RS-11-169, Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate

Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate

ML113050427
Person / Time
Site:	Byron, Braidwood
Issue date:	11/01/2011
From:	Borton K F Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML113050426	List: RS-11-169, Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
References
RS-11-169, TAC ME6587, TAC ME6588, TAC ME6589, TAC ME6590
Download: ML113050427 (43)

Text

ATTACHMENT 2 Cameron International Corporation Affidavit Supporting Withholding

ATTACHMENT 3 Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

(NON-PROPRIETARY VERSION)

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 1 NON-PROPRIETARY NRC Request 1 Section IV.1.C.vi of Attachment 5 to the June 23, 2011, submittal stated, "the current capsule withdrawal schedule shown in each Unit's PTLR [Pressure Temperature Limits Report] will be updated to reflect the latest capsule fluence, lead factor, and withdrawal EFPY [effective full power years] associated with each capsule." The updated capsule withdrawal schedules for Byron, Units 1 and 2, can be found in Tables IV.1.C.vi-1 and IV.1.C.vi-2 and for Braidwood, Units 1 and 2, in Tables IV.1.C.vi-3 and IV.1.C.vi-4 of Attachment 5 to the submittal. Although Note (b) of these Tables stated that the information was updated as part of the measurement uncertainty recapture power uprate (MUR PU), the source or reference for the latest capsule fluence, lead factor, and withdrawal EFPY associated with e ach capsule is not given. Provide the references for the updated fluence, lead factor, and withdrawal EFPY for the Byron and Braidwood units surveillance capsules. If no such references exist, provide calculation details for the updated values in this application.

Response Section IV.1.C.ii of Attachment 5 to the June 23, 2011 Byron and Braidwood Stations Measurement Uncertainty Recapture Power Uprate (MUR-PU) License Amendment Request (LAR) submittal (Reference 1) describes the methodology used for determining vessel fluence.

The discrete ordinates models used for determining vessel fluence were also used to determine surveillance capsule fluence at the time of each capsules withdrawal. Lead factors for each capsule were then calculated, where lead factor is defined as the ratio of the calculated fast (E > 1.0 MeV) fluence at the geometric center of the surveillance capsule to the corresponding maximum calculated fast fluence at the pressure vessel clad/base metal interface. The following Tables 1 and 2 provide the lead factor calculations for Byron Units 1 and 2, respectively. Tables 3 and 4 provide the lead factor calculations for Braidwood Units 1 and 2, respectively. In Tables 1 through 4, Irradiation Time indicates the time in effective full-power years (EFPY) from plant startup to the time at which each capsule was removed; these are the same values as indentified as Withdrawal EFPY in Tables IV.1.C.vi-1 through 4 in the MUR-PU LAR submittal (Reference1). Fast fluence at the capsule position and maximum fluence at the pressure vessel clad/base metal interface are also provided, along with the resultant lead

factor. The vessel and surveillance capsule fluence values contained in this letter and our original submittal (Reference 1) were calculated as part of the MUR-PU project and are not contained in any prior surveillance capsule reports. Fluence calculations were based on the NRC-approved methodologies described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, and WCAP-16083-NP-A, Revision 0, Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry, (References IV.1.C.ii-1 and IV.1.C.ii -2 in Reference 1). These methodologies meet the requirements of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, U. S.

Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (March 2001). The details of these calculations are contained in Westinghouse Calculation Notes (References 2 through 4) which can be made available for audit upon request. Prior surveillance capsule submittals to the NRC used a different methodology, one that was based on adjoint calculations and not forward transport calculations.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 2 NON-PROPRIETARY Table 1 Byron Unit 1 Surveillance Capsule Lead Factors Fast (E > 1.0 MeV) Fluence IrradiationAt Capsule Maximum Lead Capsule Time at Vessel Factor (EFPY) (n/cm

2) (n/cm 2) U 1.18 4.09 x 10 18 1.01 x 10 18 4.05 X 5.67 1.49 x 10 19 3.64 x 10 18 4.09 W 9.27 2.26 x 10 19 5.54 x 10 18 4.08 Z 14.59 3.34 x 10 19 8.12 x 10 18 4.11 V 14.59 3.16 x 10 19 8.12 x 10 18 3.89 Y 18.81 3.97 x 10 19 1.03 x 10 19 3.85 Table 2 Byron Unit 2 Surveillance Capsule Lead Factors Fast (E > 1.0 MeV) Fluence IrradiationAt Capsule Maximum Lead Capsule Time at Vessel Factor (EFPY) (n/cm

2) (n/cm 2) U 1.19 4.06 x 10 18 1.01 x 10 18 4.02 W 4.67 1.20 x 10 19 2.95 x 10 18 4.07 X 8.63 2.18 x 10 19 5.27 x 10 18 4.14 Z 14.28 3.25 x 10 19 7.91 x 10 18 4.11 V 14.28 3.07 x 10 19 7.91 x 10 18 3.88 Y 20.05 4.19 x 10 19 1.08 x 10 19 3.88 Table 3 Braidwood Unit 1 Surveillance Capsule Lead Factors Fast (E > 1.0 MeV) Fluence IrradiationAt Capsule Maximum Lead Capsule Time at Vessel Factor (EFPY) (n/cm

2) (n/cm 2) U 1.16 3.88 x 10 18 9.65 x 10 17 4.02 X 4.30 1.17 x 10 19 2.88 x 10 18 4.06 W 7.79 1.98 x 10 19 4.89 x 10 18 4.05 Z 12.01 2.79 x 10 19 6.82 x 10 18 4.09 Y 12.01 2.60 x 10 19 6.82 x 10 18 3.81 V 17.69 3.71 x 10 19 9.46 x 10 18 3.92 Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 3 NON-PROPRIETARY Table 4 Braidwood Unit 2 Surveillance Capsule Lead Factors Fast (E > 1.0 MeV) Fluence IrradiationAt Capsule Maximum Lead Capsule Time at Vessel Factor (EFPY) (n/cm

2) (n/cm 2) U 1.18 3.88 x 10 18 9.51 x 10 17 4.08 X 4.24 1.15 x 10 19 2.85 x 10 18 4.03 W 8.56 2.07 x 10 19 5.10 x 10 18 4.06 Z 12.78 2.83 x 10 19 6.83 x 10 18 4.14 Y 12.78 2.66 x 10 19 6.83 x 10 18 3.89 V 18.42 3.73 x 10 19 9.52 x 10 18 3.92 Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 4 NON-PROPRIETARY NRC Request 2 For Byron, Unit 2,Section IV.1.C.iii of Attachment 5 to the June 23, 2011, submittal stated, "For Unit 2, the limiting ART [adjusted reference temperature] values used in the development of the current P-T limit curves at 32 EFPY are slightly lower than the MUR [PU] limiting ART values (at 32 EFPY)." The NRC staff cannot verify this because this statement seems to contradict the information in Table IV.1.C.ii-1 of Attachment 5 where the maximum neutron fluence value of

2.06 E19 n/cm 2 (E > 1.0 MeV) on record (i.e., the 2006 PTLR for 32 EFPY) bounds the MUR PU maximum neutron fluence value of 1.76 E 19 n/cm 2 (E > 1.0 MeV). Provide details regarding the calculation of the Byron, Unit Nos. 1 and 2, Reactor Pressure Vessel beltline material ARTs which demonstrate how the values in this submittal are different from the corresponding values in the 2006 Byron, Unit Nos. 1 and 2, PTLRs. These ARTs will be considered as the licensing basis in support of the MUR PU license amendment request.

Response The ART calculations were performed for the Byron Units 1 and 2 reactor vessel beltline materials using the methodology contained in Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel materials, Revision 2, and the MUR-PU fluence values corresponding to 32 EFPY (EOL). The following Tables 5 and 6, provide the 1/4 thickness (1/4T) location and 3/4

thickness (3/4T) location ART calculation information for Byron Unit 1, respectively. The following Tables 7 and 8 provide the Byron Unit 2 1/4T and 3/4T ART calculation information, respectively. The limiting ART values used in the development of the current P-T limit curves along with those calculated using the MUR-PU fluence values are summarized in Tables 9 and 10 for Byron Units 1 and 2, respectively. As stated in the RAI, the maximum neutron fluence value on record (2006 PTLR) of 2.06E19 n/cm 2 bounds the MUR-PU maximum neutron fluence value of 1.76E19 n/cm 2 at 32 EFPY for Byron Unit 2. These values represent the maximum neutron fluence exposure over the entire reactor vessel. The ART values were calculated using neutron fluence values specific to each of the reactor vessel beltline materials. Even though the MUR-PU maximum neutron fluence is lower than the maximum neutron fluence on record at 32 EFPY, the MUR-PU neutron fluence value calculated specifically for the Byron Unit 2 nozzle shell forging at 32 EFPY is greater than the neutron fluence value used in

the development of the current P-T limit curves (2006 PTLR) for the nozzle shell forging at 32 EFPY. The MUR-PU neutron fluence value for the Byron Unit 2 nozzle shell forging was determined to be 5.49E18 n/cm 2 , whereas, the neutron fluence value used for this material in the P-T limits analysis of record was 5.22E18 n/cm

2. This neutron fluence increase for the Unit 2 nozzle shell forging material resulted in higher ART values at 32 EFPY for the MUR-PU as compared to those used in the development of the current P-T limit curves (2006 PTLR). The comparison of limiting ART values between the analysis of record and the MUR-PU is summarized in Table 10 below for Byron Unit 2, and results in a slight increase in limiting ART values and a decrease of P-T limits applicability from 32 EFPY to 30.5 EFPY as previously noted in Section IV.1.C.iii of Attachment 5 of the MUR-PU submittal (Reference 1) and is

therefore not contradictory.

The appropriate licensing documents (PTLR) will be updated to reflect the MUR-PU ART values along with the new applicability date of 30.5 EFPY for both the Byron Unit 2 heatup and cooldown limit curves.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 5 NON-PROPRIETARY Table 5 Calculation of the Byron Unit 1 ART Values at the 1/4T Location for 32 EFPY Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF (F) 1/4T Fluence (n/cm 2 , E > 1.0 MeV) 1/4T FF IRT NDT (a) (F) RT NDT (F) I (°F) (°F) Margin (F) ART (F) Nozzle Shell Forging 1.1 31 0.359 x 10 19 0.7173 30 22.2 0 11.1 22.2 74 Intermediate Shell Forging 1.1 26 1.063 x 10 19 1.0171 40 26.4 0 13.2 26.4 93 Using non-credible surveillance data 2.1 30.6 1.063 x 10 19 1.0171 40 31.1 0 15.6 31.1 102 Lower Shell Forging 1.1 26 1.063 x 10 19 1.0171 10 26.4 0 13.2 26.4 63 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011) 1.1 41 0.359 x 10 19 0.7173 10 29.4 0 14.7 29.4 69 Using credible Braidwood Units 1 and 2 surveillance data 2.1 26.1 0.359 x 10 19 0.7173 10 18.7 0 9.4 18.7 47 Intermediate to Lower Shell Forging Circ. Weld Seam (Heat # 442002) 1.1 54 1.033 x 10 19 1.0090 -30 54.5 0 27.2 54.5 79 Using credible surveillance data 2.1 66.5 1.033 x 10 19 1.0090 -30 67.1 0 14 28 65 Note: (b) Initial RT NDT values are measured.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 6 NON-PROPRIETARY Table 6 Calculation of the Byron Unit 1 ART Values at the 3/4T Location for 32 EFPY Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF (F) 3/4T Fluence (n/cm 2 , E > 1.0 MeV) 3/4T FF IRT NDT (a) (F) RT NDT (F) I (°F) (°F) Margin (F) ART (F) Nozzle Shell Forging 1.1 31 0.129 x 10 19 0.4706 30 14.6 0 7.3 14.6 59 Intermediate Shell Forging 1.1 26 0.383 x 10 19 0.7346 40 19.1 0 9.5 19.1 78 Using non-credible surveillance data 2.1 30.6 0.383 x 10 19 0.7346 40 22.5 0 11.2 22.5 85 Lower Shell Forging 1.1 26 0.383 x 10 19 0.7346 10 19.1 0 9.5 19.1 48 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011) 1.1 41 0.129 x 10 19 0.4706 10 19.3 0 9.6 19.3 49 Using credible Braidwood Units 1 and 2 surveillance data 2.1 26.1 0.129 x 10 19 0.4706 10 12.3 0 6.1 12.3 35 Intermediate to Lower Shell Forging Circ. Weld Seam (Heat # 442002) 1.1 54 0.372 x 10 19 0.7269 -30 39.3 0 19.6 39.3 49 Using credible surveillance data 2.1 66.5 0.372 x 10 19 0.7269 -30 48.3 0 14 28 46 Note: (b) Initial RT NDT values are measured.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 7 NON-PROPRIETARY Table 7 Calculation of the Byron Unit 2 ART Values at the 1/4T Location for 32 EFPY Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF (F) 1/4T Fluence (n/cm 2 , E > 1.0 MeV) 1/4T FF IRT NDT (a) (F) RT NDT (F) I (°F) (°F) Margin (F) ART (F) Nozzle Shell Forging 1.1 31 0.330 x 10 19 0.6948 10 21.5 0 10.8 21.5 53 Intermediate Shell Forging 1.1 20 1.057 x 10 19 1.0155 -20 20.3 0 10.2 20.3 21 Lower Shell Forging 1.1 37 1.057 x 10 19 1.0155 -20 37.6 0 17 34 52 Using credible surveillance data 2.1 18.9 1.057 x 10 19 1.0155 -20 19.2 0 8.5 17 16 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011) 1.1 41 0.330 x 10 19 0.6948 40 28.5 0 14.2 28.5 97 Using credible Braidwood Units 1 and 2 surveillance data 2.1 26.1 0.330 x 10 19 0.6948 40 18.1 0 9.1 18.1 76 Intermediate to Lower Shell Forging Circ. Weld Seam (Heat # 442002) 1.1 54 1.021 x 10 19 1.0058 10 54.3 0 27.2 54.3 119 Using credible surveillance data 2.1 66.5 1.021 x 10 19 1.0058 10 66.9 0 14 28 105 Note: (b) Initial RT NDT values are measured.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 8 NON-PROPRIETARY Table 8 Calculation of the Byron Unit 2 ART Values at the 3/4T Location for 32 EFPY Reactor Vessel Material R.G. 1.99, Rev. 2 Position CF (F) 3/4T Fluence (n/cm 2 , E > 1.0 MeV) 3/4T FF IRT NDT (a) (F) RT NDT (F) I (°F) (°F) Margin (F) ART (F) Nozzle Shell Forging 1.1 31 0.119 x 10 19 0.4524 10 14.0 0 7.0 14.0 38 Intermediate Shell Forging 1.1 20 0.381 x 10 19 0.7331 -20 14.7 0 7.3 14.7 9 Lower Shell Forging 1.1 37 0.381 x 10 19 0.7331 -20 27.1 0 13.6 27.1 34 Using credible surveillance data 2.1 18.9 0.381 x 10 19 0.7331 -20 13.9 0 6.9 13.9 8 Nozzle to Intermediate Shell Forging Circ. Weld Seam (Heat # 442011) 1.1 41 0.119 x 10 19 0.4524 40 18.5 0 9.3 18.5 77 Using credible Braidwood Units 1 and 2 surveillance data 2.1 26.1 0.119 x 10 19 0.4524 40 11.8 0 5.9 11.8 64 Intermediate to Lower Shell Forging Circ. Weld Seam (Heat # 442002) 1.1 54 0.368 x 10 19 0.7238 10 39.1 0 19.5 39.1 88 Using credible surveillance data 2.1 66.5 0.368 x 10 19 0.7238 10 48.1 0 14 28 86 Note: (b) Initial RT NDT values are measured.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 9 NON-PROPRIETARY Table 9 Comparison of the Byron Unit 1 Limiting ART Values at EOL (32 EFPY) 1/4T Limiting ART (°F) 3/4T Limiting ART (°F)

Limiting Material Existing Curves in 2006 PTLR MUR-PU Existing Curves in 2006 PTLR MUR-PU Limiting Circ. Flaw Material (Intermediate to Lower Shell Forging Circ. Weld using Credible Surveillance Data) 67 65 48 46 Limiting Axial Flaw Material (Intermediate Shell Forging using Non-credible Surveillance Data) 106 102 97 85 Table 10 Comparison of the Byron Unit 2 Limiting ART Values at EOL (32 EFPY) 1/4T Limiting ART (°F) 3/4T Limiting ART (°F)

Limiting Material Existing Curves in 2006 PTLR MUR-PU Existing Curves in 2006 PTLR MUR-PU Limiting Circ. Flaw Material (Intermediate to Lower Shell Forging Circ. Weld using Credible Surveillance Data) 107 105 89 86 Limiting Axial Flaw Material (Nozzle Shell Forging) 52 53 37 38

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 10 NON-PROPRIETARY NRC Request 3 For the upper-shelf energy (USE) evaluation,Section IV.1.C.v of Attachment 5 to the June 23,2011, submittal stated that the limiting projected 1/4 T USE value is 65 ft-Ibs for the nozzle-to-intermediate shell forging circumferential weld for Byron, Unit 1, 68 ft-Ibs for the nozzle-to-intermediate shell forging circumferential weld for Byron, Unit 2, 75 ft-Ibs for the intermediate-to-Iower shell forging circumferential weld for Braidwood, Unit 1, and 66 ft-lbs for the intermediate-to-lower shell forging circumferential weld for Braidwood, Unit 2. However, the details regarding the calculation of these limiting USE values or appropriate references are not given in Attachment 5. The 2006 Byron and Braidwood PTLRs, Revision 4, contain no current USE estimates either. Please provide details regarding the calculation of the limiting USE values for the Byron and Braidwood units or references containing this information.

Response The projected end of life (EOL) Charpy USE decreases due to the MUR-PU uprated fluence at the 1/4T location were calculated in accordance with Regulatory Guide 1.99, Revision 2 trend curves. When surveillance data was used, the decrease in USE was obtained by plotting the reduced plant surveillance data on Figure 2 of the Regulatory Guide and drawing a line parallel to the existing lines as the upper bound of all the data.

The following Tables 11 and 12 contain the USE calculation information for the Byron Units 1 and 2 reactor vessel beltline materials, respectively. Tables 13 and 14 contain the USE calculation information for the Braidwood Units 1 and 2 reactor vessel beltline materials, respectively. Note that the copper weight percents for the Byron and Braidwood beltline materials were all less than the lowest copper weight percent chemistry line (0.05% for weld and 0.10% for base metal) delineated in Figure 2 of Regulatory Guide 1.99, Revision 2. Therefore, the Regulatory Guide 1.99, Revision 2, Position 1.2 projected USE decreases in Tables 11 through 14 were conservatively determined using the lowest copper weight percent chemistry line (0.05% for weld and 0.10% for base metal) delineated in Figure 2 of Regulatory Guide 1.99, Revision 2.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 11 NON-PROPRIETARY Table 11 USE Projections for Byron Unit 1 Beltline Region Materials at EOL (32 EFPY)

Material R.G. 1.99, Rev. 2 Position Wt. % Cu 1/4T EOL Fluence (n/cm 2 , E > 1.0 MeV)

Unirradiated USE (ft-lb) Projected USE Decrease (%)

Projected EOL USE (ft-lb) Nozzle Shell Forging 1.2 0.05 0.359 x 10 19 138 15 117 Intermediate Shell Forging 1.2 0.04 1.063 x 10 19 139 20 111 Using Surveillance Capsule Data 2.2 0.04 1.063 x 10 19 139 4.2 133 Lower Shell Forging 1.2 0.04 1.063 x 10 19 150 20 120 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.2 0.03 0.359 x 10 19 77 15 65 Intermediate to Lower Shell Forging Circ. Weld Seam 1.2 0.04 1.033 x 10 19 77 20 62 Using Surveillance Capsule Data 2.2 0.04 1.033 x 10 19 77 11 69

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 12 NON-PROPRIETARY Table 12 USE Projections for Byron Unit 2 Beltline Region Materials at EOL (32 EFPY)

Material R.G. 1.99, Rev. 2 Position Wt. % Cu 1/4T EOL Fluence (n/cm 2 , E > 1.0 MeV)

Unirradiated USE (ft-lb) Projected USE Decrease (%)

Projected EOL USE (ft-lb) Nozzle Shell Forging 1.2 0.05 0.330 x 10 19 155 15 132 Intermediate Shell Forging 1.2 0.01 1.057 x 10 19 149 20 119 Lower Shell Forging 1.2 0.06 1.057 x 10 19 127 20 102 Using Surveillance Capsule Data 2.2 0.06 1.057 x 10 19 127 16 107 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.2 0.03 0.330 x 10 19 80 15 68 Intermediate to Lower Shell Forging Circ. Weld Seam 1.2 0.04 1.021 x 10 19 80 20 64 Using Surveillance Capsule Data 2.2 0.04 1.021 x 10 19 80 2 78 Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 13 NON-PROPRIETARY Table 13 USE Projections for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPY)

Material R.G. 1.99, Rev. 2 Position Wt. % Cu 1/4T EOL Fluence (n/cm 2 , E > 1.0 MeV)

Unirradiated USE (ft-lb) Projected USE Decrease (%)

Projected EOL USE (ft-lb) Nozzle Shell Forging 1.2 0.04 0.352 x 10 19 155 15 132 Intermediate Shell Forging 1.2 0.05 1.057 x 10 19 122 20 98 Lower Shell Forging 1.2 0.05 1.057 x 10 19 135 20 108 Using Surveillance Capsule Data 2.2 0.05 1.057 x 10 19 135 13 117 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.2 0.04 0.352 x 10 19 90 15 77 Intermediate to Lower Shell Forging Circ. Weld Seam 1.2 0.03 1.021 x 10 19 80 20 64 Using Surveillance Capsule Data 2.2 0.03 1.021 x 10 19 80 6.2 75 Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 14 NON-PROPRIETARY Table 14 USE Projections for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPY)

Material R.G. 1.99, Rev. 2 Position Wt. % Cu 1/4T EOL Fluence (n/cm 2 , E > 1.0 MeV)

Unirradiated USE (ft-lb) Projected USE Decrease (%)

Projected EOL USE (ft-lb) Nozzle Shell Forging 1.2 0.04 0.336 x 10 19 115 15 98 Intermediate Shell Forging 1.2 0.03 1.039 x 10 19 119 20 95 Lower Shell Forging 1.2 0.06 1.039 x 10 19 144 20 115 Using Surveillance Capsule Data 2.2 0.06 1.039 x 10 19 144 13 125 Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.2 0.04 0.336 x 10 19 90 15 77 Intermediate to Lower Shell Forging Circ. Weld Seam 1.2 0.03 1.003 x 10 19 80 20 64 Using Surveillance Capsule Data 2.2 0.03 1.003 x 10 19 80 17 66 Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 15 NON-PROPRIETARY NRC Request 4 Section IV.1.A.ii of Attachment 5 to the EGC's June 23, 2011, submittal provides generic information for only a few reactor vessel (RV) internals under the MUR PU conditions. Table Matrix-1 of NRC RS-001, Revision 0, "Review Standard for Extended Power Uprates," provides the staff's basis for evaluating the potential for extended PU to induce aging effects on RV internals. Depending on the magnitude of the projected RV internals fluence, Table Matrix-1

may be applicable to the MUR application. In the Notes to Table Matrix-1, the NRC staff states that guidance on the neutron irradiation-related threshold for irradiation-assisted stress corrosion cracking (SCC) for pressurized water reactor RV internal components are given in BAW-2248A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," and WCAP-14577, Revision 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals." The "Notes" to Table Matrix-1 state that for thermal and neutron embrittlement of cast austenitic stainless steel, SCC, and void swelling, licensees will need to provide plant-specific degradation management programs or participate in industry programs to investigate degradation effects and determine appropriate management programs. The BAW-2248A report and the WCAP-14577, Revision 1-A, have been superseded by the MRP-227 report, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," which summarized the industry's most current recommended inspection and evaluation guidelines for RV internals. The safety evaluation dated June 22, 2011, lists the limitations and conditions imposed by the NRC staff on use of the MRP-227 report. Please confirm whether you have established an inspection plan to manage the age-related degradation in the Byron and Braidwood units RV internals, or whether you have participated in the industry's initiatives on age-related degradation of PWR RV internals and plan to submit your plant-specific program consistent with the MRP-227 report guidelines. For the former case, discuss your management of the above-mentioned aging effects on RV internals and demonstrate that the management is appropriate to ensure integrity and operability of RV internals to the end of license.

Response The Exelon PWR Reactor Internals Management Program (ER-AP-333, "Pressurized Water Reactor Internals Management Program" and ER-AP-333-1001, "Pressurized Water Reactor (PWR) Internals Program") provides the necessary oversight and management to ensure the integrity and operability of reactor vessel internals consistent with the MRP-227 report requirements. Exelon is an active participant in the Materials Reliability Program (MRP) efforts relative to the development of MRP-227 for PWR reactor int ernals inspections. Exelon will prepare the necessary reactor internals aging management plan as part of the overall license renewal process. This aging management plan will be developed in accordance with NUREG-1801, "Generic Aging Lessons Learned (GALL) Report, Revision 2. NUREG-1801, Revision 2 page XI.M16A-1 identifies the recommended content for the PWR Reactor Internals Aging Management program and references MRP-227. This will ensure the Byron and Braidwood reactor internals aging management programs are developed in accordance with the NRC-approved version of MRP-227 (i.e., MRP-227-A). In accordance with NUREG-1801, Revision 2, an inspection plan will be submitted with the License Renewal Application. This is consistent with Category D plants as discussed in NRC Regulatory Issue Summary (RIS) 2011-07, dated July 21, 2011.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 16 NON-PROPRIETARY NRC Request 5 The NRC staff notes that the license amendment request (LAR), Attachment 5, MUR Technical Evaluation,Section VII.6.A, "Fire Protection Program," on page VII-5 states that, " ...an analysis of the change in combustible loading determined that the overall increase in fire loading is small and does not change the fire load classification of each affected fire zone ..." It is unclear to the NRC staff whether there are fire protection program plant modifications planned (e.g., adding new cable trays, or re-routing of existing cables) at MUR power uprate conditions. Clarify whether this request involves plant modifications, or changes to the fire protection program. If any, the staff requests the licensee to identify proposed modifications and discuss the impact of these modifications on the plant's compliance with the fire protection program licensing basis, Title 10 of the Code of Federal Regulations (10 CFR), Section 50.48, or applicable portions of Branch Technical Position, CMEB 9.5-1.

Response The MUR-PU LAR (Reference 1) does not involve plant modifications to the fire protection system, or changes to the fire protection program. The Leading Edge Flow Meter (LEFM) modification adds small amounts of combustible material in the form of instrumentation, control and power cable insulation to twelve (12) Braidwood Unit 2 fire zones, seven (7) Byron Unit 1 fire zones, and seven (7) Byron Unit 2 fire zones. The Braidwood Unit 1 LEFM modification is still under development. The Fire Protection Report (FPR), Section 2.3, currently lists cable insulation as a fire hazard in the affected fire zones and, therefore, the LEFM installation does not introduce a new fire hazard in these fire zones. The change in combustible loading is well within the available combustible loading margins. The Braidwood Unit 2 fire zone with the worst case change in fire loading (BTU/ft

2) is Fire Zone 3.2-0 (Auxiliary Building Elevation 439). The change in fire loading for this zone is 48.2 BTU/ft 2 , which amounts to a 0.35% increase in fire load. The Byron Unit 1 fire zone with the worst case change in fire loading (BTU/ft

2) is Fire Zone 3.2D-1 (Lower Cable Spreading Area). The change in fire loading for this zone is 26.5 BTU/ft 2 , which amounts to a 0.04% increase in fire load. The Byron Unit 2 fire zone with the worst case change in fire loading (BTU/ft

2) is Fire Zone 3.2D-2 (Lower Cable Spreading Area). The change in fire loading for this zone is 840.5 BTU/ft 2 , which amounts to a 0.7% increase in fire load.

The changes in fire loading do not result in a change in classification for any of the affected fire zones. The change in fire loading resulting from the LEFM modification to be installed in the Spring outage for Braidwood Unit 1 will be similar to the changes identified above for Braidwood Unit 2 and Byron Units 1 and 2. The MUR-PU LAR requests NRC approval for the modification to increase the capacity of the SG Power Operated Relief Valves (PORVs) for Byron Unit 1 and Braidwood Unit 1 only. This

modification does not add any additional combustible material and has no impact on the fire

protection system or fire protection program.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 17 NON-PROPRIETARY The other modifications associated with the MUR-PU and the Steam Generator Tube Rupture (SGTR) and Margin to Overfill (MTO) are currently under development and will be installed in accordance with 10 CFR 50.59. The impact of these modifications on the fire protection program will be evaluated in accordance with Exelons design change process.

NRC Request 6 The NRC staff notes that the LAR, Attachment 5, MUR Technical Evaluation,Section VII.6.A.i, "Fire Protection Systems," on page VII-6, states that the fire protection water is utilized to supply the spent fuel pool and cooling water to the centrifugal charging pumps. Are there any other uses of fire water pumps and water for non-fire protection uses at Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2? If so, the MUR power uprate LAR should identify the specific situations and discuss to what extent, if any, the MUR power uprate affects these "non-fire-protection" aspects of the plant fire protection system. In your response discuss how any non-fire suppression use of fire protection water will impact the need to meet the fire protection system design demands.

Response There are no design basis accidents or transients (other than fire) that credit the use of the fire protection (FP) system at Braidwood and Byron. However, there are situations in which Braidwood and Byron would use the FP system as a water source in Abnormal Operating Procedures and other procedures for situations where there is an unlikely failure or the design

basis sources of water are unavailable. As described below, the MUR-PU has no impact on the use of the FP system in these situations. The UFSAR Section 9.1.3.3 indicates in the unlikely event of a failure of the return line to the spent fuel pool (SFP), the SFP losses can be made up from the fire protection system. Braidwood procedure 1(2)BwOA REFUEL-2 and Byron procedure 1(2)BOA REFUEL-2, both titled Refueling Cavity or Spent Fuel Pool Level Loss, provide guidance to use the FP system as a last resort of non-borated make-up water to the

SFP. Under MUR-PU conditions, the heat load on the spent fuel pool is expected to increase slightly. However, because the maximum evaporation rate from the spent fuel pool under current conditions is much less than the makeup capacity of the fire protection system, and because this spent fuel pool makeup function would not be required when the fire protection system is called upon for fire protection functions, the

supply of water from the fire protection system for this makeup function is still adequate. UFSAR Table 9.2-11, Note 6 states: "An alternate cooling source is available to the centrifugal charging pumps by use of temporary hoses from the Fire Protection System (not credited in any design basis accident)."

Braidwood procedure 1(2)BwOA PRI-8, and Byron procedure 1(2)BOA PRI-7, both titled Essential Service Water Malfunction, give guidance to provide emergency cooling to the charging pumps in the unlikely event that essential service water is not available. The MUR-PU does not affect the centrifugal charging pump flow rate or fluid temperatures or pressures. Therefore, there is no effect

on the capability of the fire protection system to provide cooling water to the centrifugal charging pumps.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 18 NON-PROPRIETARY For Braidwood only, procedure BwOP SA-1, Startup and Operation of Station Air Compressors, provides guidance to use the FP system to supply emergency cooling water to station air compressors (SAC) in the event of loss of non-essential service water. The MUR-PU has no impact on the SACs, therefore, there is no effect on the capability of the fire protection system to provide cooling water to the SACs. For Byron only, procedure 0BOA PRI-7, Loss of Ultimate Heat Sink, provides guidance to use the FP system to supply alternate cooling water to the emergency diesel generators (EDG) in the unlikely event that essential service water is not available. The MUR-PU does not increase the electrical load demand on the EDGs and does not change the EDGs load profile. Therefore, the MUR-PU has no effect on the capability of the FP system to provide alternate cooling water to the EDGs. Use of the FP system is credited in certain security event scenarios. The MUR-PU does not affect these scenarios. Therefore, based on the above, the crediting of non-fire-protection uses of the fire protection systems remains valid under MUR-PU conditions. The MUR-PU does not add any new non-fire-protection uses of the fire protection systems or change the demand on the fire protection system. NRC Request 7 Section 3.4.5 of Attachment 1, (Plant Modifications) includes a list of modifications of interest. The last two bullets of that list include various Balance of Plant instrument rescaling, setpoint and alarm changes and ATWS Mitigation System time delay changes. In addition,Section VII.2.B of attachment 5 states that the MUR power uprate modification will implement the changes that are required to certain non-safety related systems, including Control Room displays and alarms. There is no other information regarding what these changes might look like or how the licensee plans to validate them. Please provide additional information regarding what potential modifications might be included, and how they will be verified and validated.

Response As stated in Section 3.4.5 of Attachment 1 and Section VII.2.B of Attachment 5 of the MUR-PU LAR submittal (Reference 1), the noted changes will be made in accordance with the requirements of 10 CFR 50.59; and verified and validated in accordance with the Exelon Configuration Change Control for Permanent Physical Plant Changes, process. Specifically, modifications identified that impact operator indications or responses will be implemented using this process. This change control process will initiate necessary changes to the simulator

hardware, procedures, and operator training. BOP instrument rescaling, setpoint and alarm point changes will not impact any control room controls, displays, or alarms. These modifications increase the scaling range for several computer points to bound MUR-PU process conditions. An instrument calibration for the new scale will be required. However, there are no physical changes required in the control room. The ATWS Mitigation System time delay involves the reduction in the ATWS Mitigation System Actuation Circuit (AMSAC) time delay relay setting to approximately 10 seconds. This change

does not have any operator impact or burden to the operator with respect to mitigating the

consequences of an ATWS event.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 19 NON-PROPRIETARY NRC Request 8 Section VII.2.A of Attachment 5, states that changes to emergency operating procedure (EOP) and abnormal operating procedure will be made in conformance with the Westinghouse EOP Setpoint Methodology. There is no additional information regarding this methodology. Please provide a description of the Westinghouse EOP Setpoint Methodology, or a reference if it is an approved methodology.

Response The Westinghouse EOP Setpoint Methodology term used in Section VII.2.A of Attachment 5 is referring to the Westinghouse Emergency Response Guidelines (ERGs), ERG Footnote Basis Document. The ERG Footnote Basis Document is part of the ERGs. The Westinghouse Emergency Response Guidelines were accepted as documented in Generic Letter 83-22, "Safety Evaluation of "Emergency Response Guidelines," which transmitted the associated NRC Safety Evaluation.

NRC Request 9 RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Uprate Applications" asks for information on this topic: "A statement confirming licensee intent to revise existing plant operating procedures related to temporary operation above 'full steady-state licensed power levels' to reduce the magnitude of the allowed deviation from the licensed power level. The magnitude should be reduced from the pre-power uprate value of 2% to a lower value corresponding to the uncertainty in power level credited by the proposed power uprate application." This was not addressed in the submittal. Please provide information to assure that they have revised the administrative controls for preventing inadvertent excursions above the new 100

percent power level.

Response Consistent with the NEI Position Statement, Guidance to Licensees on Complying with the Licensed Power Limit, as endorsed by the NRC, Byron and Braidwood Stations General Operating Procedures (BGP 100-3, Power Ascension, and BwGP 100-3, Power Ascension

5% to 100%,

respectively) proactively direct operator actions to maintain licensed power levels at or below 100%. The following guidance is provided in these procedures. During full steady state power operation, operators are directed to monitor reactor power using the 10-minute calorimetric.

o In the event the 10-minute calorimetric exceeds 100%, during steady state operation, operators are directed to initiate actions within 15 minutes to restore

the 10-minute calorimetric to less than 100%.

o While temporary power excursions greater than 100% power may occur due to reactor coolant system temperature changes, secondary plant efficiency changes, etc. and are allowable, operators are directed that the one-hour calorimetric should not be allowed to exceed 100%.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 20 NON-PROPRIETARY For planned activities that are expected to cause an increase in reactor power such that the 10-minute calorimetric will be greater than 100%, operators are directed to initiate actions such that the 10-minute calorimetric does not exceed 100%. For unplanned activities that cause the 10-minute calorimetric to exceed 100%, operators are directed not to wait for the transient to subside, but take prompt corrective

action to limit the time that the 10-minute calorimetric exceeds 100%. Since these Byron and Braidwood Station procedures refer to the 100% reactor power level and not a specific thermal power level, no procedure revisions are required to prevent operation above the licensed power level to address the change in the measurement uncertainty. The Braidwood and Byron calorimetric application on the Plant Process Computer (PPC) will be revised for the MUR-PU implementation and will execute three simultaneous calculations of

reactor power. 1) LEFM CheckPlus Based Reactor Thermal Power When the LEFM CheckPlus system is operable, reactor power will be calculated utilizing the LEFM CheckPlus derived flow, temperature and pressure. As described above, reactor ther00% reactor thermal power assumes the improved accuracy associated with the LEFM system

(+/-0.37%). 2) Normalized Reactor Thermal Power If the LEFM CheckPlus system becomes inoperable reactor thermal power will be calculated utilizing the existing feedwater venturis normalized by a multiplication factor to the LEFM CheckPlus based calorimetric prior to the LEFM becoming inoperable. As described above, reactor thermal power will be restricted to 3) Venturi Based Reactor Thermal Power If the LEFM is not returned to an operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (the Completion Time allowed for in the proposed new TRM section provided in the MUR-PU LAR submittal, (Reference 1) Attachments 3A and B), the multiplication factors are no longer used and the calorimetric will calculate reactor power based on feedwater venturi flows. Reactor 3586.6 MWt (98.3% of MUR-PU LTP)) to address the 2% uncertainty associated with the venturi based calorimetric. All calculations will occur automatically within the PPC calorimetric application. The existing uncorrected venturi-based feedwater flow will continue to be maintained and used for feedwater control and other functions.

NRC Request 10 Attachment 5 Section III.15.1 - Explain the Input Modification data base and how it is used.

Response The input modification data base is used to store information on the various Westinghouse reactor designs. Stored data consists of geometry data (volumes, flow areas, diameters, metal mass, etc) for the reactor vessel, reactor coolant loops, pressurizer, steam generators and reactor coolant pumps. This data is then accessed by a processor to construct the SATAN78

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 21 NON-PROPRIETARY input deck. SATAN78 is the first computer code used in the calculation of LOCA mass and energy (M&E) releases and calculates the blowdown releases.

NRC Request 11 Attachment 5 Section III.15.1 - Why were corrections necessary to the reactor coolant pump homologous curves? What were these corrections? Did the corrections significantly affect the results of the mass and energy release analyses?

Response The Reactor Coolant Pump (RCP) homologous curves are input by hand to the WREFLOOD computer code. In preparing the input of the homologous curves a minus sign was left out of the data string describing the head developed by the RCP. This resulted in a mis-prediction of the RCP hydraulic loss. When corrected a small benefit was noticed for the Byron/Braidwood

analyses.

NRC Request 12 Attachment 5 Section III.15.1 - Why is it assumed that containment spray is terminated at eight hours? What is the current assumption in the mass and energy release calculations and why was this assumption changed?

Response The eight hour containment spray termination assumption is consistent with the earliest Containment Spray termination permitted following a Large Break LOCA by the Emergency Operating Procedures. The prior analysis assumed that Containment Spray operated indefinitely following a Large Break LOCA event. The revised analysis changed the termination time to be consistent with the assumption for iodine removal previously approved as part of the

Alternate Source Term License Amendment (Reference 5) and to provide conservative results

for the containment response.

NRC Request 13 Attachment 5 Section III.15.1 - Why is the barrel baffle metal mass only included for upflow design plants? How significant is the omission of the barrel baffle metal mass in the current analyses?

Response The barrel baffle metal mass is to be included in all Westinghouse LOCA M&E analyses regardless if the plant is an upflow or downflow barrel baffle design. However, it was discovered that the upflow barrel baffle metal mass had been erroneously omitted for reactor vessel designs having an upflow barrel baffle. The MUR-PU LAR submittal (Reference 1) identified

that the barrel baffle metal mass omission was limited to Unit 2, however all four Byron/Braidwood units are of the upflow barrel baffle design and were missing approximately 50,000 lbms of metal for each plant. Note this issue has been resolved for all four units in both the current analysis and the MUR-PU analysis. The effect was a small penalty on the Double Ended Hot Leg (DEHL) blowdown peak (~0.05 psi) and a somewhat larger penalty for the long-term calculated peak pressure for the Double Ended Pump Suction break (~0.3 psi).

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 22 NON-PROPRIETARY NRC Request 14 Attachment 5 Section III.15.1 - Please describe or reference a description of the "identified inconsistencies" in the EPITOME computer code. Did these inconsistencies significantly affect the current mass and energy release calculation results?

Response The EPITOME computer code reads the LOCA mass and energy releases calculated by the SATAN78, WREFLOOD and FROTH computer codes and creates tables of instantaneous mass and energy releases, and also mass balance and energy balance tables. These tables are used in the FSAR and licensing reports. EPITOME also creates a binary file of the instantaneous mass and energy releases to be read by the containment computer code. The inconsistency noted in the EPITOME results was a difference between the data tables and the binary file data. The integrated break mass and energy release from the mass and energy

balance tables, when compared to integrated results from the binary file data shows a large

difference in the releases between the end of reflood and the steam generator equilibrium points. An investigation determined that the instantaneous releases from the binary file were missing some mass and energy between the end of reflood and the first or second steam generator equilibrium point. Once the binary file instantaneous releases were corrected there was an increase in the mass and energy released to the first steam generator equilibrium point for Byron/Braidwood. Note that this issue is addressed in the MUR-PU analysis. There were sufficient margins available in the current analysis such that no revision to P a was necessary.

NRC Request 15 Attachment 5 Section III.15.3 - Describe or reference the modeling of the Unit 1 Babcock and Wilcox steam generators and the Unit 2 Westinghouse Model D5 steam generators with respect to mass and energy release calculations and explain why this modeling is conservative for mass and energy release calculations.

Response The calculational models described in the NRC approved WCAP-10325-P-A, Westinghouse LOCA Mass and Energy Release Model for Containment Design, March 1979, and referenced WCAPs, were applied in the same manner and are applicable to both steam generator designs at Byron and Braidwood Stations. The overall conservatism of the WCAP-10325 model and supporting WCAPs has not changed and was used in the Byron and Braidwood MUR-PU analyses. This methodology was previously utilized and approved for Byron and Braidwood Stations as part of the stretch power uprate (Reference 6). The Byron and Braidwood Unit 1 Babcock and Wilcox replacement steam generators are similar in design to Westinghouse steam generators that they have vertical U-tubes for the heat transfer. The Westinghouse LOCA M&E model was specifically developed to model vertical U-tube steam generators and has the flexibility to model any manufacturers steam generator as long as it is of the vertical U-tube design. Differences in steam generator design are related to differences in heat transfer area, flow area, hydraulic loss, secondary mass, etc. The correct geometry, such as tube diameter, wall thickness, fluid volume, flow area, metal mass, and surface area are input to the plant specific model.

These differences in design between the two

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 23 NON-PROPRIETARY steam generators were accounted for in the Byron and Braidwood LOCA Mass and Energy

releases.

NRC Request 16 Attachment 5 Section III.15.4 and Attachment 5 Section III.16.4 - Describe or reference the computer codes and the assumptions used to derive the electrical equipment environmental qualification temperature and pressure profiles.

Response The computer code and assumptions for the Main Steam Line Break Mass and Energy Release Inside Containment Analysis are discussed in the MUR-PU LAR submittal (Reference 1), Attachment 5, Sections III.16.2 and III.16.3. The computer codes and assumptions for the LOCA Mass and Energy Release Analysis are discussed in the same submittal (Reference 1), , Sections III.15.2 and III.15.3. The electrical equipment environmental qualification (EQ) temperature profile is a composite curve that bounds the results from both the Main Steam Line Break Mass and Energy Release Inside Containment and the LOCA Mass and Energy Release Analyses. The electrical equipment EQ pressure profile is a curve that conservatively assumes the pressure equals the containment design pressure for the first twenty minutes of the event then corresponds to

saturated ambient conditions for the remaining duration of the event.

NRC Request 17 No response required.

NRC Request 18 Attachment 5 Section III.16.3 - Please describe the modeling of the two steam generator types for the MSLB analyses, especially those characteristics affecting the MSLB results, e.g., mass of water, location of nozzle, etc.

Response The Byron and Braidwood analysis used conservative steam generator modeling that is consistent with the NRC approved methodology found in WCAP-8822, Mass and Energy

Releases Following a Steam Line Rupture, September 1976. The Main Steam Line Break (MSLB) analyses use the LOFTRAN code (methodology found in WCAP-8822). The steam generator model is described in Section 4 of WCAP-7907-P-A, LOFTRAN Code Description, April 1984. The user input consists of geometric parameters and the initial thermal/hydraulic conditions, including initial steam generator (SG) water mass. The important SG input parameters that impact the MSLB results are: Initial SG water mass - this value has been set conservatively high. The secondary SG water volume at which the SG tubes are assumed to start to uncover - this value has been set conservatively low to maximize the primary-to-secondary side

heat transfer. The quality transient of the break effluent is input by the user. It is set conservatively high to maximize the vapor release which maximizes the containment pressure.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 24 NON-PROPRIETARY NRC Request 19 Attachment 5 Table III.16 What is meant by a composite curve? How is it determined?

Response The Main Steam Line Break (MSLB) mass and energy release composite curves provide the most limiting pressure or temperature at any given time during a MSLB transient based on the full spectrum of cases analyzed. A description of what the MSLB composite curves are based

on and how they are derived is provided in UFSAR Section 6.2.1.4.1, Pipe Break Blowdown -

Spectra and Assumptions.

NRC Request 20 Attachment 5 Section VI.1.F.ii - This section states that the heat loads in a limited number of areas did increase due to the MUR and that the increase was minimal. Please state which areas would experience an increase in heat load and the magnitude of the increase. The maximum value of these increases is sufficient.

Response The evaluation to support the MUR-PU evaluated the potential sources of heat input into the Auxiliary Building; however it was not a comprehensive room-by-room evaluation. The evaluation concluded that the Auxiliary Steam System is the only potential source of increased heat input to areas of the Auxiliary Building. The evaluation determined that there was no impact to ESF cubicles. The Auxiliary Steam System is fed by the Auxiliary Boiler and by Extraction Steam from the High Pressure Turbine. The Auxiliary Boiler is not impacted by the MUR-PU. The temperature of extraction steam system is expected to increase slightly (2.5 °F, 0.6%) with MUR-PU conditions; therefore the Auxiliary Steam system temperature is also expected to increase slightly. The consequent increase in heat load is not considered significant. The Auxiliary Steam system provides heating for various auxiliary services (e.g., batching Boric Acid in the Auxiliary Building). Therefore any potential increases in temperature to the Auxiliary Building areas could only be impacted by the small increase in the temperature of the Auxiliary Steam. It should also be noted that the decrease in Steam Generator Blowdown temperature is expected to offset the increase in Auxiliary Steam. Therefore, minimal impact on the areas of the Auxiliary Building is expected.

NRC Request 21 Engineering Report ER-800, Revision 1 (ADAMS Accession No. ML 111790063), contains several appendices labeled "A.1, A.2, A.3, A.4, & A.5." These appendices contain detailed calculations, the results of which appear to be summarized in Appendix C, Table I. The NRC staff is having trouble identifying the equations in the approved topical report that correspond to the calculation in these appendices. In order to confirm implementation please provide a detailed and explicit cross reference between the June 23, 2011, letter (ADAMS Accession No. ML111790030), Attachment 8a, (ADAMS Accession No, ML111790063) and the associated approved topical report equations (i.e., between ER-800 Revision 1, Appendix A.1, and ADAMS Accession No. ML102950246).

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 25 NON-PROPRIETARY Response As discussed with the NRC on October 13, 2011, during an audit of the Engineering Report ER-800, the NRC agreed to allow Exelon to provide a specific example of the development of one of

the uncertainty terms in lieu of the detailed and explicit cross reference requested above.Appendix C, Table I, of ER-800, "Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 1 Using the LEFM CheckPlus System," Revision 1, provides a summary of individual uncertainty terms that contribute to the determination of total thermal power uncertainty. This table refers to values specific to Byron Unit 1 as well as typical values obtained from Appendix A, Table A-1 in Camerons approved topical report, ER-157P-A, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System, Revision 8 (Reference 7) (herein further referred to as ER-157P-A). The topical report uses values, representative of a typical PWR plant, to determine the values in ER-157P-A, Appendix A, Table A-1. Although these values may differ from the Byron Unit 1 values in ER-800, Appendix C, Table I, the equations and methodologies are consistent.

Therefore, the values listed in ER-800, Appendix C, Table I are traceable to the approved topical report ER-157P-A. A systematic approach can be utilized to demonstrate traceability between each uncertainty term listed in ER-800, Appendix C, Table I and the approved topical report (ER-157P-A). The key steps necessary that describe this process are listed below: 1. Identify the uncertainty term description (e.g., Profile Factor) that will be traced from ER-800, Appendix C, Table I. 2. Locate the equivalent term description from ER-157P-A, Appendix A, Table A-1. 3. Obtain the item number that corresponds to the uncertainty term description. The item number is listed in front of the uncertainty term in ER-157P-A, Appendix A, Table A-1. For example, the item number for Profile Factor is 1. 4. Using the item number and the uncertainty term description, locate the corresponding section in the main body of ER-157P-A, Appendix A. 5. Within this section, locate the equations and/or methodologies used to determine the uncertainty term. Use the same equations and/or methodologies to calculate the Byron Unit 1 uncertainty term by applying the appropriate plant specific values from ER-800 or ER-829, Meter Factor Calculation and Accuracy Assessment for Byron Unit 1. An example that provides a detailed and explicit cross reference, using the above steps, is listed

below for the Profile Factor uncertainty term. The first step in understanding this uncertainty term involves the overview discussion in ER-800, Section 4.0 (page 5). The subsection titled Appendix A.3, Meter Factor (Calibration)

Uncertainties describes the elements that are used to calculate the Profile Factor uncertainty.

(Note that Cameron uses the terms Meter Factor and Profile Factor interchangeably.)

[

] a,b,c These uncertainties are addressed in ER-800, Appendix A.3, which provides a cross-reference to ER-829.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 26 NON-PROPRIETARY Following the key steps listed above applied for the Profile Factor as an example. 1. The Profile factor uncertainty term is found as the first uncertainty term in ER-800, Appendix C, Table I (shown under Hydraulics). 2. The Profile Factor uncertainty term is correspondingly found as the first uncertainty term ER-157P-A, Appendix A, Table A-1. 3. The Profile Factor uncertainty is labeled as item 1 (denoted by 1 preceding the uncertainty term in the table). 4. The corresponding section in ER-157P-A, Appendix A, is titled Profile Factor; Item 1, Table A-1 and is located on page A-17. 5. The equation used to determine the "Profile Factor" is shown in ER-157P-A, Appendix A, page A-18 as follows:

Equation with Topical Report Values

[

] a,b,c 1 Using this equation and replacing the element values with the description of elements, the equation is restated below. Equation with Topical Report Descriptions

[

] a,b,c Equation with Byron Unit 1 Values Substituting Byron Unit 1 specific values from ER-829, Table 5 (page 27) into the above equation yields the following:

[

] a,b,c The Byron Unit 1 plant specific Profile Factor of 0.17% (ER-800, Appendix C, Table I) differs from the value determined in the topical report (ER-157P-A) due to differences in values of the elements. However, the equation used to determine the Profile Factor uncertainty is consistent with ER-157P-A. A summary table is listed below that compares the values of the Profile Factor elements from ER-157P-A to the Byron Unit 1 application. The values listed for ER-157P-A are obtained from the Profile Factor uncertainty table on pages A-17 and A-18 in ER-157P-A. The values listed for the plant specific Byron Unit 1 application are obtained from ER-829, page 27.

1 The reference to +/-0.22% from ER-157P-A, Revision 8, was misquoted as +/-0.25% when it was transferred over to ER-800, Appendix C, Table I and is further discussed in the response to RAI Request 24 below.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 27 NON-PROPRIETARY a, b ,c Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 28 NON-PROPRIETARY

[ ] a,b,c NRC Request 22 Please confirm that the assumptions listed in Cameron Caldon Ultrasonics Engineering Report No. ER-157(P-A) Revision 8 and Revision 8 Errata (ADAMS Accession No ML102950246) Appendix A are valid for the Byron and Braidwood application.

Response Camerons Engineering Report ER-157P-A (Reference 7) is based on an analysis of a typical 2-loop pressurized water reactor or 2-feedwater line boiling water reactor. The assumptions in ER-157P-A, page A-1 of Appendix A, as applied to Byron and Braidwood are discussed below.

The majority of the assumptions use plant-specific values. In these cases, plant-specific values are reflected in the Byron and Braidwood Cameron engineering reports (e.g. ER-800, Rev. 1 for Byron Unit 1). Therefore, the calculated feedwater mass flow uncertainty and the total thermal power uncertainty reflect plant-specific values. Specific information involving each of the

assumptions is provided below. 1. The LEFM system measures total feedwater flow. The plant may be either a PWR or a BWR.The LEFM measures approximately 99% of total feedwater flow in the Byron and Braidwood application. The remaining 1% is measured through a tempering flow line that bypasses the LEFM. The tempering flow measurement uncertainty is determined in a plant specific calculation and then combined with other plant specific uncertainty terms. The total plant-

specific gains/losses uncertainty term is then combined with Cameron uncertainty terms in Appendix B, Table B-1 of the Byron and Braidwood Cameron engineering reports to determine the total thermal power uncertainty. 2. The feed line has an internal diameter of 27 inches, and a feedwater velocity of 20 ft/sec. The Byron and Braidwood application utilizes nominal 16 feedwater lines. The spool piece internal diameter and feedwater velocities are reflected in the Byron and Braidwood Cameron engineering reports. The resulting uncertainties in the reports were calculated based on the actual plant specific spool piece dimensions and velocities. 3. The feedwater flow measurement is downstream of the final stage of feed heating. The LEFM thus measures final water temperature. The LEFM spool piece is located downstream of the final feedwater heating for the Byron and Braidwood application and therefore measures final feedwater temperature. Therefore, this assumption is satisfied. 4. The steam pressure for the PWR and the BWR is 1000 psia. This assumption is replaced with the appropriate plant specific value.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 29 NON-PROPRIETARY

5. The pressure in the feedwater line at the point of flow measurement is measured with an accuracy of +/- 15 psi or better. A plant specific feedwater pressure measurement uncertainty calculation determined that the uncertainty is within +/- 15 psi. Therefore, this assumption is satisfied. 6. The steam supply pressure is measured by one or more utility supplied transmitters having an accuracy of +/- 15 psi. Departures from the assumed accuracy of this instrument, as well as other elements of the thermal power determination that are not part of the LEFM system, will be treated in the site specific uncertainty analysis submitted as part of his Appendix K

uprate package. This assumption is satisfied as the appropriate plant specific value is used. 7. The final feed temperature is 430

°F.This assumption is replaced with the appropriate plant specific value.

8. Fluid properties at and adjoining the operating point are noted in Attachment 1, excerpted from the 1967 ASME Steam Tables. The 1967 ASME Steam Tables were used in the plant specific analysis; this assumption is satisfied 9. Reactor Thermal Power is determined from the following equation: (1) P R = W FT (h s - h fw) +/- P LOSS Byron and Braidwood utilize the same equation as referenced above subject to plant specific contributions, such as tempering flow. 10. All errors and biases will be calculated and combined according to the procedures defined in ASME-PTC-19.1 (1985). This document defines the contributions of individual error elements through the use of sensitivity coefficients defined as follows: (Refer to topical report ER-157(P-A), Rev. 8, page A-3 for a detailed list of the equations) This assumption is applicable to the Byron and Braidwood Cameron engineering reports. 11. The LEFM algorithm is as follows: (Refer to topical report ER-157(P-A), Rev. 8, pages A-3 and A-4 for the equation) This assumption is satisfied and is reflected in the Byron and Braidwood Cameron engineering reports. 12. Spool piece and other data are in accordance with Attachment 2 to this Appendix. Byron and Braidwood plant specific values have been used and are reflected in the Byron and Braidwood Cameron engineering reports.

NRC Request 23 Regulatory Guide 1.105 Rev. 3, "Setpoints for Safety-Related Instrumentation," dated December 1999, describes a method for combining individual uncertainty terms in quadrature; this method assumes that the individual term each meet the 95/95 criteria. Please describe how each individual uncertainty term meets the 95/95 criteria.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 30 NON-PROPRIETARY Response Consistent with ASME PTC 19.1-1985 (Reference 4 in ER-800 and Assumption 10 in ER-157P-A), each individual uncertainty term listed in ER-800, "Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 1 Using the LEFM CheckPlus System," Revision 1, Appendix C, meets the 95/95 criteria through one of two methods, depending on the sample size of test data. When a large sample size of data is available, there is a 95% probability that the constructed limits of each uncertainty term contain 95% of the population of interest for the surveillance interval selected. For example, the first uncertainty term, Profile Factor, is comprised of several elements as listed in ER-829, Meter Factor Calculation and Accuracy Assessment for Byron Unit 1, Table 5, page 27. In particular, Modeling Sensitivity, uses 127 different meters (with 497 different model calibrations), including approximately four to five parametric tests per meter. This large sample size of data facilitates a comprehensive statistical analysis. Refer to ER-829, Section 4.4, page 30, for a further discussion of the statistical approach. When only a small sample size of data is available, a Students t-distribution method is used. The Students t-distribution is a statistical methodology that, along with the standard deviation, is

used to determine the 95% confidence interval of a small population of samples. For example, another element of Profile Factor is Data Scatter. The 95% confidence limits for the Data Scatter element are calculated using the Students t-distribution for the A/B/C and D flow elements as described in ER-829, Section 4.5. The remaining uncertainty terms, listed in ER-800, Appendix C, Table I, use either of the above methods, depending on sample size, to meet the 95/95 criteria.

NRC Request 24 Table I, "Reconciliation of Byron Unit 1 Nuclear Generating Station Uncertainties with Cameron Reports," of Appendix C (page 5) of Cameron Engineering Report ER-800 Revision 1 (ADAMS Accession No ML111790063) was compared with Table A-1, "Representative Thermal Power Uncertainties for a Total Feedwater Flow Measurement in a PWR or BWR Using Chordal LEFM Check and LEFM CheckPlus" of ER-157(P-A) Revision 8 and Revision 8 Errata (ADAMS Accession No.

ML102950246).The Byron document seems to misquote the numbers in the approved topical report in some places, for example: Table I identifies the ER-157P value for the Hydraulics Profile Factor as being "+/-

0.25%" while the value in ER-157(P-A) Rev.8 And Rev.8Errata is "+/- 0.22%." Table I identifies the ER-157P value for the Time Measurements as being "+/- 0.05%" while the value in ER-157(P-A) Rev.8 And Rev.8Errata is "+/- 0.06%." The Byron document also seems to indicate that in some cases the Byron Unit 1 system is credited as being better than the bounding topical report, for example: Table I identifies the Byron Unit 1 value for the Subtotal Mass Flow Uncertainty as being "+/- 0.26%" while the value in ER-157(P-A) Rev.8 And Rev.8Errata is "+/-

0.28%."

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 31 NON-PROPRIETARY Table I identifies the Byron Unit 1 value for the Feedwater Density and Feedwater Enthalpy as being "+/- 0.14%" while the value in ER-157(P-A) Rev.8 And Rev.8Errata is "+/- 0.15%."

Please explain.

Response Table I of Appendix C (page 5) of the Cameron Engineering Report, ER-800, "Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 1 Using the LEFM CheckPlus System" did misquote the value for the Hydraulics Factor Uncertainty and the Time Measurement Uncertainty given in ER-157P-A, Revision 8 (Reference 7). The ER-157P-A Revision 8 value of +/-0.22% for the Hydraulics Profile Factor" uncertainty and the value of

+/-0.06% for the Time (Time-of-Flight) Measurement uncertainty should have been used in ER-800. These values were also misquoted in Table I of the similar Engineering Reports for the

three other units: ER-801, Revision 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 2 Using the LEFM CheckPlus System," ER-802, Revision 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Braidwood Unit 1 Using the LEFM CheckPlus System," ER-803, Revision 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Braidwood Unit 2 Using the LEFM CheckPlus System." These errors do not impact the results or conclusions of the Engineering Report as they represent the values for a generic LEFM CheckPlus installation and were provided for comparison purposes only to the site specific values. Table 1 of Appendix C (page 5) of the Cameron Engineering Report, ER-800, "Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 1 Using the LEFM CheckPlus System" provided the site specific values of +/-0.26% for the Subtotal Mass Flow uncertainty and +/-0.14% for the Feedwater Density and Feedwater Enthalpy Combined Uncertainty. These are site specific values based on non-fluid tau's, actual spool piece measurements, Alden profile uncertainty values, etc.

Braidwood/Byron Stations MUR LAR Response to RAI November 1, 2011 , page 32 NON-PROPRIETARY References

1. Letter from Craig Lambert (Exelon Generation Company, LLC) to U. S. NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23, 2011 2. Westinghouse Calculation Note CN-REA-10-44, Revision 0, Byron/Braidwood Fluence Analysis to Support the Measurement Recapture Uncertainty (MUR) Program, B. W.

Amiri, September 2010. 3. Westinghouse Calculation Note CN-AMLRS-10-07, Revision 0, Braidwood Units 1 and 2 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations, A. E. Leicht, September 2010 4. Westinghouse Calculation Note CN-AMLRS-10-08, Revision 0, Byron Units 1 and 2 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations, A. E. Leicht, September 2010 5. Letter from R. F. Kuntz (NRC) to C. M. Crane (EGC), dated September 8, 2006 Byron Station. Units Nos. 1 and 2 and Braidwood Station, Units Nos. 1 and 2 - Issuance of Amendments Re: Alternate Source Term (TAC Nos. MC6221, MC6222, MC6223, MC6224 6. Letter from G. F. Dick (USNRC) to O. D. Kingsley (Exelon), Issuance of Amendments: Increase in Reactor Power, Byron Station Units 1 and 2, Braidwood Station, Units 1 and

2. (TAC NOS. MA9428, MA9429, MA9426 and MA9427), ML011420274, May 4, 2001 (Amendments 119/113) 7. Topical Report (TR) Engineering Report ER-157P, Revision 8, Caldon Ultrasonics Engineering Report ER-157P, Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM Check or CheckPlus System, dated May 11, 2009 (ML091340322)