Changes to Extend the Use of Pressure-Temperature Limits Specified in Technical Specifications Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3," dated July 6, 2005.

(2) Letter from R. V. Guzman, U. S. Nuclear Regulatory Commission, to C. M.

Crane, Exelon Generation Company, LLC, "Peach Bottom Power Station Unit No. 3 - Request for Additional Information (RAI) Regarding Proposed Pressure-TemperatureCurves (TAC No. MC7519)," dated January 26, 2006.

In Reference 1, Exelon Generation Company, LLC (Exelon), requested a change to Renewed Facility Operating License No. DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Unit

3. The proposed change would allow for extension of the use of the current Pressure-Temperature (P-T) limit curves specified in Technical Specifications (TS) Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 to 32 effective full power years (EFPY).

In Reference 2, the NRC requested additional information concerning the PBAPS, Unit 3 License Amendment Request (LAR). The attachment to this letter restates the NRC questions and provides Exelon's response to each question.

Exelon has concluded that the information provided in this response does not impact the conclusions of the: (1) Technical Analysis, (2) No Significant Hazards Consideration under the standards set forth in 10 CFR 50.92(c), or (3) Environmental Consideration as provided in the original submittal (Reference 1).

Response to Request for Additional Information PBAPS Unit 3 P-T Curve LAR Docket No. 50-278 March 15, 2006 Page 2 Enclosures 1 and 2 to this letter provide information from two versions of the same General Electric (GE) report. Enclosure 1 to this letter provides information from GE Report GE-NE-613-02119-00-01a, "Pressure-Temperature Curves for Exelon, Peach Bottom Unit 3," dated February 2002, which is the non-proprietary version of the report. Enclosure 2 to this letter provides information from GE Report GE-NE-B13-02119-00-01, which GE considers to contain proprietary information as defined in 10 CFR 2.390. The proprietary information is identified by a vertical bar in the margin. In each case, the information identified by the vertical bar in the margin is considered "trade secrets" and exempt from disclosure in accordance with the requirements of 10 CFR 2.390(a)(4). Accordingly, GE requests that the proprietary information in Enclosure 2 be withheld from public disclosure in accordance with the requirements of 10 CFR 2.390. An affidavit certifying the basis for this request for withholding, as required by 10 CFR 2.390(b)(l), is provided in Enclosure 3. The non-proprietary version of the information provided in Enclosure 2, which has the proprietary information removed, is included in . The portions of the information that have been removed are indicated by a vertical bar in the margin.

There are no regulatory commitments contained within this letter. If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 15th day of March 2006.

Respectfully, Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Response to Request for Additional Information : Excerpts from GE Report GE-NE-613-02119-00-01a "on-Proprietary Information] : Excerpts from GE Report GE-NE-B13-02119-00-01 [Proprietary Information] : GE Affidavit cc: Regional Administrator - NRC Region I w/ attachments NRC Senior Resident Inspector - PBAPS ,,

NRC Project Manager, NRR - PBAPS (6 Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection w/o Enclosure 2

ATTACHMENT PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 DOCKET NO. 50-278 PROPOSED CHANGES TO EXTEND THE USE OF PRESSURE-TEMPERATURE LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS FIGURES 3.4.9-1, 3.4.9-2, AND 3.4.9-3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION

Page 1 of 3 ATTACHMENT Peach Bottom Atomic Power Station, Unit 3 Docket No. 50-278 Proposed Changes To Extend the Use of Pressure-Temperature Limits Specified in Technical Specifications Figures 3.4.9-1, 3.4.9-2, And 3.4.9-3 Response to Request for Additional Information In Reference 1, Exelon Generation Company, LLC (Exelon), requested a change to Renewed Facility Operating License No. DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Unit

3. The proposed change would allow for extension of the use of the current Pressure-Temperature (P-T) limit curves specified in Technical Specifications (TS) Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3 to 32 effective full power years (EFPY).

In Reference 2, the NRC requested additional information concerning the PBAPS, Unit 3 License Amendment Request (LAR). Each NRC question is restated below followed by our response.

Enclosures 1 and 2 to this letter provide information from two versions of the same General Electric (GE) report. Enclosure 1 to this letter provides information from GE Report GE-NE-B13-02119-00-01a, "Pressure-Temperature Curves for Exelon, Peach Bottom Unit 3," dated February, 2002, which is the non-proprietary version of the report. Enclosure 2 to this letter provides information from GE Report GE-NE-B13-02119-00-01, which GE considers to contain proprietary information as defined in 10 CFR 2.390. The proprietary information is identified by a vertical bar in the margin. In each case, the information identified by the vertical bar in the margin is considered "trade secrets" and exempt from disclosure in accordance with the requirements of 10 CFR 2.390(a)(4). Accordingly, GE requests that the proprietary information in Enclosure 2 be withheld from public disclosure in accordance with the requirements of 10 CFR 2.390. An affidavit certifying the basis for this request for withholding, as required by 10 CFR 2.390(b)(1), is provided in Enclosure 3. The non-proprietary version of the information provided in Enclosure 2, which has the proprietary information removed, is included in . The portions of the information that have been removed are indicated by a vertical bar in the margin.

Question 1.

"Please provide the adjusted reference temperature (ART) calculations for the Peach Bottom Atomic Power Station (Peach Bottom) Unit No. 3 beltline materials at the 1/4T locations of the reactor pressure vessel based on the calculated neutron fluence values for these locations at 32 effective full-power years."

Response

The PBAPS, Unit 3 ART calculations used by General Electric (GE) to develop the latest (i.e.,

unapproved) P-T curves are provided in Enclosure 1. This information is excerpted from GE Report GE-NE-B13-02119-00-01a, Section 4.2, "Adjusted Reference Temperature for Beltline."

This information provides the basis for the calculations and summarizes the results of the ART calculations at the 1/4T location for the calculated fluence at 32 EFPY for all beltline materials, including plate and weld materials. As noted in Section 4.2.1.2 of the enclosed material,

Response to Request for Additional Information Attachment PBAPS Unit 3 P-T Curve LAR Page 2 of 3 extremely conservative fluence values were used to develop the ART calculations. This enclosure also contains Section 6.0, "References," of the GE report which provides the list of documents referred to in Section 4.2 of the report. This information is considered by GE to be non-proprietary.

Question 2.

"Please provide the pressure-temperature (P-T) calculations over the entire temperature range for the unapproved P-T curves that were based on the KIC equation in Section G-2110 of Appendix G to Section XI of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2001 Edition. Provide the adjusted pressure value for each temperature value assessed, based on the limiting ART (limiting RTNDT) values for the Peach Bottom Unit No. 3 reactor vessel. Include all parameters used in the calculation (e.g., KIT values, temperature gradients across the wall, KIC values, and any margin included for pressure and/or temperature measurement uncertainty)."

Response

The P-T calculations for the unapproved P-T curves are provided in Enclosure 2. This enclosure contains Section 4.3, "Pressure-Temperature Curve Methodology," of GE Report GE-NE-B13-02119-00-01. The information in Enclosure 2 provides the basis and methodology used for development of the curves, considering all regions of the reactor vessel. This enclosure contains information which GE considers to be proprietary. The non-proprietary version of Section 4.3, which has the proprietary information removed, is provided in Enclosure 1.

Additionally, Enclosure 1 includes Tables B-1 and B-2 from GE Report GE-NE-B13-02119 01a. These tables contain the results of the P-T calculations over the entire pressure range of operations for the reactor vessel. These tables are considered by GE to be non-proprietary.

Question 3.

"Please confirm that the P-T curves (as provided in Peach Bottom Unit No. 3 Technical Specifications, Figure 3.4.9-1, Temperature/Pressure Limits for Inservice Hydrostatic and Inservice Leakage Tests, Figure 3.4.9-2, Temperature/Pressure Limits for Non-Nuclear Heatup and Cooldown Following a Shutdown, and Figure 3.4.9-3, Temperature/Pressure Limits for Criticality, which were approved in the Peach Bottom Amendment No. 250), are based on the 1/4T location calculations for cooldown. Also, confirm that these curves are based on the limiting ART for the 1/4T location of the vessel."

Response

The P-T curves, as provided in PBAPS, Unit 3, TS Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3, which were approved in PBAPS, Unit 3, Amendment No. 250, are based on the 1/4T location for cooldown. Additionally, it is confirmed that the curves are based on the limiting ART value for the 1/4T location of the vessel.

Response to Request for Additional Information Attachment PBAPS Unit 3 P-T Curve LAR Page 3 of 3

References:

(1) Letter from P. B. Cowan, Exelon Generation Company, LLC, to U. S. Nuclear Regulatory Commission, License Amendment Request, "Proposed Changes to Extend the Use of Pressure-Temperature Limits Specified in Technical Specifications Figures 3.4.9-1, 3.4.9-2, and 3.4.9-3," dated July 6, 2005.

(2) Letter from R. V. Guzman, U. S. Nuclear Regulatory Commission, to C. M. Crane, Exelon Generation Company, LLC, "Peach Bottom Power Station Unit No. 3 - Request for Additional Information (RAI) Regarding Proposed Pressure-Temperature Curves (TAC No. MC7519)," dated January 26, 2006.

ENCLOSURE 1 PEACH BOTTOM ATOMIC POWER STATION, UNlT 3 DOCKET NO. 50-278 PROPOSED CHANGES TO EXTEND THE USE OF PRESSURE-TEMPERATURE LIMITS SPECIFIED IN TECHNICAL SPEClFlCATlONS FIGURES 3.4.9-1 ,'3.4.9-2, AND 3.4.9-3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION GENERAL ELECTRIC (GE) REPORT GE-NE-B13-02119-00-01a, "PRESSURE-TEMPERATURE CURVES FOR EXELON, PEACH BOrrOM UNlT 3" SECTIONS 4.2,4.3 AND 6.0; TABLES B-1 AND B-2 NON-PROPRIETARY INFORMATION The information in this enclosure i s from the non-proprietary version of the document GE-NE-B13-02119-00-01, which has the proprietary information removed. The portions that have been removed are indicated by a vertical bar i n the margin.

GENERAL ELECTRIC (GE) REPORT GE-NE-BI 3-02119-00-01a, "PRESSURE-TEMPERATURECURVES FOR EXELON, PEACH BOTTOM UNIT 3" SECTION 4.2

GE Nuclear Energy Non-Proprietary Version 4.2 ADJUSTED REFERENCE TEMPERATURE FOR BE1TLlNE 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 made and summarized in Table 4-4 for 32 EFPY and Table 4-5 for 54 EFPY.

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

SHIFT = ARTNDT+ Margin where, lilRTNDT= [CFI*~ (0.28 - 0.10 log 9 Margin = 2(c? + 0A2)0'5 CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = %T fluence 1 10" Margin = 2(0? + G,~)~.~

0, = standard deviation on initial RTNDT1 which is taken to be 0°F (16.4"F for electroslag welds).

GA = standard deviation on ARTNDT,28OF for welds and 17°F for base material, except that CT, need not exceed 0.50 times the ARTNDTvalue.

ART = Initial RTNDT+ SHIFT The margin term 0, has constant values in Rev 2 of 17°F for plate and 28°F for weld.

However, CTA need not be greater than 0.5 ARTNDT.Since the GElBWROG method of estimating RTNDT operates on the lowest Charpy energy value (as described in Section 4.1.2) and provides a conservative adjustment to the 50 ft-lb. level, the value of

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version ol is taken to be 0°F for the vessel plate and most weld materials, except that a1 is assumed to be 16.4OF for the beltline electroslag weld materials.

Chemistry The vessel beltline chemistries were obtained from sources including CMTRs [ I 21 and an NRC RAI submittal [ I 31. 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.Iof Rev 2 for welds and plates, respectively. As discussed in Section 4.1.2, best estimates results are used for the beltline electroslag for the Initial RTNDT

[13], therefore, the standard deviation (0,) is specified.

Fluence A bounding Limerick and Peach Bottom flux for the vessel ID wall [I41 is calculated using methods consistent with Regulatory Guide 1.190. The flux in Reference 14 is determined for the currently licensed power of 3458 MWt and is conservatively used from the beginning to the end of the licensing period (i.e., 32 and 54 EFPY). Even using the conservative flux from Reference 14 the P-T curves are only beltline limited above 1230 psig for curve A and 1290 psig for curve B for 32 EFPY. The P-T curves are beltline limited above 830 psig for curve A and 890 psig for curve B for 54 EFPY.

The peak fast flux for the RPV inner surface from Reference 14 is 1.32e9 n/cm2-s. The peak fast flux for the RPV inner surface determined from surveillance capsule flux wires removed during the outage following Fuel Cycle 7 at a full power of 3293 M w is 7.16e8 n/cm2-s[I]. Linearly scaling the Reference 1 flux by 1.05 to the currently licensed power of 3458 MWt results in an estimated flux of 7.52e8 n/cm2-s. Therefore, the Reference 14 flux bounds the flux determined from the surveillance capsule flux wire results by 76%.

The time period 32 EFPY is 1.Ole9 sec, therefore the RPV ID surface fluence is as follows: RPV ID surface fluence = I.32e9 nlcm2-s*l.Ole9 s = 1.33e18 nlcm2. This fluence of 1.33e18 nlcm2applies to Shell #2 and the Vertical Welds for Shell #2.

GE Nuclear Energy GE-NE-BI3-02 119-00-0 1a Non-Proprietary Version As shown in Reference 22 the elevation of the girth welds DE and EF are 23.9" above BAF and 138.69" above BAF, respectively. Using Figure 3-2 of Reference 14, the relative flux at 25" above BAF and 138" above BAF is 0.64. Therefore, the fluence for the girth welds and the Shell # I and #3 welds and plates can be reduced from the peak fluence by a ratio of 0.64. Therefore, the ID fluence for the girth welds and the Shell # I and #3 welds and plates will be 8.53e17 n/cm2and the 1/4T fiuence will be 5.9e17 n/cm2.

The fluence value used in this report for a power level of 3458 MWt also bounds the fluence value for a thermal optimization power (TPO) level of 3517 MWt.

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 RTNDTlchemistryl and fluence as inputs, Rev 2 was applied to compute ART. Table 4-4 lists values of beltline ART for 32 EFPY and Table 4-5 lists the values for 54 EFPY.

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version Table 4-4: Peach Bottom Unit 3 Beltline ART Values (32 EFPY)

Shell #2 Plates and Vertical Welds Thickness in inches = 6.125 Ratio at Location IPeak = 1 00 32 EFPY Peak I.D. fluence = 1 3E+18 nlcmA2 32 EFPY Peak 114 T fluence = 9.2E+17 nlcmA2 32 EFPY Peak 114 T fluence = 9.2E+17 nlcmA2 Shell #l Plates &Welds, Shell #3 Plates & Welds, and Girth Welds Thickness in inches= 6,125 Ratio at Location IPeak = 0.64 32 EFPY Peak I 0.fluence = 1.3E+18 n/cmA2 32 EFPY Peak 114 T fluence = 9.2E+17 nlcmA2 32 EFPY at Location 114 T fluence = 5.9E+17 nlcmA2 Initial 114 T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OR HEATILOT %Cu %Ni CF RTndt Fluence A RTndt ol a, Margin Shift ART "F nlcmA2 "F "F "F "F PLATES:

Shell #ILower 6-146-1 C4689-2 0.12 0.56 82 -10 5.9E+17 26 0 13 26 53 43 6-146-3 C4684-2 0.13 0.58 90 -20 5.9E+17 29 0 14 29 58 38 6-146-7 C4627-1 0.12 0.57 82 -20 5.9E+17 26 0 13 26 53 33 Shell #2 Lower-Inter 6-139-10 C2773-2 0.15 0.49 104 10 9.2E+17 42 76 0 17 34 86 6-139-11 C2775-1 0.13 0.46 87 10 9.2E+17 35 0 17 34 69 79 6-139-12 C3103-1 0.14 0.60 100 10 9.2E+17 40 0 17 34 74 84 Shell #3 Intermediate 6-146-5 C4608-1 0.12 0.55 82 10 5.9E+17 26 0 13 26 52 62 6-146-4 C4689-1 0.12 0.56 82 10 5.9E+17 26 0 13 26 53 63 6-146-2 C4654-1 0.11 0.55 74 10 5.9E+17 24 0 12 24 47 57 WELDS:

Vertical Weld Shell # l Seam Dl, 02, D3 37C065 0.182 0.181 94.5 -45 5.9E+17 30 16.4 15 45 75 30 Shell #2 Seam E l , E2, E3 37C065 0.182 0.181 94.5 45 9.2E+f7 38 16.4 19 50 88 43 Shell #3 Seam F1, F2, F3 37C065 0.182 0.181 94.5 -45 5.9E+17 30 16.4 15 45 75 30 Girth Shell 1 to 2 - Lower to 3P4000, Linde 124 Lower-IntermediateDE Flux Lot 3932 0.020 0.934 27 -50 5.9E+17 9 0 4 9 17 -33 Shell 2 to 3 Lower-Inter lP4217, Linde 124 to IntermediateEF Flux Lot 3929 0.102 0.942 137 -50 5.9E+17 44 0 22 44 88 38 Max ART 86

GENERAL ELECTRIC (GE) REPORT GE-NE-BI 3-02119-00-01a, "PRESSURE-TEMPERATURECURVES FOR EXELON, PEACH BOTTOM UNIT 3" SECTION 4.3

GE Nuclear Energy G E-NE-BI3-02119-00-01a Non-Proprietary Version 4.3 PRESSURE-TEMPERA TURE CURVE MTHODOLOGY 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 Appendix G [8] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions that a pressure-retaining component may be subjected to over its service lifetime. The ASME Code (Appendix G of Section XI of the ASME Code [6])forms the basis for the requirements of 10CFRSO 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]:

0 Closure flange region (Region A) 0 Core beltline region (Region B) 0 Upper vessel (Regions A & B) 0 Lower vessel (Regions B & C)

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

For the core not critical and the core critical curves, the P-T curves specify a coolant heatup and cooldown temperature rate of 1OO"F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the GE Nuclear Energy GE-NE-513-02119-QQ-Qla Non-Proprietary Version nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 2Q"F/hror 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, Klr, at 114T 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 heatupkooldown curve limits.

The applicable temperature is the greater of the 10CFR50 Appendix G minimum temperature requirement or the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-6:

GE Nilclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version Table 4-6: Summary of the 10CFR50 Appendix G Requirements Operating Condition and Pressure Minimum Temperature Requirement I I. Hydrostatic Pressure Test & Leak Test

1. At 5 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

1. At 5 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 5 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 f

hydrostatic pressure test

• 60°F adder is included by GE as an additional conservatism as discussed in Section 4.3.2.3 There are four vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the two regions in the remainder of the vessel (i.e., the upper vessel and lower vessel non-beltline regions). The closure flange region limits are controlling at lower pressures primarily because of 10CFR50 Appendix G [8]

requirements. The non-beltline and beltline region operating limits are evaluated according to procedures in 10CFR50 Appendix G [8], ASME Code Appendix G [6], and Welding Research Council (WRC) Bulletin 175 [I 51. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

GE Nuclear Energy GE-NE-513-02119-00-01 a Non-Proprietary Version 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 (4 .OE17 n/cm2)to cause any significant shift of RTNDT.Non-beltline components include nozzles (see Appendix E),

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

Detailed stress analyses of the non-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include IOO"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 regions are categorized under one of these two components as described in Tables 4-7 and 4-8.

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version Table 4-7: Applicable BWR/4 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B I

I CRD HYD System Return I I Stabilizer Brackets Shroud Support Attachments Core AP and Liquid Control Nozzle I

Steam Water Interface Jet Pump Instrumentation Nozzle Shell CRD and Bottom Head (B only)

Top Head Nozzles (B only) A I Recirculation Outlet Nozzle (B only) I Table 4-8: Applicable BWR/4 Discontinuity Components for Use with CRD (Bottom Head) Curves A&B

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

The P-T curves for the non-beltline region were conservatively developed for a large BWR/6 (nominal inside diameter of 251 inches). The analysis is considered appropriate for Peach Bottom Unit 3 as the plant specific geometric values are bounded by the

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version generic analysis for a large BWR/6, as determined in Section 4.3.2.1 .Ithrough Section 4.3.2.1.4. The generic value was adapted to the conditions at Peach Bottom Unit 3 by using plant specific RTNDTvalues 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.

4.3.2.7.7 -

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

In a finite element analysis , the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K,. The evaluation was modified to consider the new requirement for M, as discussed in ASME Code Section XI Appendix G [6] and shown below. The results of that computation were K, = 143.6 ksi-in1I2for 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.

The limit for the coolant temperature change rate is 20"F/hr or less.

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version The value of M, for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

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

M, = 1.85 for A 5 2 M, = 0.926 4 for 2 5 4 5 3 . 4 6 4 = 2.6206 M, = 3.21 for >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [6] and Klbis calculated from the equation in Paragraph G-2214.2 161:

KI, = M,

• opm = ksi-in2 Klb = (213) M, . Gpb = ksi-in*

The total KI is therefore:

KI = 1.5 (Kim+K t b ) + M m . (osm + (213)

• CT,~)= 143.6 ksi-in This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T - RTNoT)for a specific KI is based on the KI, the equation of Paragraph A-4200 in ASME Appendix A [I 71:

(T - RTNoT)= In [(K, - 33.2) 1 20.7341 10.02 CE Nuclear Energy GE-NE-613-02119-00-01a Non-Proprietary Version (T - RTNDT)= In [(I44 - 33.2) / 20.7341 / 0.02 (T - RTNDT) = 84°F The generic curve was generated by scaling 143.6 ksi-in2 by the nominal pressures and calculating the associated (T - RTNDT):

The highest RTNDT for the bottom head plates and welds is 42"F, as shown in Tables 4-1 and 4-3.

I GF Nuclear Fnargy GE-NE-B13-02119-90-01a Non-Proprietary Version Second, the P-T curve is dependent on the calculated KI value, and the KI value is proportional to the stress and the crack depth as shown below:

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

Generic: R / (t)12= 138 / (8)2= 49 inch (4-2)

The Peach Bottom Unit3 specific bottom head dimensions are R = 125.5 inches and t =8 inches minimum [IS], resulting in:

Peach Bottom Unit 3 specific: R I (t) = 125.5 / (8)IF2= 44.4 inch (4-3)

Since the generic v a l u e d Rl(t) is larger, the generic P-T curve is conservative when applied to the Peach Bottom Unit 3 bottom head.

GE Nuclear Energy GE-NE-Bl3-02119-00-01a Non-Proprietary Version 4.3.2.1.2 -

Core Not Critical Heatup/Cooldown Non-Beltline Curve B (Using Bottom Head)

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

Heatuplcooldown limits were calculated by increasing the safety factor in the pressure testing stresses (Section 4.3.2.1. I ) from 1.5 to 2.0.

The calculated value of KI 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 KI value for the core not critical condition is (143.6 / 1.5) . 2.0 = 191.5 ksi-in.

Therefore, the method to solve for (T - RTNDT) for a specific KIis based on the KI, equation of Paragraph A-4200 in ASME Appendix A 1171for the core not critical curve:

(T - RTNDT) = In [(KI - 33.2) / 20.7341 / 0.02 GE Nuclear Energy GE-NE-B13-02119-QQ-01 a Non-Proprietary Version (T - RTNDT) = In [( 191.5- 33.2) / 20.7341/ 0.02 (T - RTNDT) = 102°F The generic curve was generated by scaling 192 ksi-in2 by the nominal pressures and calculating the associated (T - RTNDT):

800 98 57 600 74 33 400 49 -14 The highest RTND~ for the bottom head plates and welds is 42F, as shown in Tables 4-1 and 4-3.

As discussed in Section 4-3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Tables 4-7, 4-8, and Appendix A}. With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than those of the other components being bounded. Therefore, for heatuplcooldown conditions, the CRD penetration provides bounding limits.

GE Nuclear Energy GE NE B13 02119-00-01a Non-Proprietary Version

~

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version 4.3.2.1.3 -

Pressure Test Non-Beltline Curve A (Using Feedwater NozzleAJpper Vessel Region)

The stress intensity factor, K,, for the feedwater nozzle was computed using the methods from WRC 175 [I 51 together with the nozzle dimension for a generic 251-inch BWR/6 feedwater nozzle. The result of that computation was KI = 200 ksi-in for an applied pressure of 1563 psig preservice hydrotest pressure.

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

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

Vessel Radius, R, 126.7 inches Vessel Thickness, t, 6.1875 inches Vessel Pressure, P, 1563 psig Pressure stress: S I = 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 CJ = 34.97 ksi. The factor F (ah,) from Figure A5-1 of WRC-175 is 1.4 where :

a = % ( t, + t, 2)12 =2.36 inches t, = thickness of nozzle = 7.125 inches t, = thickness of vessel = 6.1875 inches r, = 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, a/r, = 2.36 / 7.09 = 0.33. The value F(a/r,,), taken from Figure A5-1 of WRC Bulletin 175 for an ah, of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K,,is 1.5 CJ (jca)I2 F(a/r,):

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version Nominal KI = 1.5 34.97 . (n . 2.36)2 . 1.4 = 200 ksi-in2 The method to solve for (T - RTNDT) for a specific K, is based on the K,, equation of Paragraph A-4200 in ASME Appendix A [ 171for the pressure test condition:

(T - RTNDT)= In [(KI - 33.2) / 20.7341 I0.02 (T - RTNDT)= In [(200 - 33.2) / 20.7341 / 0.02 (T - RTNDT)= 104.2F The generic pressure test P-T curve was generated by scaling 200 ksi-in2 by the nominal pressures and calculating the associated (T - RTNDT),

GE Nuclear Energy GE-NE-813-02119-00-01a Non-Proprietary Version The highest RTNDTfor the feedwater nozzle materials is 40°F as shown in Table 4-2.

However, the R T N ~wasT increased to 44°F to consider the stresses in the top head nozzle together with the initial RTNDTas described below. The generic pressure test P-T curve is applied to the Peach Bottom Unit 3 feedwater nozzle curve by shifting the P vs. (T - RTNDT)values above to reflect the RTNDT value of 44°F.

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version Second, the P-T curve is dependent on the KI value calculated. The Peach Bottom Unit 3 specific vessel shell and nozzle dimensions applicable to the feedwater nozzle location [I91 and KI are shown below:

Vessel Radius, R, 125.7 inches Vessel Thickness, t, 6.125 inches Vessel Pressure, P, 1563 psig Pressure stress: CT = 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 0= 35.04 ksi. The factor F (ahn)from Figure A5-1 of WRC-175 is determined where:

a = 1/4 ( t, + t, 2)"2 =2.32 inches tn = thickness of nozzle = 6.963 inches t, = thickness of vessel = 6.125 inches rn = apparent radius of nozzle = r, + 0.29 rc=6.91 inches 4 = actual inner radius of nozzle = 6.0375 inches rc = nozzle radius (nozzle corner radius) = 3.0 inches Thus, ahn = 2.32 / 6.91 = 0.34. The value F(a/rn),taken from Figure A5-1 of WRC Bulletin 175 for an ahn of 0.34, is 1.4. Including the safety factor of 1.5, the stress intensity factor, KI, is 1.5 0 (na)"' . F(a/rn):

Nominal KI = 1.5. 35.04 (n . 2.32)'12. 1.4 = 199 ksi-in2 1

GE Nuclear Energy GE-NE-BI3-02119-QO-81a Non-Proprietary Version 4.3.2.1.4 -

Core Not Critical Neatup/Cooldown Non-Beltline Curve B (Using Feedwafer NozzleAJpper Vessel Region)

The feedwater nozzle was selected to represent non-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.

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

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

Bulletin 175 [I 51.

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 . CT (xa) - 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-NE4313-02119-00-01a Non-Proprietary Version Finite element analysis of a nozzle corner flaw was performed to determine appropriate values of F(a/r,) for Equation 4-4. These values are shown in Figure A5-1 of WRC Bulletin 175 [I 51.

The stresses used in Equation 4-4 were taken from design stress reports for the feedwater nozzle. The stresses considered are primary membrane, cpmr and primary I

bending, opb. Secondary membrane, osm,and secondary bending, stresses are included in the total KI by using ASME Appendix G [6] methods for secondary portion, KIs:

KI, = Mm (Osm + (213) . Osb) (4-5)

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version In the case where the tcJtamtaI stress exceeded yield stress, a plasticity correction factor was applied based on tmi h e recommendations of WRC Bulletin 175 Section 5.C.3 [15].

However, the correctionni r r m was not applied to primary membrane stresses because primary stresses satisfy the l a w : v s s of equilibrium and are not self-limiting. KIPand KI, are added t o obtain the total value oft e M R ' stress intensity factor, KI. A safety factor of 2.0 is applied to primary stresses for coil nonir e not critical heatup/cooldown conditions.

Once KI was calculatedd H , the following relationship was used to determine (T - RTNDT).

The method to solve foimm ~ tr (T - RTNDT) for a specific Kl is based on the KI, equation of Paragraph A-4200 in A:.dZ.:aME Appendix A [17]. The highest RTNDTfor the appropriate non-beltline component.inUt:s was then used to establish the P-T curves.

(T - R T N ~ I ~ w = In

~ ~ ~ -~33.2)

[(K, ~ ) / 20.7341 / 0.02 (4-6)

Exarnplezes=. Core Not Critical HeatuplCooldown Calculation r-f Feedwater NozzlelUpper Vessel Region The non-beltline core nmwna8 at critical heatuplcooldown curve was based on the feedwater nozzle analysis, where feedwater injection of 40°F into the vessel while at operating c o n d t t l llliiitions (551.4"F and 1050 psig) was the limiting normal or upset condition from a brittle t- RFracture perspective. The feedwater nozzle corner stresses were obtained from f i n i t i i t f e e element analysis . To produce conservative thermal stresses, a vessel and I n rirnozzle thickness of 7.5 inches was used in the evaluation.

However, a thickness oc==,.or.-f 7.5 inches is not conservative for the pressure stress evaluation. Therefore, . Y rtthe pressure stress (opm) was adjusted for the actual vessel thickness of 6.18s BE375 inches (i.e., opm= 20.49 ksi was revised to 20.49 ksi .

7.5 inched6.1875 inch--s = 24.84 ksi). These stresses, and other inputs used in the generic calculations, arclre-xe shown below:

opm= 24.84 ksi m-mr,, = 16.19 ksi Y oYs= 45.0 ksi t, = 6.1875 inch Opb = 0.22 ksi OLA t . st? = 19.04 ksi a = 2.36 inch r, = 7.08 inch t, = 7.125 inch In this case the total stri?lrr-ess, 60.29 ksi, exceeds the yield stress, oys,so the correction factor, R, is calculated tB+ t ttlt::o consider the nonlinear effects in the plastic region according to

CE Nuclear Energy CE-NE-B13-02119-00-01a Non-Proprietary Version the following equation based on the assumptions and recommendation of WRC Bulletin 175 [I 51. (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.)

(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:

Gpm = 24.84 ksi osm = 9.44 ksi opb = 0.13 ksi Gsb = 11.10 ksi The value of M m for an inside axial postulated surface flaw from Paragraph G-2214.1 [6]

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

M, = 1.85 for A 5 2 M, = 0.926 4 for 22fi13.464 = 2.845 Mm = 3.21 for &>3.464 The value F(a/rn),taken from Figure A5-1 of WRC Bulletin 175 for an ah, of 0.33, is therefore, F (a / r) = 1.4 KIp is calculated from Equation 4-4:

KIP = 2.0 (24.84 + 0.13) . (X . 2.36) . 1.4 KIP = 190.4 ksi-in KIs is calculated from Equation 4-5:

KI, = 2.845 . (9.44 + 2/3 . 11.10)

GE Nuclear Energy GE-ME-B13-02119-00-01a Non-Proprietary Version K,, = 47.9 ksi-in2 The total KI is, therefore, 238.3 ksi-in*.

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 KI; 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 KI value of 238 ksi-in*, the pressure is 1050 psig and the hot reactor vessel temperature is 551.4F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by

- 40) / (551.4 - 40). From KI the associated (T - RTNDT)can be calculated:

(Tsaturatlon GE Nuclear Energy GE-NE-B13-~2119-00-01 a Non-Proprietary Version Core Not Critical Feedwater Nozzle KI and (T R T N ~ ~ )

as a Function of Pressure

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

The highest non-beltline RTNDT for the feedwater nozzle at Peach Bottom Unit 3 is 40°F as shown in Table 4-2. The generic curve is applied to the Peach Bottom Unit 3 upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDTvalue of 44°F as discussed in Section 4.3.2. I.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-B13-02119-0Q-Q1~

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

4.3.2.2.1 -

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

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

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G IS]for comparison with K,c, 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 KI, equation of Paragraph A-4200 in ASME Appendix A [I71 for the pressure test condition:

KIm . SF = Klc = 20.734 exp[0.02 (T - RTNDT)] + 33.2 (4-9)

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

GE Nuclear Energy GE-bJE-BI3-02119-00-81a Non-Proprietary Version GE's current practice for the pressure test curve is to add a stress intensity factor, Kit, for a coolant heatupkooldown rate of 2Q°F/hrto provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant heatupkooldown rate of 1OO"F/hr. The Klt calculation for a coolant heatupkooldown rate of 1OO"F/hr is described in Section 4.3.2.2.3 below.

4.3.2.2.2 -

Calculations for the Beltline Region Pressure Test This sample calculation is for a pressure test pressure of 1105 psig at 32 EFPY. The following inputs were used in the beltline limit calculation:

Adjusted RTNDT= Initial RTNDT+ Shift A = 10+76=86"F (Based on ART values in Section 4.2)

Vessel Height H = 874.75 inches Bottom of Active Fuel Height B = 216.3 inches Vessel Radius (to inside of clad) R = 125.5 inches Minimum Vessel Thickness (without clad) t = 6.125 inches I

Pressure is calculated to include hydrostatic pressure for a full vessel:

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

= 1105 + (874.75 - 216.3) 0.0361 = 1129 psig Pressure stress:

G = PRW (4-11)

= 1.129 . 125.5 / & I 2 5 = 23.1 ksi The value of M, 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"' = 2.475. The resulting value obtained was:

GE ~ ~ Energy u ~ ~ ~ a ~ GE-NE-B?3-82119-00-01a Non-Proprietary Version M m = 1.85 for &52 M, = 0.926 & for 2 5 4 5 3 . 4 6 4 = 2.29 M m = 3.21 for &>3.464 The stress intensity factor for the pressure stress is Kim = Mm . 0.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'FFlhr instead of 100"F/hr.

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

Using the Kl, equation of Paragraph A4200 in ASME Appendix A [ I 71, Kim = 52.9, and Kit= 2.28 for a 20"F/hr coolant heatupkooldown rate with a vessel thickness, t, that includes cladding:

(T - RTNDT) = ln[(l.5 . KI, + Kit - 33.2) / 20.7341 / 0.02 (4-12)

= ln[(l.5 . 52.9 + 2.28 - 33.2) / 20.7341IO.02

= 42.4"F T can be calculated by adding the adjusted RTNDT:

T = 42.4 + 86 = 128°F for P = 1105 psig 4.3.2.2.3 -

Beltline Region Core Not Critical HeatupKooldown The beltline curves for core not critical heatupkooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section XI Appendix G [6]:

Klc = 2.0 . KI, +KI{ (4- 13) where KI, is primary membrane K due to pressure and Klt is radial thermal gradient K due to heatuplcooldown.

GE Nuclear Energy GE-NE-&13-02119-00-0la Non-Proprietary Version The pressure stress intensity factor K,, is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.

The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient Mt from Figure G-2214-7 of ASME Appendix G [6] by the through-wall temperature gradient AT, 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:

d 2T(x,t) / d x2 = 1 / p (ST(x,t) / 3t) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and p is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that dT(x,t) / 3t = dT(t) / dt = G, where G is the coolant heatuplcooldown rate, normally 1OO"F/hr. The differential equation is integrated over x for the following boundary conditions:

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

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

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

T = Gx'/ 2p - GCx / p + To (4-1 5)

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

GE Nudear Energy GE-NE-B13-Qthl19-00-01a Non-Proprietary Version The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G [6]. Therefore, ATwcalculated 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 [I51 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 Calculations for the Beltline Region Core Not Critical HeatupKooldo wn 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 K,, 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 Klt are:

Coolant heatup/cooldown rate, normally 1OO"F/hr G = 100 "F/hr Minimum vessel thickness, including clad thickness C = 0.526 ft (6.313 inches)

Thermal diffusivity at 550°F (most conservative value) p = 0.354 ft2/ hr [21J 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-313-02119-00-01a Non-Proprietary Version AT = GC2/2P (4-16)

= I 0 0 . (0.526)2/ (2 0.354) = 39°F I

The analyzed case for thermal stress is a 1/4T flaw depth with wall thickness of C. The corresponding value of Mt (=0.2916) can be interpolated from ASME Appendix G, Figure G-2214-2 [6]. Thus the thermal stress intensity factor, Klt = Mt . AT = 11.42, can be calculated. K, 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 . Kl, + Kit) - 33.2) / 20.7341/0.2 (4-17)

= ln[(2 . 52.9 + 11.42 - 33.2) / 20.7341/0.02

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

T=70+86=156"F f o r P = 1105psig 4.3.2.3 CLOSURE FLANGE REGION 10CFR50 Appendix G [8] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT. In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves.

However, some closure flange requirements do impact the curves, as is true with Peach Bottom Unit 3 at low pressures.

The approach used for Peach Bottom Unit 3 for the bolt-up temperature was based on a conservative value of (RTNDT+ 60), or the LST of the bolting materials, whichever is greater. The 60°F adder is included by GE for two reasons: 1) the pre-1971 requirements of the ASME Code Section Ill, Subsection NA, Appendix G included the GE Nuclear Energy GE-ME-Bl3-0~119-00-01a Non-Proprietary Version 60°F adder, and 2) inclusion of the additional 60°F requirement above the RTNDT provides the additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Tables 4-1 and 4-3, the limiting initial RTNDT for the closure flange region is represented by both the top head and vessel shell flange materials at 10°F and the LST of the closure studs is 70°F; therefore, the bolt-up temperature value used is 70°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 RTNDTof the closure region. Curve A temperature must be no less than (RTNDT+ 90°F) and Curve B temperature no less than (RTNDT+ 120°F).

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

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

The shutdown margin, provided in the Peach Bottom Unit 3 Technical Specification, is calculated for a water temperature of 68°F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68°F limit, further extensive calculations would be required to justify a lower temperature. The 70°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.

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version 4.3.2.4 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFRSO, APPENDIX G Curve C, the core critical operation curve, is generated from the requirements of 10CFRSO Appendix G [8], Table 1. Table 1 of [81requires that core critical P-T limits be 40°F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40°F for pressures above 312 psig.

Table 1 of 10CFR50 Appendix G [8] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT+ 60°F) at pressures below 312 psig. This requirement makes the minimum criticality temperature 70"F, based on an RTNDTof 10°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 1140 psig). The requirement of closure region RTN~T + 160°F does cause a temperature shift in Curve C at 312 psig.

GENERAL ELECTRIC (GE) REPORT GE-NE-B13-02119-00-01a, "PRESSURE-TEMPERATURE CURVES FOR EXELON, PEACH BOTTOM UNIT 3" SECTION 6.0

GE Nuclear Energy GE-NE-4313-02119-00-01a Non-Proprietary Version

6.0 REFERENCES

1 T. A. Caine, "Peach Bottom Atomic Power Station, Unit 3 Vessel Surveillance Materials Testing and Fracture Toughness Analysis," GE-NE, San Jose, CA, July 1995, (SASR 90-50, Revision I ) .

2. GE Drawing Number 729E762, "Reactor Vessel Thermal Cycles," GE-APED, San Jose, CA, Revision 0. Peach Bottom Units 2 and 3 RPV Thermal Cycle Diagram (GE Proprietary).

3. GE Drawing Number 135B9990, "Nozzle Thermal Cycles," GE-APED, San Jose, CA, Revision 1. Peach Bottom Units 2 and 3 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 VesselsSection XI, Division 1," Code Case N-588 of the ASME Boiler &

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

6. "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section Ill 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-B13-02119-00-01a Non-Proprietary Version

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

Report for BWR Owners1Group, San Jose, California, September 1994 (GE Proprietary).

11. Letter from B. Sheron to R.A. Pinellili'Safety Assessment of Report NEDC-32399-PI Basis for GE RTNDTEstimation Method, September 1994, " USNRC, December 16, 1994.

12. QA Records & RPV CMTR's and Purchase Specification:

12.1. Peach Bottom Unit 3 - A Nuclear Vessel Fabricated by One Manufacturer for Another Manufacturer initiated by Babcock and Wilcox (B&VV) and completed by BVI Nuclear (CBIN) - (QA Records & RPV CMTR's Peach Bottom Unit 3 GE PO# 205-B1156, Manufactured by B&W and GE PO#205-H4641, Manufactured by CBIN), B&W Contract No. 610-0146-51 and CBI Contract No. 69-5128.

12.2. GE Document #21A1111, Rev. 9, "Purchase Specification for the Reactor Pressure Vessel," GE-APED, San Jose, CA.

13. a). Letter GL 92-01, Rev.1, Supp. 1 From Garrett D. Edwards, Director-Licensing,To U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, "Peach Bottom Atomic Power Station, Units 2 and 3 Response to Request for Additional lnformation Concerning Generic Letter 92-01! Revision 1, Supplement 1, "Reactor Vessel Structural Integrity", Dated November 24, 1998 b). Nuclear Regulatory Commission Letter From Bartholomew C. Buckley, Sr.

Project manager, Section 2, Office of Nuclear Reactor Regulation, to James A.

Hutton, Director-Licensing, PECO Energy Company, "Closure of TAC Nos MA1203 and MA1204 - Response to the Request for Additional lnformation to Generic Letter 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity" for Peach Bottom Atomic Power Station, Units 2 and 3.

14. Wu, Tang, "Limerick /Peach Bottom Neutron Flux Evaluation," GE-NE, San Jose, CA, May 2001, (GE-NE-B11-00842-00-01, Rev. 0).

GE Nuclear Energy GE-ME-513-02119-00-01 a Non-Proprietary Version

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

Welding Research Council Bulletin 175, August 1972.

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

19. Bottom Head and Feedwater Nozzle Dimensions:

19.1. "Lower Hd Bottom Segment Assembly for Peach Bottom Ill Nuclear Reactor Vessel", Chicago Bridge & lron Co., Drawing No. 4, Rev. 11, (GE VPF # 2753-5-9) 19.2. "Exhibit A Fabrication Report Summary Stress Report for Peach Bottom Ill R.P.V. P.0 205-H4641, Contract 69-5128," GE PO# 205-B1156, Manufactured by B8W and GE PO#205-H4641, Manufactured by CBIN, B&W Contract No. 610-0146-51 and CBI Contract No. 69-5128, (GE VPF # 2753-133-I), Sheet A-I for the Bottom Head and A-28 for the Feedwater Nozzle.

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

22. GE VPF # 2752-124-1, "CB&I Heat No's. & Seam Identification for Peach Bottom Ill Nuclear Reactor Vessel," Chicago Bridge & lron Company, (CB&I Contract No 69-5 128 Drawing #R 1, Rev. 6)

GENERAL ELECTRIC (GE) REPORT GE-NE-B13-02119-00-01a, "PRESSURE-TEMPERATURECURVES FOR EXELON, PEACH BOTTOM UNIT 3" TABLE B-1

GE Nuclear Energy GE-NE-B13-02119-00-01 a Non-Proprietary Version TABLE B-1, Peach Bottom Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "F/hr for Curves B & C and 20 "Flhr for Curve A FOR FIGURES 5-1 5-2, 5-3, 5-5, 5-6, AND 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLlNE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 0 68.0 70.0 70.0 68.0 70.0 70.0 10 68.0 70.0 70.0 68.0 70.0 70.0 20 68.0 70.0 70.0 68.0 70.0 70.0 30 68.0 70.0 70.0 68.0 70.0 70.0 40 68.0 70.0 70.0 68.0 70.0 70.0 50 68.0 70.0 70.0 68.0 70.0 70.0 60 68.0 70.0 70.0 68.0 70.0 70.0 70 68.0 70.0 70.0 68.0 70.0 70.0 80 68.0 70.0 70.0 68.0 70.0 70.0 90 68.0 70.0 70.0 68.0 70.0 70.0 100 68.0 70.0 70.0 68.0 70.0 70.0 110 68.0 70.0 70.0 68.0 70.9 70.0 120 68.0 70.0 70.0 68.0 74.7 70.0 130 68.0 70.0 70.0 68 .O 78.2 70.0 140 68.0 70.0 70.0 68.0 81.4 70.0 150 68.0 70.0 70.0 68.0 84.2 70.0 160 68.0 70.0 70.0 68.0 86.9 70.0 170 68.0 70.0 70.0 68.0 89.5 70.0 180 68.0 70.0 70.0 68.0 91.9 70.0 190 68.0 70.0 70.0 68.0 94.2 70.0 200 68.0 70.0 70.0 68.0 96.3 70.0 210 68.0 70.0 70.0 68.0 98.3 70.0 220 68.0 70.0 70.0 68.0 100.3 70.0 230 68.0 70.0 70.0 68.0 102.1 70.0 240 68.0 70.0 70.0 68.0 103.9 70.0 250 68.0 70.0 70.0 68.0 105.6 70.0 260 68.0 70.0 70.0 68.0 107.2 70.0 270 68.0 70.0 70.0 68.0 108.8 70.0 6-2

GE Nuclear Energy GE-NE-513-02119-00-81a Non-Proprietary Version TABLE B-I . Peach Bottom Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "Flhr for Curves B & C and 20 "F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, AND 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 280 68.0 70.0 70.0 68.0 110.3 70.0 290 68.0 70.0 70.0 68.0 111.8 70.0 300 68.0 70.0 70.0 68.0 113.2 70.0 310 68.0 70.0 70.0 68.0 114.5 70.0 312.5 68.0 70.0 70.0 68.0 114.9 70.0 312.5 68.0 100.0 100.0 68.0 130.0 130.0 320 68.0 100.0 100.0 68.0 130.0 130.0 330 68.0 100.0 100.0 68.0 130.0 130.0 340 68.0 100.0 100.0 68.0 130.0 130.0 350 68.0 100.0 100.0 68.0 130.0 130.0 360 68.0 100.0 100.0 68.0 130.0 130.0 370 68.0 100.0 100.0 68.0 130.0 130.0 380 68.0 100.0 100.0 68 .O 130.0 130.0 390 68.0 100.0 100.0 68.0 130.0 130.0 400 68.0 100.0 100.0 68.0 130.0 130.0 410 68.0 100.0 100.0 68.0 130.0 130.0 420 68.0 100.0 100.0 68.0 130.0 130.0 430 68.0 100.0 100.0 68.0 130.0 130.0 440 68.0 100.0 100.0 68.0 130.0 130.0 450 68.0 100.0 100.0 68.0 130.1 130.0 460 68.0 100.0 100.0 68.0 131.1 130.0 470 68.0 100.0 100.0 68.0 132.0 130.0 480 68.0 100.0 100.0 68.0 132.9 130.0 490 68.0 100.0 100.0 68.0 133.7 130.0 500 68.0 100.0 100.0 68.0 134.6 130.0 510 68.0 100.0 100.0 68.0 135.4 130.0 520 68.0 100.0 100.0 68.2 136.2 130.0 530 68.0 100.0 100.0 70.2 137.0 130.0 540 68.0 100.0 100.0 72.1 137.8 130.0 550 68.0 100.0 100.0 73.9 138.6 130.0 560 68.0 100.0 100.0 75.7 139.4 130.0 5-3

GE-NE-Bl3-02119-00-01a Non-Proprietary Version

~ ~~

TABLE B-1. Peach Bottom Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "Flhr for Curves B & C and 20 "F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, AND 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) ("F) (OF) 570 68.0 100.0 100.0 77.4 140.1 130.0 580 68.0 100.0 100.0 79.0 140.9 130.0 590 68.0 100.0 100.0 80.6 141.6 130.0 600 68.0 100.0 100.0 82.2 142.1 130.0 610 68.0 100.0 100.0 83.7 142.6 130.0 620 68.0 100.0 100.0 85.1 143.0 130.0 630 68.0 100.0 100.0 86.5 143.4 130.0 640 68.0 100.0 100.0 87.9 143.8 130.0 650 68.0 100.0 100.0 89.2 144.2 130.0 660 68.0 100.0 100.0 90.5 144.7 130.0 670 68.0 100.9 100.0 91.8 145.1 130.0 680 68.0 101.9 100.0 93.1 145.5 130.0 690 68.0 102.8 100.0 94.3 145.9 130.0 700 69.2 103.7 100.0 95.4 146.3 130.0 710 70.7 104.6 100.0 96.6 146.7 130.0 720 72.1 105.4 100.0 97.7 147.1 130.0 730 73.5 106.3 100.0 98.8 147.5 130.0 740 74.8 107.1 100.0 99.9 147.9 130.0 750 76.1 108.0 100.0 101.0 148.2 130.9 760 77.4 108.8 100.0 102.0 148.6 131.8 770 78.6 109.6 100.0 103.0 149.0 132.7 780 79.8 110.3 100.0 104.0 149.4 133.6 790 81.O 111.1 100.0 105.0 149.8 134.4 800 82.2 111.9 100.0 105.9 150.1 135.3 810 83.3 112.6 100.7 106.9 150.5 136.1 820 84.4 113.4 101.9 107.8 150.9 136.9 830 85.5 114.1 103.1 108.7 151.2 137.8 840 86.5 114.8 104.3 'f09.6 151.6 138.6 850 87.6 115.5 105.5 110.4 151.9 139.3 860 88.6 116.2 106.6 111.3 152.3 140.1 870 89.6 116.9 107.7 112.1 152.6 140.9 B-4

GE Nuclear Energy GE-NE-Bl3-02119-00-01a Non-Proprietary Version TABLE B-1. Peach Bottom Unit 3 P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "Fhr for Curves B & C and 20 "F/hr for Curve A FOR FIGURES 5-1, 5-2,5-3,5-5,5-6, AND 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) ("F) (OF) (OF) 880 90.5 117.6 108.8 113.0 153.0 141.6 890 91.5 118.3 109.9 113.8 153.3 142.4 900 92.4 118.9 110.9 114.6 153.7 143.1 910 93.4 119.6 112.0 115.4 154.0 143.8 920 94.3 120.2 113.0 116.1 154.4 144.5 930 95.1 120.9 113.9 116.9 154.7 145.2 940 96.0 121.5 114.9 117.7 155.0 145.9 950 96.9 122.1 115.9 118.4 155.4 146.6 960 97.7 122.7 116.8 119.1 155.7 147.3 970 98.6 123.3 117.7 119.9 156.0 147.9 980 99.4 123.9 118.6 120.6 156.4 148.6 990 100.2 124.5 119.5 121.3 156.7 149.2 1000 101.o 125.1 120.3 122.0 157.0 149.8 1010 101.7 125.7 121.2 122.6 157.3 150.5 1020 102.5 126.2 122.0 123.3 157.6 151.1 1030 103.3 126.8 122.8 124.0 158.0 151.7 1 040 104.0 127.4 123.6 124.6 158.3 152.3 1050 104.7 127.9 124.4 125.3 158.6 152.9 1060 105.4 128.5 125.2 125.9 158.9 153.5 1070 106.2 129.0 126.0 126.5 159.2 154.1 1080 106.9 129.5 126.7 127.2 159.5 154.7 1090 107.6 130.1 127.5 127.8 159.8 155.2 1100 108.2 130.6 128.2 128.4 160.1 155.8 1105 108.6 130.8 128.6 128.7 160.3 156.1 1110 108.9 131.I 128.9 129.0 160.4 156.3 1120 109.6 131.6 129.7 129.6 160.7 156.9 1130 110.2 132.1 130.4 130.2 161.O 157.4 1140 110.9 132.6 131.1 130.7 161.3 158.0 1150 111.5 133.1 131.7 131.3 161.6 158.5 1160 112.1 133.6 132.4 131.9 161.9 159.0 1170 112.8 134.1 133.1 132.4 162.2 159.6 B-5

GE-NE-813-02219-00-01a Non-Proprietary Version TABLE B-1. Peach Bottom 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, AND 5-7 BOTTOM UPPER 32 EFPY BOTTOM UPPER 32 EFPY HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 1180 113.4 134.6 133.7 133.0 162.5 160.1 1190 114.0 135.1 134.4 133.5 162.7 160.6 1200 114.6 135.5 135.0 134.1 163.0 161.1 1210 115.2 136.0 135.7 134.6 163.3 161.6 1220 115.8 136.5 136.3 135.2 163.6 162.1 1230 116.3 136.9 136.9 135.7 163.9 162.6 1240 116.9 137.4 137.5 136.2 164.2 163.1 1250 117.5 137.8 138.1 136.7 164.4 163.6 1260 118.0 138.3 138.7 137.2 164.7 164.0 1270 118.6 138.7 139.3 137.7 165.0 164.5 1280 119.1 139.2 139.9 138.2 165.2 165.0 1290 119.7 139.6 140.5 138.7 165.5 165.5 1300 120.2 140.0 141.0 139.2 165.8 165.9 1310 120.7 140.5 141.6 139.7 166.1 166.4 1320 121.3 140.9 142.1 140.2 166.3 166.8 1330 121.8 141.3 142.7 140.6 166.6 167.3 1340 122.3 141.7 143.2 141.1 166.8 167.7 1350 122.8 142.1 143.8 141.6 167.1 168.2 1360 123.3 142.6 144.3 142.0 167.4 168.6 1370 123.8 143.0 144.8 142.5 167.6 169.0 1380 124.3 143.4 145.4 142.9 167.9 169.4 1390 124.8 143.8 145.9 143.4 168.1 169.9 1400 125.3 144.2 146.4 143.8 168.4 170.3 B-6

GENERAL ELECTRIC (GE) REPORT GE-NE-B13-02119-00-01a,

'"RESSURE-TEMPERATURE CURVES FOR EXELON, PEACH BOTTOM UNIT 3" TABLE 13-2

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version TABLE 5-2.Peach Bottom Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "Flhr for Curves B & C and 20 "F/hr for Curve A FOR FIGURES 5-10, 5-11, AND 5-12 NON-BELTLINE BOTTOM UPPER RPV & BOTTOM UPPER RPV & AND 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) 0 68.0 70.0 68.0 70.0 70.0 10 68.0 70.0 68.0 70.0 70.0 20 68.0 70.0 68.0 70.0 70.0 30 68.0 70.0 68.0 70.0 70.0 40 68.0 70.0 68.0 70.0 70.0 50 68.0 70.0 68.0 70.0 75.1 60 68.0 70.0 68.0 70.0 84.0 70 68.0 70.0 68.0 70.0 91.2 80 68.0 70.0 68.0 70.0 97.2 90 68.0 70.0 68.0 70.0 102.3 100 68.0 70.0 68.0 70.0 106.8 110 68.0 70.0 68.0 70.9 110.9 120 68.0 70.0 68.0 74.7 114.7 130 68.0 70.0 68.0 78.2 118.2 140 68.0 70.0 68.0 81.4 121.4 150 68.0 70.0 68.0 84.2 124.2 160 68.0 70.0 68.0 86.9 126.9 170 68.0 70.0 68.0 89.5 129.5 180 68.0 70.0 68.0 91.9 131.9 190 68.0 70.0 68.0 94.2 134.2 200 68.0 70.0 68.0 96.3 136.3 210 68.0 70.0 68.0 98.3 138.3 220 68.0 70.0 68.0 100.3 140.3 230 68.0 70.0 68.0 102.1 142.1 240 68.0 70.0 68.0 103.9 143.9 250 68.0 70.0 68.0 105.6 145.6 260 68.0 70.0 68.0 107.2 147.2 270 68.0 70.0 68.0 108.8 148.8 8-7

GE Nuclear Energy GE-blE-B13-02?19-00-01a Non-Proprietary Version TABLE B-2. Peach Bottom Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "Flhr for Curves B & C and 20 "Flhr for Curve A FOR FIGURES 5-10, 5-11, AND 5-12 NON-BELTLINE BOTTOM UPPER RPV & BOTTOM UPPER RPV & AND 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) 280 68.0 70.0 68.0 110.3 150.3 290 68.0 70.0 68.0 111.8 151.8 300 68.0 70.0 68.0 113.2 153.2 310 68.0 70.0 68.0 114.5 154.5 312.5 68.0 70.0 68.0 114.9 154.9 312.5 68.0 100.0 68.0 130.0 170.0 320 68.0 100.0 68.0 130.0 170.0 330 68.0 100.0 68.0 130.0 170.0 340 68.0 100.0 68.0 130.0 170.0 350 68.0 100.0 68.0 130.0 170.0 360 68.0 100.0 68.0 130.0 170.0 370 68.0 100.0 68.0 130.0 170.0 380 68.0 100.0 68.0 130.0 170.0 390 68.0 100.0 68.0 130.0 170.0 400 68.0 100.0 68.0 130.0 170.0 410 68.0 100.0 68.0 130.0 170.0 420 68.0 100.0 68.0 130.0 170.0 430 68.0 100.0 68.0 130.0 170.0 440 68.0 100.0 68.0 130.0 170.0 450 68.0 100.0 68.0 130.1 170.1 460 68.0 100.0 68.0 131.1 171.1 470 68.0 100.0 68.0 132.0 172.0 480 68.0 100.0 68.0 132.9 172.9 490 68.0 100.0 68.0 133.7 173.7 500 68.0 100.0 68.0 134.6 174.6 510 68.0 100.0 68.0 135.4 175.4 520 68.0 100.0 68.2 136.2 176.2 530 68.0 100.0 70.2 137.0 177.0 540 68.0 100.0 72.1 137.8 177.8 550 68.0 100.0 73.9 138.6 178.6 B-8

GE Nuclear Energy GE-NE-Bl3-02Il9-0O-OZa Non-Proprietary Version TABLE 8-2. Peach Bottom Unit 3 Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 "Flhr for Curves B & C and 20 "F/hr for Curve A FOR FIGURES 5-10, 5-11, AND 5-12 NON-BELTLINE BOnOM UPPER RPV & BOTTOM UPPER RPV & AND 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) 560 68.0 100.0 75.7 139.4 179.4 570 68.0 100.0 77.4 140.1 180.1 580 68.0 100.0 79.0 140.9 180.9 590 68.0 100.0 80.6 141.6 181.6 600 68.0 100.0 82.2 142.1 182.1 610 68.0 100.0 83.7 142.6 182.6 620 68.0 100.0 85.1 143.0 183.0 630 68.0 100.0 86.5 143.4 183.4 640 68.0 100.0 87.9 143.8 183.8 650 68.0 100.0 89.2 144.2 184.2 660 68.0 100.0 90.5 144.7 184.7 670 68.0 100.9 91.8 145.1 185.1 680 68.0 101.9 93.1 145.5 185.5 690 68.0 102.8 94.3 145.9 185.9 700 69.2 103.7 95.4 146.3 186.3 710 70.7 104.6 96.6 146.7 186.7 720 72.1 105.4 97.7 147.1 187.1 730 73.5 106.3 98.8 147.5 187.5 740 74.8 107.1 99.9 147.9 187.9 750 76.1 108.0 101.o 148.2 188.2 760 77.4 108.8 102.0 148.6 188.6 770 78.6 109.6 103.0 149.0 189.0 780 79.8 110.3 104.0 149.4 189.4 790 81.O 111.1 105.0 149.8 189.8 800 82.2 111.9 105.9 150.1 190.1 810 83.3 112.6 106.9 150.5 190.5 820 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 B-9

GE-NE-B13-Q2119-00-01a Non-Proprietary Version TABLE 8-2. Peach Bottom 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 5-10, 5-11, AND 5-12 NON-BELTLINE BOTTOM UPPER RPV & BOTTOM UPPER RPV & AND HEAD BELTLINE AT HEAD BELTLINE AT BELTLINE AT 32 EFPY 32 EFPY 32 EFPY PRESSURE CURVEA CURVE A CURVE B CURVE B CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) 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 96.9 122.1 118.4 155.4 195.4 960 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.o 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.o 1140 I 10.9 132.6 130.7 161.3 201.3 5-10

GE Nuclear Energy GE-NE-B13-02119-00-01a Non-Proprietary Version TABLE 8-2. Peach Bottom 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 5-10, 5-11, AND 5-12 NON-BELTLINE BOTTOM UPPER RPV & BOTTOM UPPER RPV & AND 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) ("F) (OF) (OF) (OF) 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 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 116.3 136.9 135.7 163.9 203.9 1240 116.9 137.5 136.2 164.2 204.2 1250 117.5 138.1 136.7 164.4 204.4 1260 118.0 138.7 137.2 164.7 204.7 1270 118.6 139.3 137.7 165.0 205.0 1280 119.1 139.9 138.2 165.2 205.2 1290 119.7 140.5 138.7 165.5 205.5 1300 120.2 141.O 139.2 165.9 205.9 1310 120.7 141.6 139.7 166.4 206.4 1320 121.3 142.1 140.2 166.8 206.8 1330 121.8 142.7 140.6 167.3 207.3 1340 122.3 143.2 141.I 167.7 207.7 1350 122.8 143.8 141.6 168.2 208.2 1360 123.3 144.3 142.0 168.6 208.6 1370 123.8 144.8 142.5 169.0 209.0 1380 124.3 145.4 142.9 169.4 209.4 1390 124.8 145.9 143.4 169.9 209.9 1400 125.3 146.4 143.8 170.3 210.3 B-I 1