ML12286A014

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License Amendment Request - Extended Power Uprate, Attachment 4, NEDO-33566, Rev. 0, Safety Analysis Report, Constant Pressure Power Uprate, Page 2-223 Through Page 2-358
ML12286A014
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 09/30/2012
From:
GE-Hitachi Nuclear Energy Americas
To:
Office of Nuclear Reactor Regulation
References
DRF 0000-0107-2302 NEDO-33566, Rev 0
Download: ML12286A014 (136)


Text

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.5.6 Additional Considerations 2.5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System Regulatory Evaluation Nuclear power plants are required to have redundant onsite emergency power supplies of sufficient capacity to perform their safety functions (e.g., power diesel engine-driven generator sets), assuming a single failure. The review focused on increases in emergency diesel generator electrical demand and the resulting increase in the amount of fuel oil necessary for the system to perform its safety function. The regulatory acceptance criteria for the emergency diesel engine fuel oil storage and transfer system are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects, including missiles, pipe whip, and JI forces associated with pipe breaks; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-17, insofar as it requires onsite power supplies to have sufficient independence and redundancy to perform their safety functions, assuming a single failure.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-4, Draft GDC-24, Draft GDC-39, Draft GDC-40, and Draft GDC-42.

Current GDC-17 is applicable to PBAPS as described in UFSAR Sections 8.4.8 and 8.5.6, "Compliance with Safety Guides."

The Diesel Engine Fuel Oil Storage and Transfer capability is described in PBAPS UFSAR Section 8.5, "Standby AC Power Supply and Distribution," and Appendix F, "Interaction of Units 2 and 3."

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In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the Diesel Engine Fuel Oil Storage and Transfer capability is documented in NUREG-1769, Section 2.3.3.16. Management of aging effects on the Diesel Engine Fuel Oil Storage and Transfer capability is documented in NUREG-1769, Section 3.3.16.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.8 of the CLTR addresses the effect of CPPU on other systems not addressed in the CLTR. It concludes that systems not specifically addressed in the CLTR are not significantly affected by the power uprate. The Emergency Diesel Engine Fuel Oil Storage and Transfer system is not addressed in the CLTR, and this disposition applies to PBAPS.

There is no change to ESW loads with EPU. EPU conditions are achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the required pump motors.

In addition, UFSAR Section 8.5.2 defines the time periods verses loading requirements for each diesel. No increase in electrical equipment demand on the EDGs is expected as a result of EPU with the exception of additional HPSW pump motor and RHR heat exchanger cross-tie MOV loads. These changes to EDG loads have been evaluated and shown to remain within the EDG 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> (3000kW) ratings. Therefore, under emergency conditions, the electrical supply and distribution components are adequate.

The RHR heat exchanger cross-tie modification requires the loading of an additional HPSW pump for long term containment cooling. This additional loading was evaluated and found to be acceptable.

No increase in flow or pressure is required of any other AC powered ECCS equipment. Therefore, the amount of power required to perform safety-related functions (pump and valve loads) is not increased with EPU, and the current emergency power system remains adequate. The changes to the maximum EDG loading are those associated with the additional HPSW pump motor and MOV loads associated with the RHR heat exchanger cross-tie modification.

EDG fuel oil consumption based on the RHR heat exchanger cross-tie modification which will cause an increase in the total EDG loading limit beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and resulting positive margins are based on a proposed increase in the TS requirement for the EDG Fuel Oil Storage Tanks and credit for the EDG Fuel Oil Day Tanks.

HPSW, ESW and ECW flow requirements are not changing with EPU except that one additional HPSW pump will be credited at one hour post LOCA as part of the RHR heat exchanger cross-tie modification to eliminate CAP credit for ECCS pump NPSH. The changes to EDG loading associated with the RHR heat exchanger cross-tie modification have been evaluated and shown to remain within the EDG 2000 hr. ratings. Individual pump load demand will not change for 2-224

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EPU. No physical changes to the Emergency Diesel Engine Fuel Oil Storage and Transfer system are necessary, except for the fuel oil storage tank level alarm setpoint changes. A change to the TSs (Section 3.8.3) to increase the minimum required EDG fuel oil storage capacity from 31,000 gallons per tank to 33,000 gallons per tank is required.

Conclusion The amount of required fuel oil for the EDGs has been reviewed and found to adequately account for the effects of EPU on fuel oil consumption. Exelon concludes that the fuel oil storage and transfer system will continue to provide an adequate amount of fuel oil to allow the diesel generators to meet the onsite power requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the fuel oil storage and transfer system.

2.5.6.2 Light Load Handling System (Related to Refueling)

Regulatory Evaluation The light load handling system (LLHS) includes components and equipment used in handling new fuel at the receiving station and the loading of spent fuel into shipping casks. The review covered the avoidance of criticality accidents, radioactivity releases resulting from damage to irradiated fuel, and unacceptable personnel radiation exposures.

The review focused on the effects of the new fuel on system performance and related analyses. The regulatory acceptance criteria for the LLHS are based on: (1) GDC-61, insofar as it requires systems containing radioactivity be designed with appropriate confinement and with suitable shielding for radiation protection; and (2) GDC-62, insofar as it requires that criticality be prevented.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967.

The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

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For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-66, Draft GDC-67, Draft GDC-68, and Draft GDC-69. Current GDC-62 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 License Amendments 175 and 178 (Reference 62),

respectively.

The LLHS is described in PBAPS UFSAR Sections 10.2, "New Fuel Storage," 10.3, "Spent Fuel Storage," and 10.4, "Tools and Servicing Equipment."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the LLHS is documented in NUREG-1769, Sections 2.3.3.1 and 2.3.3.18. Management of aging effects associated with the LLHS is documented in NUREG-1769, Sections 3.3.1 and 3.3.18.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.8 of the CLTR addresses the evaluation of the impact of the EPU on several plant systems that were not addressed elsewhere in that report. The LLHS (related to Fuel Handling) is one of the systems so evaluated (see Table 2.5-8, Item 18). CLTR Section 6.8 is supported by ELTRI (Reference 2), Section 5.12 and Appendix J, also previously approved by the NRC for use as guidelines for EPUs. The results of this evaluation are described below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

Meets CLTR Other systems Meets Ds ti I

Disposition The EPU has been found to not have any significant effect on the LLHS.

The LLHS meets the CLTR disposition.

Conclusion The effects of the new fuel on the ability of the LLHS to avoid criticality accidents have been reviewed.

Exelon concludes that PBAPS has correctly applied the conclusion of the CLTR Section 6.8 as no effect. Based on this review, Exelon further concludes that the LLHS will continue to meet the requirements of the current licensing basis for radioactivity releases and 2-226

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) prevention of criticality accidents. Therefore, Exelon finds the proposed EPU acceptable with respect to the LLHS.

2.5.7 Additional Review Areas (Plant Systems)

NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.8 of the CLTR addresses the evaluation of the effect of the EPU on several plant systems that were not addressed elsewhere in that report. The systems included in this evaluation are listed in Table 2.5-8.

CLTR Section 6.8 is supported by ELTRI (Reference 2), Section 5.12 and Appendix J, also previously approved by the NRC for use as guidelines for EPUs. The results of this evaluation are described below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

~

~

i~&3/44CLTR" JI. BAPS Othe sysemsMeets CLTR Othe systDisposition The EPU has been found to not have any significant effect on the systems in Table 2.5-8.

The assessment of other systems meets the CLTR disposition.

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Table 2.5-1 Appendix R Fire Event Key Inputs Core Thermal Power MWt 3951 Rated Core Flow Mlbm/hr 102.5 Initial Dome Pressure psia 1050 Initial Water Level (AVZ) inch 562.0 Decay Heat N/A ANS 5.1-1979 (1)

Fuel Type N/A GNF2 Initial SP (torus water) Temperature OF 86 Initial containment/WW airspace pressure psia 15.05 Initial SP (torus) Water Volume ft3 125,100 Initial containment/WW airspace temperature OF 86 Initial Drywell Temperature OF 135 Initial Drywell Pressure psia 15.05 RHR Heat Exchanger K-value BTU/sec-°F 305 RHR Flow Rate gpm 8600 RHR Pump Horsepower (1 pump) hp 2000 CS Flow Rate gpm 3125 Service Water Temperature OF 86 SRV Capacity at Reference Pressure lbm/hr 800,000 psig 1080 Water source for HPCI and RCIC N/A CST CST Protected Capacity for HPCI/RCIC Supply gallons 173,000 Note:

(1)

Decay heat does not include 2-sigma uncertainty. Decay heat includes additional terms specified in GEH SIL 636 (Reference 63).

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Table 2.5-2 Appendix R Fire Event Evaluation Results Peak Fuel Cladding Temperature ('F) 1468 1490 No core heat-up2

< 1500°F Maximum Operator Action Time to Open 27.5 26.5 Not Calculated 6 See Note 3 ADS valves (minute)

PeakRPVDome

< 1325 5 1145.3 1145.3

< 1325 Pressure (psig)

Peak Suppression Pool Bulk 205 204.4 205.8

< 281 Temperature ('F)

Notes:

1. Using SAFER/GESTR-LOCA and SHEX methodologies.
2. Initial steady-state fuel cladding temperature.
3.

The maximum ADS actuation time should allow the core to remain covered with a short fuel uncovery period permitted, providing the PCT acceptance criterion is met.

4. Reactor vessel pressure remains low enough to ensure no risk of reactor vessel overpressure.
5. The peak vessel pressure is bounded by the Appendix R criteria of 1375 psig as shown in Reference 64. In the current licensing basis analysis, automatic cycling of the SRVs controls the peak RPV dome pressure, which is also the case at EPU conditions.
6. Controlled depressurization with rate of 1 00°F/ hour starting at 210 minutes after event initiation.

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Table 2.5-3 Appendix R Evaluation Results EPU Containment (SHEX) for Shutdown Method C (LPCI) without Spurious Operation of SRV - Sequence of Events 0 seconds Reactor Trips automatically, or is tripped by the operator due to fire.

Offsite Power is lost. All high pressure systems are considered conservatively unavailable.

MS turbine trip, loss of reactor FW and MSIV closure occurs due to fire or due to LOOP.

SRVs open due to high reactor pressure and subsequently close as 10 seconds reactor pressure drops. SRVs cycle several times; maintaining the reactor at high pressure until the operator can initiate corrective actions.

When indicated reactor level reaches 366.3 inches AVZ (TAF), the 26.5 minutes operator manually opens 3 SRVs, starts one RHR pump and aligns RHR in the LPCI mode.

33.4 minutes LPCI starts injecting into the core.

Reactor vessel water level is restored above TAF in the downcomer 43.8 minutes rein region.

70 minutes ASDC mode is initiated to remove long-term decay heat.

Torus cooling is initiated when 1 HPSW pump is started and aligned 150 minutes to RHR heat exchanger.

180 minutes Torus cooling interruption for 10 minutes due to dual unit interaction.

6.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> The peak suppression pool temperature of 204.4 'F is reached.

17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> Reactor reaches cold shutdown conditions 2-230

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Table 2.5-4 Appendix R Evaluation Results EPU Containment (SHEX) Shutdown Method A (RCIC) without Spurious Operation of SRV

- Sequence of Events 0 seconds Reactor Trips automatically, or is tripped by the operator due to fire.

Offsite Power is lost.

MS turbine trip, loss of reactor FW and MSIV closure occurs due to fire or due to LOOP.

SRV's open due to high reactor pressure and subsequently close as 10 seconds reactor pressure drops. SRV's cycle several times; maintaining the reactor at high pressure until the operator can initiate a corrective action.

10 minutes RCIC is automatically initiated at low RPV water level.

180 minutes Torus cooling is initiated when 1 HPSW pump is started and aligned to RHR heat exchanger.

210 minutes The RPV depressurization starts at 1 00°F/hr Torus cooling interruption for 10 minutes due to High Drywell 270 minutes Pressure concurrent with Low Reactor Pressure Vessel Pressure (HDWP/LRPVP) LOCA signal on Appendix R fire unit.

326 minutes RPV pressure reaches 150 psig.

Torus cooling interruption for 10 minutes due to dual unit interaction.

RCIC is secured due to suppression pool water level approaching the 510 minutes SRV tail pipe level limit. ASDC mode is initiated to remove long-term decay heat.

10.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> The peak suppression pool temperature of 205.8 'F is reached.

18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Reactor reaches cold shutdown conditions.

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Table 2.5-5 SGTS Iodine Removal Capacity Parameters

))

2. The PBAPS SGTS has a deluge system instead of incorporating minimum cooling flow to prevent desorption in the case of increased decay heating.

This is considered acceptable for that purpose.

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Table 2.5-6 Spent Fuel Pool Response 3 FPCCS pumps, 3 FPCCS HXs. 1,665 gpm total SFP flow, 2,400 gpm total SW flow.

Start of Max. SFP Time to Boil Makeup Flow SW Temperature Offload Temperature from Max.

Required at (0F)

(hours after Tmru Temperature Boiling shutdown)

(°F)

(hours)

(gpm) 90 80 140 11.4 49 1 RHR pump, 1 RHR heat exchanger. 5,000 gpm SFP flow, 4,500 gpm HPSW flow.

HPSW Start of Max. SFP Time to Boil Makeup Flow Temperature Offload Temperature from Max.

Required at (hours after Temperature Boiling (OF) shutdow.n)

(OF)

(hours)

(gpm) 92 150 140 6.0 88 No ri~ml 'ffload, S1nge; Failuve 2 FPCCS pumps, 2 FPCCS HXs. 1,110 gpm total SFP flow, 1,600 gpm total SW flow.

Start of Max. SFP Time to Boil Makeup Flow SW Temperature Offload Temperature from Max.

Required at (OF)

(hours after Tmru Temperature Boiling shutdown)

(0F)

(hours)

(gpm) 90 200 150 12.3 40 2-233

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Table 2.5-7 EPU TBCCW Impact Notes:

1 With restrictions on Iso-Phase Bus Duct modifications 2-234

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Table 2.5-8 Basis for Classification of No Significant Effect 2ý-235

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.4-4 2-240

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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 4 4

4 2-242

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Figure 2.5-1 Appendix R Evaluation Results EPU Containment (SHEX) for Shutdown Method C (LPCI) without Spurious Operation of SRV - Suppression Pool Temperature PBAPS EPU App R Case CIB Suppression Pool Temperature 250 200 I-Cu E 100 so I-I 50 0

5 10 is 20 26 Time (hour) 2-245

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Figure 2.5-2 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) for Shutdown Method C (LPCI) without Spurious Operation of SRV - Water Level in Hot and Average Channel PEACH BOTTOM HOT CHANEL EPU AVERAGE CHANNEL STOP OF ACTIVE F "L APPENDIX-R 60.

40.

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0.

0.8 1.2 t.8 2.010' wifti i*

TIME (SECOND) 2-246

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Figure 2.5-3 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) for Shutdown Method C (LPCI) without Spurious Operation of SRV-Reactor Vessel Pressure PEACH BOTTOM VESSEL PRESSURE EPU APPENDIX-R 0.8 U.SC-Ld

0.

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0.8 1.2 1.8

2.

-10"

,1 Fr.

TIME (SECONO]

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Figure 2.5-4 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) for Shutdown Method C (LPCI) without Spurious Operation of SRV-Peak Cladding Temperature PEACH BOTTOM PEAK CLAD TEMPE 1VR EPU APPENDIX-R 5.

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0.8 1.2 A.8 2.0103 20110W! "i,*

TIME (SECOND) 2-248

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Figure 2.5-5 Appendix R Evaluation Results EPU Fuel Heatup (SAFER) Shutdown Method C (LPCI) without Spurious Operation of SRV - Water Level Outside the Shroud PEACH BOTTOM

& REGION S EPU REGION 7 DO(N3*OhER LEVEL APPEPCIX-R TOP OF ACTIVE FUL 60.

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2. 0108 W101 TIME (SECOND]

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Figure 2.5-6 Appendix R Evaluation Results EPU Containment (SHEX) for Shutdown Method A (RCIC) without Spurious Operation of SRV - Suppression Pool Temperature PBAPS EPU App R Case Al Suppression Pool Temperature 250 200 150 E

9100/

50 0

5 10 15 20 25 Tlime (hour) 2-250

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Figure 2.5-7 Appendix R Evaluation Results EPU Containment (SHEX) for Shutdown Method A (RCIC) without Spurious Operation of SRV - RPV pressure PBAPS EPU App R Case Al RPV Pressure 14200 1200 8000 S600 400 200 0.o M.

10 15 20 25 Time (hour) 2-251

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Figure 2.5-8a Reactive Capability Curve - Unit 2 ESTIMATED REACTIVE CAPABILITY CURVES 4 Pole 1800 RPM 1530000 kVA 2 00 Volts.0.920 PF 0.510 SCR 75.00 PSIG 82 Pressure 605 Vohts Excitation 46 Deg. C Cold Gas 165 Ft Atitude Base Capability 150 75 PIG 2 0.3o0 100 06 z

0 oC,

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?OPIGH 0..9 II J520FRO'4-2a 0

w i

1 0 50 100 150 2000 MEGAWATTS 2-252

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Figure 2.5-8b Reactive Capability Curve - Unit 3 ESTIMATED REACTIVE CAPABIUTY CURVES 4 Pole 1800 RPM 1530000 WA 22000 Volts 0.900 PF 0.540 SCR 75.00 P$1G F12 Pressure 705 Volts Excitation 46Deg.CColdGas 165Ft Atitude 10,-.......

IBase CapabU4t 15000 PS-10 000 ----------

-0.0--------.5w 0

J529FR03-2a

-1000 0

50w 1000 150D 2=oo MEGAWATTS 2-253

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.6 Containment Review Considerations 2.6.1 Primary Containment Functional Design Regulatory Evaluation The containment encloses the reactor system and is the final barrier against the release of significant amounts of radioactive fission products in the event of an accident. The review for the primary containment functional design covered: (1) the temperature and pressure conditions in the DW and WW due to a spectrum of postulated LOCAs; (2) suppression pool dynamic effects during a LOCA or following the actuation of one or more RCS safety/relief valves; (3) the consequences of a LOCA occurring within the containment; (4) the capability of the containment to withstand the effects of steam bypassing the suppression pool; (5) the suppression pool temperature limit during RCS safety/relief valve operation; and (6) the analytical models used for containment analysis. The regulatory acceptance criteria for the primary containment functional design are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and that such SSCs be protected against dynamic effects; (2) GDC-16, insofar as it requires that reactor containment be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment; (3) GDC-50, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any LOCA; (4)

GDC-13, insofar as it requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation and for accident conditions, as appropriate, to assure adequate safety; and (5) GDC-64, insofar as it requires that means be provided to monitor the reactor containment atmosphere for radioactivity that may be released from normal operations and from postulated accidents.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967.

The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

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Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, with the exception of current GDC-13, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H:

Draft GDC-10, Draft GDC-12, Draft GDC-13, Draft GDC-16, Draft GDC-17, Draft GDC-40, Draft GDC-42, Draft GDC-49, Draft GDC-50, Draft GDC-54, and Draft GDC-56.

Current GDC-64 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 ODCM License Amendments 102 and 104 (Reference 35), respectively.

The primary containment is described in PBAPS UFSAR Sections 4, "Reactor Coolant System,"

5.2, "Primary Containment," 7.3, "Primary Containment and Reactor Vessel Isolation Control System," 14.6, "Analysis of Design Basis Accidents," Appendix C, "Structural Design Criteria,"

and Appendix M, "Containment Report."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the primary containment is documented in NUREG-1769, Section 2.3.2.3. Management of aging effects on the primary containment is documented in NUREG-1769, Section 3.2.3.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs.

Section 4.1 of the CLTR addresses the effect of CPPU on Primary Containment Functional Design. The results of this evaluation are described below.

The PBAPS UFSAR provides the containment responses to various postulated accidents that validate the design basis for the containment. EPU operation changes some of the conditions for the containment analyses. For example, the Short-Term DBLOCA containment response during the blowdown is governed by the blowdown flow rate. This blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel pressure and the mass and energy of the vessel fluid inventory, which change slightly with EPU. Also, the long-term heat-up of the suppression pool following a LOCA or a transient is governed by the ability of the RHR to remove decay heat. Because the decay heat depends on the initial reactor power level, the long-term containment response is affected by EPU. The containment response was reanalyzed to demonstrate the plant's capability to operate with a rated power increase to 3951 MWt. The key plant parameters used to model and analyze the plant response at EPU are provided in Table 2.6-2.

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The analyses of containment pressure and temperature responses, as described in Section 2.6.1.1, were performed at a power level of 102% of EPU RTP in accordance with ELTRI using GEH codes and models. The M3CPT code was used to model the short-term containment pressure and temperature response.

The modeling used in the M3CPT analyses is described in References 65 and 66. References 65 and 66 describe the basic containment analytical models used in GEH codes. Reference 67 describes the more detailed RPV model (LAMB) used for determining the vessel break flow in the containment analyses for EPU.

The LAMB code models the recirculation loop as a separate pressure node. It also allows for inclusion of flashing in the pipe and vessel during the blowdown and flow choking at the jet pump nozzles when the conditions warrant. The use of the LAMB blowdown flow in M3CPT was identified in ELTRI by reference to the LAMB code qualification in Reference 67.

The SHEX code was used to model the long-term containment pressure and temperature response. The key models in SHEX are based on models described in Reference 66. The GEH containment analysis methodologies have been applied to all BWR power uprate projects performed by GEH and accepted by the NRC.

Original long-term containment analyses did not credit passive heat sinks in the DW, WW airspace, and suppression pool. This conservative assumption was identified to the NRC as Assumption 6 of Attachment 1 to the March 12, 1993 GE letter referenced in Reference 8.

Long-term containment analyses performed for PBAPS EPU now includes credit for these passive heat sinks. This is herein identified as a change in methodology. These long-term containment analyses continue to conservatively neglect any heat loss from the containment through the containment walls to the reactor building or environs (Assumption 8 of the same GE letter).

The effects of EPU on the containment dynamic loads due to a LOCA or SRV discharge have also been evaluated as described in Section 2.6.1.2. The containment hydrodynamic loads have been defined generically for Mark I plants as part of the Mark I Containment Long-Term Program (LTP) (Reference 68) and approved by the NRC in Reference 69. The PBAPS plant-specific dynamic loads were defined in References 70 and 71, using the NRC approved methods of Reference 68. The evaluation of the LOCA containment dynamic loads is based primarily on the results of the short-term analysis described in Section 2.6.1.2.

The SRV discharge load evaluation is based on no changes in the SRV opening setpoints for EPU.

The metal-water reaction energy versus time relationship is calculated using the method described in USNRC RG 1.7 (Reference 72) as a normalized value (fraction of reactor thermal power). All of the energy from the metal-water reaction is assumed transferred to the reactor coolant in the first 120 seconds into the LOCA. The metal-water reaction energy represents a very small fraction of the total shutdown energy transferred to the coolant.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

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Pool Temperature Response

((

Meets CLTR Disposition Wetwell Pressure Meets CLTR Disposition Drywell Temperature Meets CLTR Disposition Meets CLTR Drywell Pressure Disposition Containment Dynamic Loads Meets CLTR Disposition Containment Isolation Meets CLTR Disposition Motor-Operated Valves Meets CLTR Disposition Hardened Wetwell Vent System Meets CLTR Disposition Meets CLTR Equipment Operability

))'

Disposition 2.6.1.1 Containment Pressure and Temperature Response The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. As a result of this, the pool temperature response, WW pressure, DW temperature, and DW pressure need to be addressed.

Short-term and long-term containment analysis results are reported in the UFSAR. The short-term analysis is directed primarily at determining the DW pressure response during the initial blowdown of the reactor vessel inventory to the containment following a large break inside the DW.

The long-term analyses are directed primarily at the suppression pool temperature response, considering the decay heat addition to the suppression pool. The DBLOCA, SSLB LOCAs, and loss of RHR normal shutdown cooling (NSDC) function event were all reanalyzed for EPU.

Peak values of the containment pressure and temperature responses to the DBLOCA are given in Table 2.6-1.

The impact of local suppression pool temperatures during SRV discharges was 2-257

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) addressed in accordance with the NUREG-0783 (Reference 73) criteria. Peak suppression pool (torus) temperatures resulting from the postulated ATWS, SBO, and 10 CFR 50, Appendix R Fire events are given in Table 2.6-3.

The effect of EPU on the events which yield the limiting containment pressure and temperature response is provided below.

2.6.1.1.1 Long-Term Suppression Pool Temperature Response 2.6.1.1.1.1 Bulk Pool Temperature The long-term bulk pool temperature response for EPU is evaluated for the limiting DBLOCA in Section 14.6.3.3 (Case C') of the UFSAR. This DBLOCA is an instantaneous guillotine break of the recirculation loop suction line (RSLB). For the PBAPS EPU, small break LOCAs were also analyzed at EPU conditions.

The GE Safety communication SC 06-01 (Reference 74) identified a potential issue.

The potential issue is that a single failure that eliminated only the RHR heat exchanger could prove more limiting than the typically analyzed scenario of the single failure of an entire AC electrical power source. The PBAPS RHR system is configured with 2 loops of RHR, with each loop having its own separate injection point to the RPV, and with each loop having its own separate return to the suppression pool. Each loop is comprised of 2 RHR pumps, each pump having its own separate suction from the suppression pool, and with each pump having its own separate heat exchanger on its discharge.

This configuration is such that GE Safety Communication SC 06-01 has been determined to not be applicable to PBAPS.

The acceptability of ECCS pump NPSH based on the containment analysis suppression po01 temperature response is demonstrated in Section 2.6.5.2.

RHR and CS pump flows can be throttled to decrease required NPSH from that required at pump run-out flow conditions, provided containment cooling requirements are still satisfied.

See Section 2.6.5.1 for the bulk pool temperature response analyses and results at EPU RTP.

2.6.1.1.1.2 Local Pool Temperature with SRVDischarge The local pool temperature limit for SRV discharge was specified in NUREG-0783 (Reference 73) because of concerns resulting from unstable condensation observed at high pool temperatures in plants without quenchers. Quencher devices such as the T-quenchers used in PBAPS mitigate these loads. The peak local suppression pool temperature at PBAPS has been evaluated for EPU, with the same scenario assumptions as evaluated in the original analysis of Reference 75, and meets the NUREG-0783 criteria. This evaluation demonstrated a minimum subcooling of approximately 20'F locally at the quencher. This meets a criteria included in NUREG-0783 to ensure that the exiting quencher steam is condensed before posing a steam ingestion potential for any ECCS pump suction. Therefore, the peak local suppression pool temperature at PBAPS remains acceptable at EPU conditions.

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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.6.1.1.1.3 Drywell to Wetwell Steam Bypass Leakage The acceptance criterion for Mark I plants such as PBAPS, with regard to steam bypass leakage, is that the measured leakage is not greater than the leakage that would result from a one-inch diameter opening. This maximum bypass leakage is confirmed by plant tests as directed in the PBAPS TS (TS surveillance requirement SR 3.6.1.1.2).

As a result of ((

)), all pressure and temperature responses were evaluated and have been found to be within limits. Therefore; the containment pressure and temperature response meets all CLTR dispositions.

2.6.1.2 Containment Dynamic Loads The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU.

As a result, containment dynamic loads are addressed in the following sections.

2.6.1.2.1 Loss-of-Coolant Accident Loads The LOCA containment dynamic loads analysis for EPU is primarily based on the short-term RSLB LOCA analyses and compliance with generic criteria developed through testing programs.

The analyses were performed as described in Section 2.6.1.1 with break flows calculated using a more detailed RPV model (Reference 67). The NRC approved use of this model for the EPU containment evaluations in Reference 2. These analyses also provide calculated values for the controlling parameters for the dynamic loads throughout the blowdown. The key parameters are DW and WW pressures, vent flow rates and suppression pool temperature. The LOCA dynamic loads considered in the EPU evaluations include pool swell, CO and chugging. For Mark I plants like PBAPS, the vent thrust loads were also evaluated.

The results of the EPU pool swell evaluation confirmed that the current pool swell load definition remains bounding. The containment response conditions for EPU are within the range of test conditions used to define CO loads for the plant. The containment response conditions for EPU are within the conditions used to define the chugging loads. The vent thrust loads at EPU conditions were calculated to be less than the plant-specific values calculated during the Mark I Containment LTP for all but four locations. For these four locations, the vent thrust load exceeds the plant-specific value calculated during the Mark I Containment LTP by no greater than 2.5%,

while the minimum margin to allowable stress limits was never less than 13%. The increase in vent thrust loads, therefore, cannot reduce the margin to stress allowable limits below 10.5%.

The LOCA dynamic loads at EPU conditions are therefore still within the allowable stress limits.

The Mark I Containment Program Load Definition Report (LDR) Table 4.5.1-1 (Reference 68) defines the onset and duration times for chugging based on break size.

For the small and intermediate break sizes, chugging lasts for a duration of 900 seconds. Chugging starts five seconds after the break for the IBA event and 300 seconds after the break for the SBA event.

Discussion of the chugging duration time is provided in Sections 2.2, 2.3, and 4.4.1.1 of the LDR. For the load definition, chugging is assumed to end when reactor pressure is reduced to or 2-259

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) below the DW pressure, essentially stopping break flow and therefore vent steam flow. This vessel depressurization for the IBA and SBA events is due to manual initiation of ADS. The load definition of the LDR does not include any credit for operation of containment (drywell) sprays.

However, emergency operating procedures (EOPs) for PBAPS include direction to initiate DW sprays prior to WW pressure exceeding 9.0 psig. Containment analyses performed for PBAPS EPU have shown that WW pressure will exceed this DW spray initiation pressure of 9.0 psig before 900 seconds following initiation of the event. Initiation of DW sprays, will rapidly reduce DW pressure and stop chugging. Therefore, containment analyses performed for PBAPS EPU confirm that the chugging duration times used in the original PBAPS load combinations remain applicable and bounding* for operation at EPU conditions. The basis for these chugging duration times is, however, changed from manual operation of ADS to operation of DW sprays.

2.6.1.2.2 Safety Relief Valve Loads The SRV loads include SRVDL loads, suppression pool boundary pressure loads, and drag loads on submerged structures.

The SRV opening setpoint pressure, the initial water leg in the SRVDL, the SRVDL geometry, and the suppression pool geometry influence these loads. The SRV loads were evaluated for two different actuation phases; initial actuation and subsequent actuation.

For the initial SRV actuation following an event involving RPV pressurization, the SRV flow rate controlling the SRV loads is dependent upon the SRV capacity and SRV opening setpoint pressure, which are not changed for EPU. EPU reduces the time between subsequent SRV actuations, however the EPU analysis confirms that the reflood height used in the original

.PBAPS analysis remains bounding.

The SRV opening setpoint pressure values, which are in part the basis for the SRVDL loads and the SRV loads on the suppression pool boundary and submerged structures, are not changed.

The effect of EPU on the load definition for subsequent SRV actuations was evaluated. Using the same methodology and assumptions as identified in the original analysis of Reference 71, the load definition for subsequent SRV actuations remains bounding and applicable for operation at EPU conditions.

2.6.1.2.3 LOCA Pressure and Temperature Loads The Reference 70 load definition report (PULD) provided LOCA-induced pressure and temperature results from DBLOCA, intermediate break accident (IBA), and small break accident (SBA) events as an input for subsequent use in the Reference 71 structural analysis. The IBA and SBA events were re-evaluated at 102% EPU RTP using initial conditions and assumptions consistent with the Reference 70 analysis. The results of the PBAPS EPU analysis show that all DW and WW pressure and temperatures at EPU conditions are bounded by the values *of Reference 70 with the exception of the peak WW and SP temperature for the SBA. At EPU conditions, the SBA peak WW and SP temperature is 148'F, which does not bound the Reference 70 result of 122°F. The current PBAPS Mark 1 Long Term Program Plant Unique 2-260

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Analysis (Reference 71) determines that the maximum internal pressure and thermal loads for the torus occur during an IBA. The design load combinations with thermal loads are all based on the bounding IBA thermal loads. The Reference 70 result for the IBA is 155'F which bounds the SBA peak WW and SP temperature of 148'F cited above. The evaluation of WW and SP piping that have the SBA temperature as a structural load combination input is contained in Section 2.2.2.2.2.2 (Other Piping Evaluation).

Given the fact that current containment dynamic load evaluations remain bounding and applicable for plant operation at EPU conditions, and that the current SRV load definition is still applicable, all CLTR dispositions are met.

2.6.1.3 Containment Isolation The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU.

However, the system designs for containment isolation are not affected by EPU.

The capabilities of isolation actuation devices to perform during normal operations and under post-accident conditions have been determined to be acceptable.

Therefore, the PBAPS containment isolation capabilities are not adversely affected by the EPU and all CLTR dispositions are met.

The capabilities of MOVs within the scope of GL 89-10 and GL 96-05 to perform during design basis conditions have been evaluated. In most cases, the values of existing parameters bound the values expected under EPU conditions. In a few cases, there are slight increases above the current value. In those cases, the effect of the slight increase on the MOV has been evaluated in accordance with the requirements of the station MOV program. Any changes necessary will be completed prior to EPU implementation.

2.6.1.4 Generic Letter 89-16 GL 89-16 recommended the installation of a hardened vent capable of relieving the pressure from the WW air space to replace the non-pressure bearing vent path(s) previously credited for severe accident management. In response to GL 89-16, PBAPS agreed to install a hardened vent from the Torus vapor space to the atmosphere in accordance with industry criteria. Using BWR Owners' Group design criteria, the vent was sized to exhaust sufficient steam to prevent the containment pressure from exceeding the primary containment pressure limit of 60 psig with a constant heat input equal to 1% of 3458 MWt. However, because the hardened WW vent was conservatively designed, it will continue to provide adequate pressure protection with a constant heat input equal to 1% of RTP under EPU conditions (3951 MWt) also. Therefore, the existing PBAPS response to GL 89-16 remains valid for EPU, and all CLTR dispositions are met.

2.6.1.5 Generic Letter 96-06 GL 96-06 identified potential problems with equipment operability and containment integrity during DBA conditions as a result of (1) water hammer and/or two-phase flow conditions in cooling water systems serving the containment air coolers and (2) thermally induced overpressurization of isolated piping sections in containment.

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The PBAPS response to GL 96-06 credited the maintenance of sufficient overpressure in the cooling water lines with preventing steam from forming during the design-basis scenarios of interest. Under EPU conditions, sufficient overpressure in the cooling water lines is maintained, thereby preventing water hammer under DBA conditions.

The PBAPS response to GL 96-06 also included the installation of relief valves on lines penetrating primary containment that were susceptible to thermally induced over-pressurization during DBA conditions. The relief valve sizing was based on a DW temperature significantly higher than that expected under EPU conditions. In addition, the relief valve installed capacity was much greater than the required capacity. As a result, the slight increase in DW temperature with EPU does not affect the adequacy of the previous corrective action.

Therefore, the existing PBAPS response to GL 96-06 remains valid for EPU, and all CLTR dispositions are met.

Conclusion The containment temperature and pressure transient has been reviewed and was found to adequately account for the increase of mass and energy resulting from the proposed EPU. The review also demonstrated that containment systems will continue to provide sufficient pressure and temperature mitigation capability to ensure that containment integrity is maintained. Exelon concludes that containment systems and instrumentation will continue to be adequate for monitoring containment parameters and release of radioactivity during normal and accident conditions and will continue to meet the requirements of the current licensing basis following implementation of the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to primary containment functional design.

2.6.2 Subcompartment Analyses Regulatory Evaluation A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high-energy piping and would limit the flow of fluid to the main containment volume in the event of a postulated pipe rupture within the volume. The review for subcompartment analyses covered the determination of the design differential pressure values for containment subcompartments. The review focused on the effects of the increase in mass and energy release into the containment due to operation at EPU conditions, and the resulting increase in pressurization. The regulatory acceptance criteria for subcompartment analyses are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and that such SSCs be protected against dynamic effects; and (2) GDC-50, insofar as it requires that containment subcompartments be designed with sufficient margin to prevent fracture of the structure due to the calculated pressure differential conditions across the walls of the subcompartments.

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Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-40, Draft GDC-42, and Draft GDC-49.

The primary containment is described in PBAPS UFSAR Sections 5.2, "Primary Containment,"

12.2.1, "Reactor Buildings," 14.6, "Analysis of Design Basis Accidents," Appendix C, "Structural Design Criteria," and Appendix M, "Containment Report."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the primary containment is documented in NUREG-1769, Section 2.3.2.3 and 2.4.2. Management of aging effects on the primary containment is documented in NUREG-1769, Sections 3.2.3, 3.5.1 and 3.5.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 10.1 of the CLTR addresses the effect of CPPU on Subcompartment Analyses. The results of this evaluation are described below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

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(CLTR PBAPS TpcDisposition Resutip Meets CLTR Liquid lines Disposition As stated in Section 10.1 of the CLTR, EPU may increase subcooling in the reactor vessel, which may lead to increased break flow rates for liquid line breaks.

An annular structure of reinforced concrete is located inside the DW around the RPV in order to provide thermal and radiation shielding, and is called the Sacrificial Shield Wall (SSW). The SSW is designed to withstand the differential pressure that would develop across the Wall as a result of a high pressure pipe break within the annulus (i.e., between the RPV and the SSW).

The calculations to determine the maximum pressure difference on SSW due to the limiting recirculation suction line break between the RPV and the SSW, were updated to treat the break flow as subcooled liquid. This conservative change in the methodology reduced the design margin more than the effect of the EPU. The results of the updated conditions including the effects of the EPU and the limiting off-rated condition along the MELLLA operating domain upper boundary (minimum recirculation pump speed (MPS) point with FFWTR) indicate that the design limit of SSW pressure difference is not exceeded.

Critical Mass Flux (ibm/sec-ft2) 8000.0 (Saturated) 9603.0 (Subcooled) 9786.1 (Subcooled) 13304.3 (Subcooled)

N/A Maximum SSW 34.0 42.4 43.2 59.2 72.0 DP (psid)

This evaluation also included consideration of the jet force impinging on the shield plugs'in the shield wall, penetrations and the stagnation pressure in the annulus resulting from the FWLB pressurizing the shield annulus. With consideration of steam flashing for AP and the jet pressure (impingement on shield plug), the results of the updated conditions including the effects of the EPU and the off-rated conditions indicate that the design limit for shield plug pressure difference is not exceeded.

Subcompartment Pressurization Evaluation As discussed earlier, the differential pressure loading on the SSW and shield plug is not significantly affected by the EPU. The peak SSW pressure load resulting -from the.limiting 2-264

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) recirculation suction line break at CLTP and at EPU conditions remains below the SSW design differential pressure. The peak shield plug pressure load resulting from the limiting FWLB at CLTP and at EPU conditions remains below the shield plug design differential pressure.

Therefore, the subcompartment analyses meet all CLTR dispositions.

Conclusion The change in predicted pressurization resulting from the increased mass and energy release following the proposed EPU has been reviewed.

It was found that the containment SSCs important to safety will continue to be protected from the dynamic effects resulting from pipe breaks and that the subcompartments will continue to have sufficient margins to prevent fracture of the structure due to pressure difference across the walls following implementation of the proposed EPU. Exelon concludes that the plant will continue to meet the current licensing basis for the proposed EPU. Therefore, Exelon finds the proposed EPU acceptable with respect to subcompartment analyses.

2.6.3 Mass and Energy Release 2.6.3.1 Mass and Energy Release Analysis for Postulated Loss of Coolant Regulatory Evaluation The release of high-energy fluid into containment from pipe breaks could challenge the structural integrity of the containment, including subcompartments and systems within the containment.

The review covered the energy sources that are available for release to the containment and the mass and energy release rate calculations for the initial blowdown phase of the accident. The regulatory acceptance criteria for mass and energy release analyses for postulated LOCAs are based on: (1) GDC-50, insofar as it requires that sufficient conservatism be provided in the mass and energy release analysis to assure that containment design margin is maintained; and (2) 10 CFR 50, Appendix K, insofar as it identifies sources of energy during a LOCA.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967.

The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

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Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-49.,

The mass and energy release analysis is described in PBAPS UFSAR Sections 5, "Containment,"

14.6.3.3, "Primary Containment Response," Appendix C, "Structural Design Criteria," and Appendix M, "Containment Report."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also

,referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs.

Section 4.1 of the CLTR addresses the effect of CPPU on Containment System Performance. The results of this evaluation are described below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

CLTR PBAPS tropie Disposition Result Drywell Temperature Meets CLTR Dislpositionl Drywell Pressure

]

Disposition 2.6.3.1.1 Drywell Temperature The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. The assumption of constant pressure minimizes the effect on other aspects of the containment evaluation. The bounding DW temperature occurs during a break of a steam line. A spectrum of steam line break sizes have been evaluated to ensure a bounding DW EQ temperature profile is established (Figure 2.6-10).

The analysis has been performed in accordance with NUREG-0588 (Reference 76), and the most limiting DW temperature from this analysis is shown in Table 2.6-1. The DW shell has a design temperature limit of 281'F, also shown in Table 2.6-1. Although the DW environment may see temperatures as high as 340'F, the most limiting temperature for the DW shell has been analyzed to be 281 'F. Therefore, the most limiting DW shell temperature remains bounded by the design limit. The effects of the dual unit interaction were analyzed and the DW, WW and containment shell temperature response remains bounded by the results reported above.

The WW gas space peak temperature response was calculated assuming a heat and mass transfer model between the pool and WW gas space that is calculated mechanistically.

Table 2.6-1 2-266

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2.6.3.1.2 Short-Term Containment Pressure Response The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. The assumption of constant pressure minimizes the effect on other aspects of the containment evaluation.

The short-term containment response analysis was performed for the limiting DBLOCA that assumes a double-ended guillotine break of a recirculation suction line to demonstrate that EPU does not result in exceeding the containment design limits. The short-term analysis covers the blowdown period during which the maximum DW pressure and WW pressure occur. The analysis was performed at 102% of EPU RTP level.

The time-dependent results of the limiting short-term analysis are presented in Figures 2.6-4 through 2.6-9 and are summarized in Table 2.6-1. Table 2.6-1 also includes comparisons of the pressure values calculated for EPU to the design pressures and to pressure values from previous calculations based on the current power. The maximum calculated containment pressure for EPU remains within the design value, and thus, is acceptable and applicable CLTR dispositions are met.

The short-term analysis was performed at EPU conditions for three different initial containment conditions. The Design case (D) considers the most limiting initial containment conditions of 70'F in the DW and 2.5 psig in the DW and WW. The Bounding case (B) considers initial containment conditions of 125°F (two standard deviations below the PBAPS average temperature) in the DW and 2.0 psig (two standard deviations above the PBAPS average pressure) in the DW and WW, bounding normal operation.

A Reference case (R) is also evaluated that assumes initial conditions of 145°F in the DW and 0.75 psig in the DW and WW -

initial conditions used in the PBAPS power rerate analysis (Reference 64). The Design case (D) was also performed at CLTP conditions to provide comparison for evaluating the impact of operation at EPU conditions.

Conclusion The mass and energy release has been reviewed and found to adequately address the effects of the proposed EPU and appropriately accounts for the sources of energy identified in 10 CFR 50, Appendix K. Based on this, Exelon finds that the mass and energy release analysis meets the requirements in the current licensing basis for ensuring that the analysis is conservative.

Therefore, Exelon finds the proposed EPU acceptable with respect to mass and energy release for a postulated LOCA.

2.6.4 Combustible Gas Control in Containment Regulatory Evaluation Following a LOCA, hydrogen and oxygen may accumulate inside the containment due to chemical reactions between the fuel rod cladding and steam, corrosion of aluminum and other 2-267

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The review covered: (1) the production and accumulation of combustible gases; (2) the capability to prevent high concentrations of combustible gases in local areas; (3) the capability to monitor combustible gas concentrations; and (4) the capability to reduce combustible gas concentrations.

The review primarily focused on any effect that the proposed EPU may have on hydrogen release assumptions, and how increases in hydrogen release are mitigated. The regulatory acceptance criteria for combustible gas control in containment are based on: (1) 10 CFR 50.44, insofar as it requires that plants be provided with the capability for controlling combustible gas concentrations in the containment atmosphere; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; (3) GDC-41, insofar as it requires that systems be provided to control the concentration of hydrogen or oxygen that may be released into the reactor containment following postulated accidents to ensure that containment integrity is maintained; (4) GDC-42, insofar as it requires that systems required by GDC-41 be designed to permit appropriate periodic inspection; and (5) GDC-43, insofar as it requires that systems required by GDC-41 be designed to permit appropriate periodic testing.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-49.

Combustible Gas Control is described in PBAPS UFSAR Sections 5, "Containment," 14.6.3.3, "Primary Containment Response," Appendix C, "Structural Design Criteria," and Appendix M, "Containment Report."

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Technical Evaluation The revised 10 CFR 50.44 (68 FR 54123, dated September 16, 2003) does not define a design basis LOCA hydrogen release and eliminates the requirements for hydrogen control systems to mitigate such releases. PBAPS License Amendment Numbers 256 and 259 for Units 2 and 3 respectively, issued in 2005 (Reference 77),

eliminated the requirements for the hydrogen/oxygen monitors and PBAPS License Amendment Numbers 274 and 278 for Units 2 and 3 respectively, issued in 2010 (Reference 78), eliminated the requirements for the CAD system.

Conclusion The containment combustible gas control system was reviewed and it was found that the effects of the proposed EPU have been adequately addressed.

The system will continue to have sufficient capability following the implementation of the proposed EPU. Exelon concludes that the containment combustible gas control system will continue to meet the requirements of the current licensing basis, as well as 10 CFR 50.44. Therefore, Exelon finds the proposed EPU acceptable with respect to combustible gas control in containment.

2.6.5 Containment Heat Removal Regulatory Evaluation Fan cooler systems, spray systems, and RHR systems are provided to remove heat from the containment atmosphere and from the water in the containment WW. The review in this area focused on: (1) the effects of the proposed EPU on the analyses of the available NPSH to the containment heat removal system pumps; and (2) the analyses of the heat removal capabilities of the spray water system and the fan cooler heat exchangers. The regulatory acceptance criteria for containment heat removal are based on GDC-38, insofar as it requires that a containment heat removal system be provided, and that its function shall be to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that 2-269

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Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-41 and Draft GDC-52.

The containment heat removal systems are described in PBAPS UFSAR Sections 4.8, "Residual Heat Removal System," 5.2.3.7, "Primary Containment and Ventilation System," 5.2.4.3, "Primary Containment Characteristics During Reactor Blowdown," and 6, "Core Standby Cooling Systems."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).

The license renewal evaluation associated with the containment heat removal system is documented in NUREG-1769, Section 2.3.2.5.

The management of the effects of aging on the containment heat removal system is documented in NUREG-1769, Section 3.2.5.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Sections 4.1 and 4.2 of the CLTR address the effect of CPPU on Containment Heat Removal. The results of this evaluation are described below:

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

CMeet CLTRS Topic Disposition

~

Result Pool Temperature Response Disposition E

Meets CLTR ECCS Net Positive Suction Head

]Disposition 2.6.5.1 Pool Temperature Response Section 4.2.6 of the CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. The long-term bulk pool temperature response for EPU is evaluated for the limiting RSLB DBLOCA in Section 14.6.3.3 (Case C') of the UFSAR.

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For the PBAPS EPU, small pipe breaks were also analyzed at EPU conditions. The most severe, highest, peak bulk suppression pool temperature for all LOCA break types and sizes is provided in Table 2.6-1.

In cases where the event does not result in rapid reactor depressurization, operators will manage reactor depressurization in accordance with existing procedures. Guidance will be provided for the operators to anticipate the rise in torus temperature resulting from an interruption in torus cooling caused by receiving a LOCA signal when a unit is depressurized below 450 psig. In addition, entry into the ASDC mode of RHR will be managed to account for the approximate 10°F rise in torus temperature. This allows control of reactor pressure while considering torus temperature and the effect on NPSH. The long term scenarios described below include this consideration and delayed reactor depressurization and entry into the ASDC mode of RHR such that NPSH is maintained.

Pool Temperature Response - RSLB DBLOCA The analysis of the RSLB DBLOCA was performed at 102% of EPU RTP. The calculated SP and WW temperature responses are presented in Figure 2.6-1, the DW and WW temperature responses are presented in Figure 2.6-2, and the peak values for LOCA bulk pool temperature for the CLTP and the EPU RTP case are compared in Table 2.6-1. The EPU analysis was performed using a decay heat table based on ANS/ANSI 5.1-1979 with 2-sigma adders with additional actinides and activation products per GE SIL 636 (Reference 63). No modifications were made to this standard. The analysis assumed the worst-case single active failure and concurrent loss of off-site power. For the first 10 minutes of the DBLOCA, there is no operator action. As such, CS (one loop with two pumps) and RHR (one loop with two pumps and another loop with only one pump) automatically align to provide cooling to the fuel by either spray (CS) or flooding (LPCI). With a LOOP and a single active failure (SAF) of one EDG, a third CS pump in the other loop is expected to initiate as well. However, operators are assumed to stop this pump when they take manual control. Ten minutes after the event initiation, the operators take manual control and align one loop of RHR (one RHR pump, one RHR heat exchanger and one HPSW pump providing cooling flow to the RHR heat exchanger) to provide containment cooling with a flow rate of 8600 gpm.

All three modes of containment cooling were evaluated, SPC, containment spray cooling (CSC), and coolant injection cooling (CIC). The other two RHR pumps are manually turned off, and CS flow maintains core cooling. At one hour after event initiation, the RHR heat exchanger cross-tie is placed in service (see EPPU LAR Attachment 9 for description of the RHR heat exchanger cross-tie modification), which results in a containment cooling configuration of 1 RHR pump with flow split between each of the two RHR heat exchangers (total RHR flow of 8600 gpm). The RHR heat exchanger cross-tie modification results in a significant increase in the capacity of a single RHR pump to cool the containment and suppression pool. Table 2.6.1 also provides results for CLTP benchmarked with the RHR heat exchange performance used for the EPU evaluation. This benchmarking provides a direct comparison of the effect of the higher decay heat at EPU conditions. PBAPS analysis uses a RHR heat exchanger performance that is supported by RHR heat exchanger performance testing.

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PBAPS ensures that the actual heat removal capability remains above the required or analyzed value through heat exchanger thermal performance testing. The performance testing acceptance criteria will be revised based on the EPU analyses to reflect the minimum acceptable heat removal capability that is greater than the value assumed in the analyses.

The resulting calculated peak bulk suppression pool temperature at EPU RTP for the DBLOCA is 186'F.

The preceding paragraph description of the DBLOCA analysis performed at 102% EPU RTP treats the two PBAPS units singularly, i.e. the analysis does not consider the interaction of the two units whereas an assumed LOOP on both PBAPS units concurrent with both a DBLOCA on one of the PBAPS units and a worst-case single failure (loss of one EDG) results in both Units 2 and 3 sharing the remaining three EDGs which could impact the SP temperature response on the DBLOCA unit.

As stated in Appendix F of the PBAPS UFSAR, design consideration has been given to the possible effects of interaction of Units 2 and 3 on the diesel generators. For the postulated case of DBA conditions for one unit, including the total LOOP, it has been recognized that a loss of DW cooling to the non-accident unit could eventually result in HDWP for that unit. The ECCS logic at PBAPS requires a HDWP/LRPVP to initiate LPCI injection. Therefore the HDWP will result in the generation of a LOCA signal only when the non-accident unit is depressurized below 450 psig. This design feature allows the operator to control the non-accident unit and take preemptive actions as described below so that automatic actuations do not result in overloading of the shared EDGs.

The non-accident unit LOCA signal will result in the start and alignment of the ECCS pumps to the LPCI mode of operation. As stated in the PBAPS UFSAR, Appendix F, in order to prevent overloading of the EDGs the RHR pumps are electrically interlocked between units at the breaker level to preclude the operation of the RHR pumps of both units on one diesel generator.

This logic could result in tripping RHR pumps on the accident unit; therefore operators are instructed to secure RHR and HPSW pumps on the accident unit prior to depressurizing the non-accident unit below 500 psig in Abnormal Operating Procedures (AOPs). After the non-accident unit is depressurized below 450 psig and the ECCS pumps start, adequate core cooling is verified, unnecessary pumps are secured and containment cooling is restored on the accident unit.

The analysis of the above scenario, herein called "dual unit interaction", for the RSLB DBLOCA was performed as described above for the single unit RSLB DBLOCA except containment cooling is assumed interrupted due to dual unit interaction when the accident unit suppression pool temperature is 100F below the peak suppression pool temperature that would be experienced by the accident unit if there was no containment cooling interruption due to dual unit interaction. The depressurization of the non-accident unit is controlled and the operators are instructed to verify a margin of 10'F exists, on the accident unit, to the NPSH curves in the EOPs prior to depressurizing the non-accident unit below 450 psig so the effect on the accident unit is acceptable and doesn't result in loss of NPSH for the ECCS pumps.

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The accident unit containment cooling interruption is assumed for a period of 10-minutes due to the dual unit interaction. After the 10-minute interruption, containment cooling on the accident unit is restored with the same containment cooling configuration as existed prior to the dual unit interaction interruption. The resulting calculated peak bulk suppression pool temperature at EPU RTP for the RSLB DBLOCA with dual unit interaction is 187.2 0F.

Evaluation of the containment response for the non-accident unit was also evaluated. With the conservative assumption that the RSLB DBLOCA occurs concurrently with a LOOP, reactor isolation and scram will occur on the non-accident unit. With the conservative assumption that the SAF during the RSLB DBLOCA is loss of an EDG, insufficient electrical capacity will exist to place in service two HPSW pumps and thereby initiate the RHR heat exchanger cross-tie in the non-accident unit. In addition, loss of a specific EDG will result in the inability to place in service the NSDC mode of the RHR system in order to achieve cold shutdown of the non-accident unit.

The capability of the non-accident unit to achieve cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> was analyzed at 102% of EPU RTP and ANS/ANSI 5.1-1979 with 2-sigma adders with additional actinides and activation products per GE SIL 636 (Reference 63). This analysis includes the assumption of reactor shutdown initiated by a LOOP (for both PBAPS units) with concurrent loss of one division power (for the accident/non-accident unit). The loss of one division of power prevents the use of the NSDC mode of the RHR system. HPCI is conservatively assumed available and provides reactor inventory makeup until SP temperature reaches 180'F. If SP temperature reaches 180'F, HPCI is secured because HPCI availability cannot be assured with SP temperature greater than 180'F. No credit is assumed for CST volume. DW cooling fans in the non-accident unit are assumed as unavailable. When non accident unit SP temperature reaches 11 0°F, but no sooner than ten minutes after reactor shutdown; the operators commence manual reactor depressurization and reactor cooldown at a rate of 100'F/hr. At one-hour after reactor shutdown, the operators align one loop of RHR (one RHR pump, one RHR heat exchanger and one HPSW pump providing cooling flow to the RHR heat exchanger) in SPC mode with a flow rate of 8600 gpm. At one-hour following the start of reactor depressurization, it is assumed that a LOCA signal (on the non-accident unit) may occur on HDWP commensurate with low reactor pressure. This assumed timing for the LOCA signal on the non-accident unit is based on a depressurization rate of 100lF/hr. This LOCA signal will stop all HPSW flow, start all available low-pressure ECCS pumps, and realign the RHR pumps to LPCI mode in the non-accident unit.

The analysis assumes this operation continues for 10 minutes, after which operators are assumed to stop all but one RHR pump (if SP temperature is below 180'F, HPCI will also be available),

restart the HPSW flow to one RHR heat exchanger in one RHR loop, and reestablish SPC. A second interruption of containment cooling on the non-accident unit is assumed due to dual unit interaction, as described above, when the non-accident unit suppression pool temperature is 10°F below the peak suppression pool temperature that would be experienced by the non-accident unit if there was no containment cooling interruption due to dual unit interaction. Note that this assumption of a second containment cooling interruption is conservative for the large pipe break DBLOCA because, for the DBLOCA, the LOCA signal on the accident unit will occur within 2-273

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) the first minute of the accident progression. This second containment cooling interruption for the non-accident unit was assumed in order to provide a bounding analysis of the non-accident unit for all events in which the RHR heat exchanger cross-tie modification is not available.

The non-accident unit containment cooling interruption is assumed for a period of 10-minutes due to the dual unit interaction. After the 10-minute interruption, containment cooling on the non-accident unit is restored with the same containment cooling configuration as existed prior to the dual unit interaction interruption.

When RPV pressure reaches 150 psig, the analysis assumes that the operators will maintain the RPV at this pressure.

Containment cooling is maintained using RHR SPC mode. The containment spray mode of RHR may be initiated to maintain the containment shell temperature less than the containment design temperature of 281'F. When bulk suppression pool temperature is below a pre-determined value (180'F) and decreasing, operators are assumed to flood the RPV to the level of the MSLs, open the ADS valves to completely depressurize the RPV, and initiate ASDC in the CIC mode. Cooldown of the RPV to cold shutdown conditions on the non-accident unit is accomplished with ASDC.

Cold shutdown is achieved when bulk reactor liquid water temperature is below 200'F, which is conservative for PBAPS with a TS definition for Cold Shutdown of 212'F. The peak bulk SPC temperature for this analysis at EPU conditions is 203.8'F, and the time to achieve cold shutdown was 35.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The resulting time-dependent bulk suppression pool temperature response is presented in Figure 2.6-1.b.

The results of this evaluation of the second PBAPS (non-accident) unit response is applicable and bounding for small break LOCAs and other accidents/events on the other PBAPS unit because the scenario includes a bounding dual unit interaction and uses the minimum equipment available to reach cold shutdown.

Pool Temperature Response -Loss of RHR Normal Shutdown Cooling Function Event Analyses were also performed to confirm the ability of PBAPS to reach cold shutdown conditions within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, considering the inability to cool down the reactor using the NSDC mode of the RHR system. This method to achieve cold shutdown is termed ASDC. ASDC can be established in various ways. Because the method of CIC mode results in the greatest heat addition to the suppression pool, only CIC mode was evaluated in this analysis. This method establishes ASDC using a single RHR pump aligned in CIC mode to both flood the vessel and to provide containment cooling, which results in the most limiting peak bulk suppression pool temperature.

The suppression pool temperature response for the analysis of the loss of normal RHR shutdown cooling function event with cold shutdown achieved by ASDC also represents the bounding bulk suppression pool temperature response for a small liquid line break LOCA wherein SPC mode is used in lieu of CSC mode.

The analysis results provide a peak bulk suppression pool temperature for the small liquid line break LOCA of 186°F when the accident unit was treated singularly and 186.7'F when additional containment cooling interruption is considered due to dual unit interaction.

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The capability of the ASDC method to achieve cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> was analyzed at 102% of EPU RTP and ANS/ANSI 5.1-1979 with 2-sigma adders with additional actinides and activation products per GE SIL 636 (Reference 63). Reactor shutdown is assumed initiated by a LOOP with concurrent loss of one division power. The loss of one division power prevents the use of the NSDC mode of the RHR system. HPCI is conservatively assumed available and provides reactor inventory makeup until reactor pressure decreases below the HPCI isolation pressure, after which low-pressure ECCS provides reactor inventory makeup. If HPCI is not available, ADS would be used to rapidly reduce reactor pressure to allow low-pressure ECCS to provide vessel makeup. Such use of ADS results in a faster heatup of the suppression pool.

With reactor pressure at the time of peak pool temperature the same, the total (integrated) sensible heat addition to the suppression pool remains the same, but the total (integrated) decay heat to the pool at the time of peak pool temperature is less for the fast pool heatup. In addition, the heat removed from the pool is greater for the faster pool heatup. Thus, a faster pool heatup results in a lower peak pool temperature. For this reason, the assumption of crediting the HPCI as available is conservative. No credit is assumed for CST volume.

When suppression pool temperature reaches 1 10°F, but no sooner than ten minutes following initiation of the event, the operators commence manual reactor depressurization and reactor cooldown at a rate of 100'F/hr. At ten minutes after initiation of the event, the operators align one loop of RHR (one RHR pump, one RHR heat exchanger and one HPSW pump providing cooling flow to the RHR heat exchanger) in SPC mode with a flow rate of 8600 gpm. At one hour after event initiation, the RHR heat exchanger cross-tie is placed in service, wherein the RHR flow from the single RHR pump is split, with half of the pump discharge directed to a second RHR heat exchanger.

When RPV pressure reaches the NSDC RPV pressure permissive of 70 psig, the analysis assumes that the operators will maintain the RPV at this pressure.

Containment cooling is maintained using SPC mode.

Just before suppression pool temperature peaks (ten minutes is assumed in the analysis), it is conservatively assumed that a LOCA signal may occur on HDWP commensurate with low reactor pressure. This LOCA signal will stop all HPSW flow, start all available low-pressure ECCS pumps, and realign the RHR pumps to LPCI mode. The analysis assumes this operation continues for 10 minutes, after which operators are assumed to stop the additional low-pressure ECCS pumps (two RHR and all CS pumps), restart the HPSW flow to both RHR heat exchangers in the remaining one RHR loop, and reestablish SPC.

When bulk suppression pool temperature is below a pre-determined value (160'F) and decreasing, operators are assumed to choose to attempt NSDC. The analysis assumes operators take 30 minutes to attempt NSDC, after which time they determine NSDC cannot be accomplished.

Another five minutes is conservatively assumed to realign and establish containment cooling using either SPC or CIC mode. Because the CIC mode results in a more limiting suppression pool temperature response, this mode was assumed in the analysis.

Operators then initiate action to establish ASDC by opening the ADS valves and allowing 2-275

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) reactor vessel water level to increase above the elevation of the MSL nozzles, flooding the MSLs, and flowing out the ADS valves (liquid flow) and returning to the suppression pool. Cold shutdown is achieved when bulk reactor liquid water temperature is below 200'F, which is conservative for PBAPS with a TS definition for Cold Shutdown of 212'F. The peak bulk SPC temperature for this analysis at EPU conditions is 186°F, and the time to achieve cold shutdown was 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.

Accomplishing ASDC using either SPC or CSC modes would be expected to take slightly longer to achieve cold shutdown.

However, the time to achieve cold shutdown assuming CIC is significantly earlier than the acceptance limit time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Therefore additional analysis runs using either SPC or CSC modes were considered unnecessary.

The resulting time-dependent bulk suppression pool temperature response is presented in Figure 2.6-3 and the peak bulk suppression pool temperature at 102% of EPU RTP is 1867F.

The preceding paragraph description of the loss of RHR NSDC function event performed at 102% EPU RTP treats the two PBAPS units singularly, i.e. the analysis does not consider the interaction of the two units whereas an assumed LOOP on both PBAPS units concurrent with the loss of one division power can result in multiple containment cooling interruptions, which could impact the SP temperature response on the PBAPS unit experiencing the loss of RHR NSDC function event. For the discussion that follows, the PBAPS unit which is evaluated for the SP temperature response is herein called the "Loss of NSDC Event" unit. The other PBAPS unit is herein called the "Second PBAPS" unit.

The above scenario, herein called "dual unit interaction", to determine the capability of the ASDC method to achieve cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> was analyzed as described above for the single unit event, except as follows. At one-hour following the start of reactor depressurization, it is assumed that a LOCA signal (on the Loss of NSDC Event unit) may occur on HDWP commensurate with low reactor pressure. This assumed timing for the LOCA signal on the Loss of NSDC Event unit is based on a depressurization rate of 100°F/hr. This LOCA signal will stop all HPSW flow, start all available low-pressure ECCS pumps, and realign the RHR pumps to LPCI mode in the Loss of NSDC Event unit. The analysis assumes this operation continues for 10 minutes, after which operators are assumed to stop the additional low-pressure ECCS pumps (two RHR and all CS pumps), restart the HPSW flow to both RHR heat exchangers in the remaining one RHR loop, and reestablish SPC or CSC as applicable. A second interruption of containment cooling on the Loss of NSDC Event unit is assumed due to dual unit interaction when the Loss of NSDC Event unit suppression pool temperature is 10°F below the peak suppression pool temperature that would be experienced by the Loss of NSDC Event unit if there was no containment cooling interruption due to dual unit interaction. The depressurization of the Second PBAPS unit is controlled and the operators are instructed to verify a margin of 10°F exists, on the Loss of NSDC unit, to the NPSH curves in the EOPs prior to depressurizing the Second PBAPS unit below 450 psig so the effect on the Loss of NSDC unit is acceptable and doesn't result in loss of NPSH for the ECCS pumps.

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The Loss of NSDC Event unit containment cooling interruption is assumed for a period of ten minutes due to the dual unit interaction. After the ten minute interruption, containment cooling on the Loss of NSDC Event unit is restored with the same containment cooling configuration as existed prior to the dual unit interaction interruption. When RPV pressure reaches the NSDC RPV pressure permissive of 70 psig, the analysis assumes that the operators will maintain the RPV at this pressure. Containment cooling is maintained using SPC mode.

When bulk suppression pool temperature is below a pre-determined value (160'F) and decreasing, operators are assumed to choose to attempt NSDC. The analysis assumes operators take 30 minutes to attempt NSDC, after which time they determine NSDC cannot be accomplished. Another 5 minutes is conservatively assumed to realign and establish containment cooling using either SPC or CIC mode. Because the CIC mode results in a more limiting suppression pool temperature response, this mode was assumed in the analysis. Operators then initiate action to establish ASDC by opening the ADS valves and allowing reactor vessel water level to increase above the elevation of the MSL nozzles, flooding the MSLs, and flowing out the ADS valves (liquid flow) and returning to the suppression pool. Cold shutdown is achieved when bulk reactor liquid water temperature is below 2007F, which is conservative for PBAPS with a TS definition for Cold Shutdown of 212'F. The peak bulk SPC temperature for this analysis at EPU conditions is 186.77F, and the time to achieve cold shutdown was 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. Accomplishing ASDC using either SPC or CSC modes would be expected to take slightly longer to achieve cold shutdown.

However, the time to achieve cold shutdown assuming CIC is significantly earlier than the acceptance limit time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Therefore additional analysis runs using either SPC or CSC modes are unnecessary.

The resulting time-dependent bulk suppression pool temperature response is presented in Figure 2.6-3a.

Pool Temperature Response -Small Steam Break LOCA For a large DBLOCA, sufficient flow from the break can be established to provide long-term core cooling to achieve cold shutdown. For a small break LOCA, such as the SSLB, the ASDC method may be required to achieve cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following initiation of the event. Therefore, the SSLB analyses performed for EQ profiles in the DW are also used to evaluate suppression pool temperature for small LOCA events, and to confirm the ability to achieve cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following initiation of the event.

For this small LOCA analysis, a spectrum of SSLBs is evaluated. Initial reactor conditions are consistent with operation at 102% of EPU RTP, and the same decay heat, relaxation and metal-water reaction energies are assumed as is used for the large DBLOCA analysis. Consistent with the large DBLOCA assumptions, a complete LOOP and a worst-case single failure are also assumed for this analysis. Reactor vessel inventory makeup is provided by one loop of CS (two pumps) and three RHR pumps in LPCI mode. For smaller steam breaks, HPCI may be available for vessel makeup, but is not credited for 10 minutes to ensure a bounding DW temperature is evaluated.

At 10 minutes, operators turn off two of the RHR pumps, and align the remaining RHR pump to provide containment cooling with a flow of 8600 gpm through one RHR heat exchanger, and one 2-277

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HPSW pump providing cooling flow to the RHR heat exchanger. When DW pressure exceeds 2.0 psig, operators initiate WW spray, with the remainder of the RHR flow remaining in SPC mode. Prior to WW pressure exceeding 9.0 psig, operators initiate DW spray and stop all SPC flow. When bulk suppression pool temperature exceeds 11 0IF (but not earlier than 10 minutes following initiation of the event), operators initiate a controlled reactor vessel cooldown at 1007F per hour. At one hour from initiation of the event, operators establish the RHR heat exchanger cross-tie to the other RHR heat exchanger in the same loop, such that a total RHR flow rate of 8600 gpm is maintained to the DW and WW spray headers. When reactor pressure is decreased below the pressure permissive for NSDC (70 psig), operators maintain DW and WW spray cooling with the RHR heat exchanger cross-tie in service, and maintain reactor vessel pressure as low as possible to limit steam flow from the break. When bulk suppression pool temperature is below a pre-determined value (170'F for EPU), operators open one or more ADS valves and increase vessel water level in the reactor vessel to the MSL nozzles using the one loop of CS until water flows from the open ADS valves back to the suppression pool, establishing ASDC.

DW and WW sprays continue to be used to provide containment and SPC, bulk reactor water temperature decreases below 200'F and cold shutdown is achieved prior to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> from initiation of the event, which is conservative for plants with cold shutdown defined at a higher temperature.

The resulting time-dependent bulk suppression pool temperature response is presented in Figure 2.6-11 and the peak bulk suppression pool temperature at 102% of EPU RTP is 187°F.

The preceding two paragraphs description of the SSLB analysis performed at 102% EPU RTP treats the two PBAPS units singularly, i.e. the analysis does not consider the interaction of the two units whereas an assumed LOOP on both PBAPS units concurrent with both a SSLB on one of the PBAPS units and a worst-case single failure (loss of one emergency diesel generator (EDG)) results in both Units 2 and 3 sharing the remaining three EDGs which could impact the SP temperature response on the small break LOCA unit.

The analysis of the above scenario, herein called "dual unit interaction", for the SSLB was performed as described above for the single unit analysis except as noted in the following paragraphs.

Only the limiting (0.01sqft break size), with respect to peak suppression pool temperature, SSLB LOCA was analyzed for dual unit interaction.

At 10 minutes, operators turn off two of the RHR pumps and align the remaining RHR pump to provide containment cooling using with a flow of 8600 gpm through one RHR heat exchanger and one HPSW pump providing cooling flow to the RHR heat exchanger. When DW pressure exceeds 2.0 psig, operators initiate WW spray, with the remainder of the RHR flow remaining in SPC mode. Prior to WW pressure exceeding 9.0 psig, operators initiate DW spray and stop all SPC flow.

When bulk suppression pool temperature exceeds 110'F (but not earlier than 10 minutes following initiation of the event), operators initiate a controlled reactor vessel cooldown at 100'F per hour. At one hour from initiation of the event, operators establish the RHR heat exchanger cross-tie to the other RHR heat exchanger in the same loop, such that a total RHR flow rate of 8600 gpm is maintained to the DW and WW spray headers. At one-hour 2-278

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) following the start of reactor depressurization it is assumed that a LOCA signal on the SSLB LOCA unit may occur on HDWP commensurate with low reactor pressure. This timing for the HDWP/low reactor pressure LOCA signal on the small break LOCA/accident unit is based on the depressurization rate of 1 00°F/hr mentioned above.

This LOCA signal will stop all HPSW flow, start all available low-pressure ECCS pumps, and realign the RHR pumps to LPCI mode in the SSLB LOCA/accident unit. The analysis assumes this interruption in SPC continues for 10 minutes, after which operators are assumed to stop the additional low-pressure ECCS pumps (two RHR and unnecessary CS pumps), restart the HPSW flow to both RHR heat exchangers in the remaining one RHR loop, and reestablish SPC or CSC as applicable. A second interruption of containment cooling on the accident unit is assumed due to dual unit interaction when the accident unit suppression pool temperature is 10°F below the peak suppression pool temperature that would be experienced by the accident unit if there was no containment cooling interruption due to dual unit interaction. The depressurization of the non-accident unit is controlled and the operators are instructed to verify a margin of 1 0F exists, on the accident unit, to the NPSH curves in the EOPs prior to depressurizing the non-accident unit below 450 psig. This ensures the effect on the accident unit is acceptable and doesn't result in loss of NPSH for the ECCS pumps. The accident unit cooling interruption is assumed for a period of ten minutes due to the dual unit interaction. After the ten minute interruption, containment cooling on the accident unit is restored with the same containment cooling configuration as existed prior to the dual unit interaction interruption. When reactor pressure is decreased below the pressure permissive for NSDC (70 psig), operators maintain DW and WW spray cooling with the RHR heat exchanger cross-tie in service, and maintain reactor vessel pressure as low as possible to limit steam flow from the break. When bulk suppression pool temperature is below a pre-determined value (170'F for EPU), operators open one or more ADS valves and increase vessel water level in the reactor vessel to the MSL nozzles using the one loop of CS until water flows from the open ADS 'valves back to the suppression pool, establishing ASDC. DW and WW sprays continue to be used to provide containment and SPC, bulk reactor water temperature decreases below 200'F and cold shutdown is achieved prior to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> from initiation of the event, which is conservative for plants with cold shutdown defined at a higher temperature. The resulting time-dependent bulk suppression pool temperature response is presented in Figure 2.6-11 a and the peak bulk suppression pool temperature at 102%

of EPU RTP is 187.6°F 2.6.5.2 ECCS Net Positive Suction Head Section 4.2.6 of the CLTR states that EPU rated thermal power operation increases the reactor decay heat, which increases the heat addition to the suppression pool following a large break DBLOCA, SSLB LOCA, Loss of RHR NSDC event, Stuck Open Relief Valve (SORV) with RPV isolation event, Appendix R event, SBO event, and ATWS event.

Following these accidents and non-design basis events, the RHR and CS pumps operate to provide the required core and containment cooling. Adequate NPSH margin (NPSH available minus NPSH required) is required during this period to assure the essential pump operation. The NPSH margin for the 2-279

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ECCS pumps is evaluated for the limiting conditions following these accidents and non-design basis events. The limiting NPSH conditions depend on the pump flow rates, debris loading on the suction strainers (for debris generating events), pipe frictional losses, suppression pool level, and suppression pool temperature. Maximum torus pressure is assumed to be 14.638 psia. No CAP is used for calculating NPSHA. EPU calculations for ECCS pump NPSH are consistent with RG 1.82 (Reference 79) for the DBLOCA and meet the requirements of recent NRC guidance with respect to CAP (Reference 80).

CAP is credited in the current analysis basis at CLTP for PBAPS. Table 2.6-4 provides a summary of the current NPSH analysis of record for the RHR and CS pumps.

The torus water level for the NPSH analysis of each event is adjusted for the minimum drawdown level consistent with RG 1.82 (Reference 79) requirements. For the LOCA events, drawdown includes the suppression pool / RPV break flow inventory that is held up in the DW and the pressure-suppression system vent headers.

The torus water level for the RSLB DBLOCA NPSH analysis is 12.40 feet for short term operation. The torus water level for RSLB DBLOCA long-term operation, SSLB LOCA, loss of RHR NSDC event, SORV with RPV isolation, and safe shutdown of the second unit NPSH analysis is 13.94 feet.

For SBO, ATWS, and Appendix R fire event methods "A", "B", and "D" torus water level will increase from the initial nominal torus water level due to the use of either RCIC or HPCI for RPV makeup with the RCIC or HPCI inventory being supplied from the CST. Therefore, for the NPSH analysis of SBO, ATWS, and Appendix R fire event methods "A", "B", and "D",

inventory addition from the CST is credited in the torus water level calculation. The torus water level without CST credit for Appendix R fire event method "C" NPSH analysis is 13.85 feet.

The ECCS pump Net Positive Suction Head Required (NPSHR) used in all NPSH margin evaluations at EPU conditions contains a 21% uncertainty for DBAs that include a pipe break, i.e., the RSLB DBLOCA and small break LOCAs, and 0% uncertainty for other DBAs and special events.

The NPSHR including uncertainty (effective NPSHR or NPSHReff), is determined as follows:

NPSHReff = (1 + uncertainty)NPSHR3%

NPSHR3% is the ECCS pump vendors NPSH corresponding to a 3% reduction in pump head during testing (NPSHR 3%). The NPSHR~ff is used in calculation of all ECCS pump NPSH margins.

The NPSH margins were calculated assuming system flow rates that meet or exceed ECCS pump operational requirements for the analyzed accidents and transients.

A listing of the safety analysis ECCS pump flow rate and the ECCS pump flow rate used in the. corresponding NPSH evaluation are shown in Table 2.6-5. The ECCS pump flows used in the NPSH analysis, except for the RHR pump flow rate for the RSLB DBLOCA short-term operation, were increased by a factor of 1/40.97 (1.015) to account for the possible reduction of pump total developed head, 3-percent, when NPSHR3% curves are utilized for comparison to NPSHa.

For the RSLB 2-280

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DBLOCA, the RHR short-term NPSH analysis flow rate of 10,600 gpm is approximately 3%

higher than the containment safety analysis assumed short-term flow rate of 10,300 gpm.

The ECCS pumps have been analyzed for operation time within the maximum erosion zone, defined as NPSH margin ratios between 1.2 and 1.6. The margin ratio is defined by NPSHA/NPSHR 3%. If this ratio is less than 1.6, the operating time spent below 1.6 has been determined and shown in Table 2.6-5.

Consideration of ECCS suction strainer debris loading within the NPSH evaluations at EPU conditions is consistent with the PBAPS current analysis of record for the large break (RSLB)

DBLOCA event. For PBAPS EPU, the SSLB and ATWS events also include ECCS suction strainer debris loading in the NPSH evaluations for these events. Table 2.6-5 indicates how ECCS suction strainer debris loading is considered for each of the analyzed events.

Following a RSLB DBLOCA, the RHR and CS pumps operate to provide the required core and containment cooling. For short-term RSLB DBLOCA operation (less than 10 minutes), the RHR pumps are assumed to operate in the NPSH evaluation at 10600 gpm and the CS pumps are assumed to operate in the NPSH evaluation at 3973 gpm. For long-term operation (greater than 10 minutes), the RHR pumps are assumed to operate in the NPSH evaluation at 8732 gpm, and the CS pumps are assumed to operate in the NPSH evaluation at 3493 gpm. The NPSH margins for the,RHR and CS pumps were evaluated for the limiting conditions following a RSLB DBLOCA. The limiting NPSH conditions depend on the total pump flow rate, debris loading on the suction strainers, pipe frictional losses, suppression pool level, and suppression pool temperature. The maximum suppression pool temperature, NPSHA, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) are listed in Table 2.6-5. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-12a and 2.6-12b for the RHR pumps and Figures 2.6-13a and 2.6-13b for the CS pumps.

The PBAPS suppression pool temperature response was evaluated at 102% of EPU RTP for a spectrum of small steam breaks (0.01 ft2, 0.05 ft2, 0.1 ft2, 0.25 ft 2, 0.50 ft2 and 1.00 ft2). Except for the smallest of these breaks, 0.01 ft2, the suppression pool temperature response results in lower pool temperatures than for the DBLOCA. For all SSLB cases, during the first 10 minutes of the event the peak suppression pool temperature is at least 20°F lower than for the DBLOCA.

In addition, during the first 10-minutes of the SSLB LOCA, the RHR and CS pumps are either not in operation or are operating at minimum flow. Therefore, the RHR and CS NPSH margins for all events except for the long-term response of the 0.01 ft2 break case are bounded by the results of the DBLOCA analysis.

The NPSH margins for the ECCS pumps were evaluated for the limiting conditions following a 0.01 ft2 small steam break. For the SSLB event NPSH analysis, the RHR pumps are assumed to operate in the NPSH evaluation at 8732 gpm, and the CS pumps are assumed to operate in the NPSH evaluation at 3493 gpm. The limiting NPSH conditions depend on the pump flow rates, debris loading on the suction strainers, pipe frictional losses, suppression pool level and 2-281

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) suppression pool temperature. No changes to any of debris loading characteristics on the suction strainers, pipe frictional losses, and LOCA minimum suppression pool level result from the implementation of EPU.

The maximum suppression pool temperature, NPSHA and NPSH margin are listed in Table 2.6-5. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-14a and 2.6-14b.

The PBAPS loss of RHR NSDC event NPSH evaluation analyzed two modes of achieving cold shutdown with ASDC:

  • CIC where RHR provides both SPC and reactor cooling
  • SPC where RHR provides SPC and CS provides reactor cooling.

The CIC mode of ASDC results in the highest suppression pool temperature response and therefore provides a more conservative suppression pool temperature input for the NPSH evaluation of the CS pump than does the SPC mode of ASDC operation. HPCI is also assumed to operate during this event, with an assumed primary water suction source from the suppression pool, to provide RPV inventory make-up with reactor pressure above the HPCI isolation pressure.

The assumption of HPCI operation is conservative for the determination of peak suppression pool temperature. However, HPCI pump suction from the suppression pool can be limited to suppression pool temperature below 180°F, the maximum allowed temperature for HPCI operation. HPCI can be secured prior to suppression pool temperature reaching 180OF with no impact on the ability to ensure core cooling. The HPCI pump NPSH margin at 180°F suppression pool temperature is 1.0 ft. with an assumed HPCI flow rate of 5150 gpm.

For the NPSH analysis of the loss of RHR NSDC event, the RHR pumps are assumed to operate at 8732 gpm (CIC and SPC), and the CS pumps are assumed to operate at 3493 gpm (SPC). The maximum suppression pool temperature, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) are listed in Table 2.6-5. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response.

Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-15a and 2.6-15b. This NPSH analysis does not include an assumed pipe break, therefore no ECCS strainer debris loading is assumed and no uncertainty is applied in the determination of NPSHeff.

The Section 2.6.5.1 suppression pool temperature response analysis for the loss of RHR NSDC function event with cold shutdown achieved by ASDC method is also applicable for a small liquid line break LOCA wherein SPC mode is used in lieu of CSC mode. The suppression pool peak temperature response for both of these events is bounded by the suppression pool temperature response for the limiting SSLB LOCA.

Both the small liquid line break and the SSLB involve line breaks, therefore strainer debris loading and 21% uncertainty are applied to the NPSH evaluation. Because the SSLB results in a higher suppression pool temperature, the NPSHA and NPSH margin analysis for the SSLB 2-282

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LOCA will bound the NPSH analysis for the small liquid line break. The RHR pump and CS pump NPSHA and NPSH margins reported in Table 2.6-5 for the SSLB LOCA are bounding for the small liquid line break LOCA.

The suppression pool bulk temperature response due to a SORV with RPV isolation event was evaluated. For this event, RHR operates in SPC and CS operates to provide RPV coolant inventory makeup at low reactor pressure. HPCI also operates during this event to provide RPV inventory make-up with reactor pressure greater than 150 psig; however, the primary water suction source for HPCI is the CST. For the SORV event NPSH analysis, the RHR pumps are assumed to operate at 8732 gpm and the CS pumps operate at 3493 gpm. The maximum suppression pool temperature, NPSHA, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) are listed in Table 2.6-5. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-16a and 2.6-16b.

The RHR pumps operate in all Appendix R safe shutdown analysis cases (see Section 2.11.1.2.2 for detailed description). In the ECCS NPSH evaluation, a RHR flow rate of 8732 gpm is used.

Appendix R Case Al has the highest suppression pool temperature of 205.8'F. This temperature, bounds the suppression pool temperature response for Appendix R cases A2, B 1, B2, D I and D2.

Appendix R case CIB has a peak pool temperature of 204.4'F, however the assumed suppression pool level is lower for this case than for Case Al.

Therefore, this case is quantitatively evaluated for RHR pump NPSH. Based on the results of the Appendix R analysis at EPU, the suppression pool temperature response for Appendix R Case C I1B bounds the results for Appendix R Case C213.

Appendix R Case C2B is the same as Appendix R C1B with exception of an assumed SORV during the event. The RHR pump NPSH for Appendix R Case CIA is also quantitatively evaluated because, although the peak suppression pool temperature is significantly lower than either Appendix R Case Al or Appendix R Case C1B, the time of

,operation in the maximum erosion zone may be limiting for this Case CIA. Based on the results of the Appendix R analysis at EPU, the suppression pool temperature response for Appendix R Case CIA bounds the results for Appendix R Case C2A. HPCI pump operation is credited for Appendix R Methods B and D, but suction is from the CST only.

The limiting Appendix R event NPSH case for the CS pumps is Appendix R Case CIA, with an assumed flow rate of 3493 gpm. Appendix R Case C2A is the same as Appendix R CIA with exception of an assumed SORV during the event. Based on the results of the Appendix R analysis at EPU, the suppression pool temperature response for Appendix R Case CiA bounds the results for Appendix R Case C2A.

For PBAPS, all Appendix R fire event scenarios also terminate with reactor cooldown to cold shutdown conditions by ASDC method.

The maximum suppression pool temperature, NPSHA, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) for the limiting Appendix R scenarios are listed in Table 2.6-5.

For Appendix R Case Al, the suppression pool (torus) water level at the 2-283

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) maximum suppression pool temperature is 15.9 ft. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. Time-history plots of the NPSHA and NPSH margin ratios for Appendix R case Al are shown in Figures 2.6-17a and 2.6-17b. Time-history plots of the NPSHA and NPSH margin ratios for Appendix R case C1A are shown in Figures 2.6-18a and 2.6-18b. Time-history plots of the NPSHA and NPSH margin ratios for Appendix R case CIB are shown in Figures 2.6-19a and 2.6-19b.

For the SBO event, the only ECCS pump operating with suction from the suppression pool is one RHR pump. The assumed RHR pump flow for the SBO NPSH evaluation is 8732 gpm. The maximum suppression pool temperature, NPSHA, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) for the SBO scenario is listed in Table 2.6-5. The suppression pool (torus) water level at the maximum suppression pool temperature is 15.4 ft.

The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. The HPCI pumps are also credited for the SBO event, operating for a maximum of 30 minutes with suction from the CST only. Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-20a and 2.6-20b.

For the ATWS event, the only ECCS pumps operating with suction from the suppression pool are the RHR pumps. The HPCI pumps supply makeup to the RPV with alignment to the CST.

The CS pumps are not credited for the ATWS event. The assumed RHR pump flow for the ATWS event NPSH analysis is 8732 gpm.

The maximum suppression pool temperature, NPSHA, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) for the ATWS event are listed in Table 2.6-5. The suppression pool (torus) water level at the maximum suppression pool temperature is 15.8 ft. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-21a and 2.6-21b.

EPU RTP operation increases the reactor decay heat, which increases the heat addition to the suppression pool following a DBLOCA or other non-design basis events. The peak suppression pool temperature for the analyzed accidents and transients is within the design capability of the ECCS pumps. Adequate NPSHA is demonstrated, and no credit for CAP is needed during a DBA or special event.

The ECCS pump operating time in the maximum erosion zone

(<1.6 margin ratio) is much less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for any event.

The preceding descriptions of the DBLOCA, small steam breaks and loss of RHR NSDC event analyses treat the two units singularly. As stated in Section 2.6.1, an assumed LOOP concurrent with a worst case single failure (loss of one EDG) results in both Units 2 and 3 sharing the remaining three EDGs which could impact the suppression pool temperature response on the accident unit. The impact of this dual unit interaction is discussed in Section 2.6.5.1. The effects on NPSH are included in the associated figures referenced above for each of these events and in Table 2.6-5.

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The suppression pool temperature response during shutdown and cooldown of the second (non-accident) PBAPS unit during a DBLOCA (on the other PBAPS unit) concurrent with LOOP and loss of EDG is discussed in Section 2.6.5.1. This event results in the second PBAPS unit entering into ASDC in the CIC mode in order to achieve cold shutdown conditions. For the second (non-accident) unit safe shutdown NPSH analysis, the RHR pumps are assumed to operate at 8732 gpm, which is conservatively higher than the RHR pump flow rate assumed in the safety analysis that provides the suppression pool temperature response. CS is not analyzed for this event but could be operated within the limits of the EOPs. HPCI is assumed available for part of this event, however HPCI can be secured prior to the suppression pool temperature reaching the 180'F qualification limit for the HPCI pump (The RPV will be depressurized below the pressure at which RHR in LPCI mode can provide RPV makeup before the suppression pool temperature reaches the HPCI NPSH limit). Torus water level for the NPSH evaluation is 13.94 ft. (consistent with the previously discussed NPSH evaluation for the loss of RHR NSDC event with cold shutdown achieved by ASDC). Because there is no pipe break in the second (non-accident) unit, no ECCS suction strainer debris loading is assumed and no uncertainty is applied in the determination of NPSHReff. The maximum suppression pool temperature, NPSH margin, and the operating time in the maximum erosion zone (<1.6 margin ratio) are listed in Table 2.6-5.

Time-history plots of the NPSHA and NPSH margin ratios are shown in Figures 2.6-22a and 2.6-22b.

The debris generated and transported following a LOCA that can cause ECCS strainer head loss includes fiber, reflective metal insulation, qualified coating, dirt/dust, rust flakes, sludge, and unqualified coating.

The ECCS strainer design debris load, which is primarily based on the methodology in the BWR Utility Resolution Guidance for ECCS Suction Strainer Blockage (NEDO-32686), was evaluated and is not impacted by EPU. The ECCS suction strainers are manufactured by ABB.

The ECCS pumps have been analyzed for plant specific conditions and have sufficient NPSH margin to perform satisfactorily under all accident and transient conditions. Therefore, all CLTR dispositions are met for ECCS pump NPSH at EPU conditions for PBAPS.

Conclusion The containment heat removal systems were reviewed and it was found that the effects of the proposed EPU have been adequately addressed. The systems will continue to be able to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

Exelon concludes that the containment heat removal systems will continue to meet the requirements of the current licensing basis and ensure adequate NPSH margin for the ECCS pumps without credit for CAP. Therefore, Exelon finds the proposed EPU acceptable with respect to containment heat removal systems.

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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.6.6 Secondary Containment Functional Design Regulatory Evaluation The secondary containment structure and supporting systems of dual containment plants are provided to collect and process radioactive material that may leak from the primary containment following an accident.

The supporting systems maintain a negative pressure within the secondary containment and process this leakage.

The review covered: (1) analyses of the pressure and temperature response of the secondary containment following accidents within the primary and secondary containments; (2) analyses of the effects of openings in the secondary containment on the capability of the depressurization and filtration system to establish a negative pressure in a prescribed time; (3) analyses of any primary containment leakage paths that bypass the secondary containment; (4) analyses of the pressure response of the secondary containment resulting from inadvertent depressurization of the primary containment when there is vacuum relief from the secondary containment; and (5) the acceptability of the mass and energy release data used in the analysis. The review primarily focused on the effects that the proposed EPU may have on the pressure and temperature response and drawdown time of the secondary containment, and the effect this may have on offsite dose. The regulatory acceptance criteria for secondary containment functional design are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and be protected from dynamic effects (e.g., the effects of missiles, pipe whipping, and discharging fluids) that may result from equipment failures; and (2) GDC-16, insofar as it requires that reactor containment and associated systems be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967.

The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

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For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-10, Draft GDC-40, and Draft GDC-42.

The secondary containment systems are described in PBAPS UFSAR Sections 5.3, "Secondary Containment System," 12.2.1, "Reactor Buildings," Appendix A. 10, "High Energy Pipe Break Outside the Primary Containment," and Appendix C, "Structural Design Criteria."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the secondary containment systems is documented in NUREG-1769, Sections 2.3.2.8 and 2.4.2. Management of aging effects on the secondary containment systems is documented in NUREG-1769, Sections 3.2.8 and 3.5.2.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 4.5 of the CLTR addresses the effect of CPPU on the SGTS.

The SGTS is designed to maintain secondary containment at a negative pressure and to provide an elevated release path for the exhaust air for removal of fission products potentially present during abnormal conditions. By minimizing ground level release and providing for an elevated release point for the airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA. The SGTS fission product control and removal function evaluation is described in Section 2.5.2.1.

Generic bounding analyses have been performed with results located in Section 4.5 of the CLTR. The results of this evaluation are given below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

Meets CLTR Flow capacity

"[

))Disposition The CLTR states that the core inventory of iodine and subsequent loading on the SGTS filter or charcoal adsorbers are affected by EPU.

The design flow capacity of the SGTS was selected to maintain the secondary containment at the required negative pressure to minimize the potential for exfiltration of air from the reactor 2-287

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)) The PBAPS HEPA filters are satisfactory for EPU operation.

The secondary containment structure, openings, and pathways, and drawdown time are unaffected by EPU. Because the maximum dome pressure is also not changed for EPU, there is no effect to the ability of secondary containment to contain mass and energy released to it. There is no increase in mass and energy released to secondary containment for EPU. The secondary containment temperature and pressure are not evaluated further in the CLTR because there is no effect as a result of EPU. Therefore, the evaluation of the SGTS ability to maintain secondary containment at a negative pressure and contain radionuclides is adequate for this topic.

Therefore, the flow capacity meets all CLTR dispositions.

Conclusion A review of the secondary containment pressure and temperature transient has confirmed the ability of the secondary containment to provide an essentially leak-tight barrier against uncontrolled release of radioactivity to the environment. The review determined: 1) there is no significant increase of mass and energy that would result from the implementation of the proposed EPU; and 2) the secondary containment will continue to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.

Exelon concludes that the secondary containment and associated systems will continue to meet the requirements of the current licensing basis.

Therefore, Exelon finds the proposed EPU acceptable with respect to secondary containment functional design.

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Table 2.6-1 PBAPS Containment Performance Results Peak Drywell Pressure (psig) 2 7 47.8 (Reference 81) 49.5 (D) 50.4 (D) 48.7 (B) 56.0 Peak Drywell 337.9 340 3406 281 4 Temperature (0F) 3 (Reference 82)

Peak Bulk Pool 205.63 180 186 281 4 Temperature (0F) f (Reference 82)

(187.2)

Peak Wetwell 29.21 Pressure (psig) 5 (Reference 82) 32.3 32.4 56.0 Peak Wetwell 205.56 175 181 2814 Temperature (0F) f (Reference 82)

(184)

Notes:

1. Containment analyses performed for the EPU use methods that are similar to the methods used for the CLTP analyses. The analysis at CLTP with the EPU Model uses the plant inputs defined for the EPU model including improved RHR heat exchanger performance (as discussed in Section 2.6.1.1.1).
2.

Most limiting values obtained from Short-Term Analysis.

3.

Most limiting values of drywell atmosphere temperature obtained from Long-Term Steam Line Break Analysis (0.01 sq fi) performed for EQ of equipment in the drywell.

4.

Temperature limit is the design temperature for the containment vessel (shell). Maximum calculated drywell shell temperature is 281°F, which does not exceed the drywell shell design limit temperature of 281°F

5. Peak values for long-term DBLOCA analysis. A peak bulk suppression pool temperature of 186°F was calculated for the loss of RHR NSDC event for EPU, which is also applicable for a small liquid line break LOCA. A peak bulk suppression pool temperature of 1877F was calculated for the SSLB LOCA analysis performed for maximum drywell temperature for EQ, and is therefore the most limiting peak bulk suppression pool temperature for all LOCA break sizes. The peak WW temperatures were also analyzed for the various events.

DBLOCA resulted in the peak WW temperature for all events of 181°F.

Parenthetical values show results corresponding dual unit interaction event.

6.

Refer to Figure 2.6-10 for SHEX output.

7.

Two cases are reported, Design (D) and Bounding (B). The Design case assumes initial containment pressures of 2.5 psig and initial drywell temperature of 70'F. This case was performed at CLTP and EPU reactor conditions in order to provide a comparison for the impact of operation at EPU conditions. The Bounding case assumes initial containment pressures of 2.0 psig and initial drywell temperature of 125'F -

conditions corresponding to the lower bound PBAPS normal operating DW temperature and conditions corresponding to the upper bound PBAPS normal operating containment pressure.

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Table 2.6-2 Long-Term Containment Response Key Analysis Input Values 1

Reactor A

Initial power level 3,527.16 1

102% current rated power MWt 2

102% uprated power MWt 4,030 B

Normal FWT at 102%

uprated power (102%

of OF 384 3951MWt)

C Initial vessel dome pressure 1 At 102% current rated power psia 1,068 2 At 102% uprated power psia 1,068 D

Decay heat model 1 Short-term DBLOCA ANS 5 1971 +

20%

1979 ANS 5.1 2 Long-term+2 E

Vessel volumes 1 Total vessel free volume cuft 20,682 2 Vessel liquid volume cuft 11,790 3 Vessel attached piping volume cuft 4,142 F

Vessel related masses (used in long-term calculation) 1 Liquid mass in recirculation loops Ibm 51,292 2

Steam mass in all MSLs between the RPV nozzle and lbm 3,216 second MSIV 3 Liquid mass in the HPCI piping between the RPV nozzle Ibm 349 and first normally closed valve 4 Liquid mass in the RCIC piping between the RPV nozzle Ibm 67 and first normally closed valve 5 Liquid mass in the LPCI piping from the vessel nozzle to Ibm 0

the first normally closed valve 6 Liquid mass in the RHR shutdown line the RPV and the Ibm 22,103 first normally closed valve Liquid mass in the CS piping between the RPV nozzle lbm 8,684 and the first normal closed valve.

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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 1 Time at which MSIVs start to close (RSLB and MSLB) 0.5 sec 2 Time at which MSIVs become fully closed (RSLB) sec 3.5 3 Time at which MSIVs become fully closed (MSLB) sec 3.5 2

Drywell/Vent System A

Total drywell free volume (including vent system) cuft 175,800 0.0(N)

Initial drywell pressure B

(N = miNimum, B = Bounding, X = maXimum) psig 2.0(B) 2.5(X)

Initial drywell temperature 70(N)

C 0F 125(B)

(N = miNimum, B = Bounding, X maXimum) 145(X)

Initial drywell relative humidity 20(N)

D%

(N = miNimum, X = maXimum) 100(X)

E Submergence of downcomers 1 Low water level ft 4.0 2 High water level ft 4.4 F

Loss coefficient for vent system including entrance and real 5.17 exit losses (based on vent exit flow area)

G Drywell holdup volume cuft 4,416 3

Wetwell/Suppression Pool Initial suppression pool volume (including water in vents)4 1 Low water level (LWL) cuft 122,900 2 High water level (HWL) cuft 127,300 Initial suppression pool temperature 70(N)

B

°F 86(B)

(N = miNimum, B = Bounding, X = maXimum) 95(X)

C Initial wetwell airspace free volume 44 1 LWL cuft 132,000 2 HWL cuft 127,700 2-291

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D Initial wetwell airspace pressure (N = miNimum, B = Bounding, X = maXimum) psig 0.0(N) 2.0(B) 2.5(X) 70(N)

Initial wetwell airspace temperature 0F 86(B)

(N = miNimum, B = Bounding, X = maXimum) 95(X)

Initial wetwell airspace relative humidity 100(N)

(N = miNimum, X = maXimum) 100(X)

RIHR Heat exchanger K-value (the first value is used prior to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, when the RHR heat exchanger cross-tie is NOT in BTU/sec°F 3.05 / 500 effect. The second value is used after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the RHR heat exchanger cross-tie is in effect.

Service Water Temperature OF 92 Drywell spray flow rate (1 RHR pump) gpm 7,867 Wetwell spray flow rate (1 RHR pump) gpm 733 RHR flow rate in pool cooling mode gpm 8,600 Wetwell-to-Drywell Vacuum Breakers Pressure difference between wetwell and drywell for sid 0.5 vacuum breakers to be open Number of valve systems 12 Flow area of one vacuum breaker line (per valve) sqft 1L62 Total loss coefficient per valve system including valves, real 2.78 nozzles and piping real 2-292

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Table 2.6-3 PBAPS Peak Suppression Pool (Torus) Temperatures for Postulated ATWS, Station Blackout, and 10 CFR 50 Appendix R Events Limiting ATWS (Pressure Regulator Failure Open Event) 168.3 0F Station Blackout 198.0°F Appendix R Fire 205.80F 2-293

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Table 2.6-4 PBAPS CAP Credit Summary - CLTP Analysis of Record RHR Pump DBLOCA 205.8 Yes (700 ft3) 26.0 7.03 6.1352 2.1128 10000 Small break Bounded by No 26.0 Not Calculated Not Calculated Bounded by 10000 LOCA DBLOCA DBLOCA SBO 190 No 26.0 4.629 2.1119 6.0062 10000 ATWS 190 No 26.0 4.629 2.1119 6.0062 10000 SORV 190 No 26.0 4.629 2.1119 6.0062 10000 Appendix R 206 No 26.0 6.8514 5.7563 2.6298 10000 COPR - Containment Overpressure Required MCPA - Minimum Containment Overpressure Available 2-294

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Table 2.6-4 PBAPS CAP Credit Summary - CLTP Analysis of Record (Continued)

Core Spray Pump DBLOCA 205.8 Yes (700 ft3) 26.75 7.03 4.7754 5.3779 3125 Small break Bounded by No 26.75 Not Calculated Not Calculated Bounded by 3125 LOCA DBLOCA DBLOCA SBO 190 No 26.75 4.629 0.695 9.3873 3125 ATWS 190 No 26.75 4.629 0.695 9.3873 3125

  • SORV 190 No 26.75 4.629 0.695 9.3873 3125 Appendix R 206 No 26.75 6.8514 4.3483 6.0109 3125 COPR - Containment Overpressure Required MCPA - Minimum Containment Overpressure Available 2-295

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Table 2.6-5 ECCS Pumn EPU NPSH Summarv ST DBLOCA RHR 10300 10600 159 31.97 30.25 1.72 21 0.15 No LTLT RHR 8600 "8732 186 23.21 19.36 3.85 21

<6 Yes DBLOCA LT DBLOCA (1)

RH-R 8600 8732 187.2 22.69 19.36 3.33 21

<7 Yes ST DBLOCA CS 3913 3973 159 32.91 26.62 6.29 21 0.1 No LT CS 3125 3493 186 24.98 24.20 0.78 21

<15 Yes DBLOCA LT DBLOCAT1)

CS 3125 3493 187.2 24.46 24.20 0.26 21

<15 Yes Small Steam Line Break RHR 8600 8732 187 22.78 19.36 3.42 21

<4 Yes LOCA Small Steam line Break RHR 8600 8732 187.6 22.51 19.36 3.15 21

<4 Yes LOCA ')

Small Steam Line Break CS 3125 3493 187 24.54 24.20 0.34 21

<14 Yes LOCA Small Steam line Break CS 3125 3493 187.6 24.28 24.20 0.08 21

<14 Yes LOCA (')

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Table 2.6-5 ECCS Pump EPU NPSH Summary (continued)

Loss of RHR NSDC RHR 8600 8732 186 25.74 16.00 9.74 0

0 No Loss of RHR RHR 8600 8732 186.7 25.43 16.00 9.43 0

<1 No NSDC ()

LossofRHR CS 3125 3493 186 26.00 20.00 6.00 0

<11 No NSDC LossofRHR CS 3125 3493 186.7 25.69 20.00 5.69 0

<11 No NSDC SORV R-R 8600 8732 180 28.22 16.00 12.22 0

0 No SORV CS 3125 3493 180 28.48 20.00 8.48 0

<6 No Appendix R asenAI RHR 8600 8732 205.8 17.21 16.0 1.21 0

<14 No Case Al AppendixR RHR 8600 8732 196.6 20.51 16.0 4.51 0

<20 No Case CIA AppendixR RHR 8600 8732 204.4 16.03-16.0 0.03 0

<18 No Case CIB Appendix R CS 3125 3493 196.6 20.77 20.0 0.77 0

<40 No Case CIA 2-297

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SBO RHR 8600 8732 198.0 21.30 16.0 5.30 0

<5 No ATWS RHR 8600 8732 168.3 31.54 16.0 15.54 0

0 Yes Second (non-accident)

PBAPS unit R-R 8600 8732 203.8 16.49 16.00 0.49 0

<26 No safe shutdown(1)

Note 1: This event includes the effect of the dual unit interaction discussed in PUSAR Section 2.6.5.1.

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Figure 2.6-1 EPU SP and WW Temperature Response to RSLB DBLOCA (CIC)

PBPS RSLB10 CIC - EPUMP 300 (Late) Peak SP Temperature is 186"F@ 2.89 hours0.00103 days <br />0.0247 hours <br />1.471561e-4 weeks <br />3.38645e-5 months <br /> (Late) Peak Wetwetl Temperature is 181 "F @ 3.54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> 250 200 I150s 100 50 0

0 6

12 18 24 Time (hours) 2-299

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Figure 2.6-1a 300 250 200 L.

50 0

L 0

PO 0000544 DATE: 03152012 EPU SP and WW Temperature Response to RSLB DBLOCA Dual Unit Interaction (CIC) 5 10 15 20 Time (hr) 2-300

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Figure 2.6-lb EPU SP Temperature Response of Non-Accident Unit during Safe Shutdown 300 250 E

ISO 0r 11200 CL 50 PBAPS Non-Accident Unit Safe Shutdown

]Peak Buk SP A mperatureIs 203.8 Feb.69 hours7.986111e-4 days <br />0.0192 hours <br />1.140873e-4 weeks <br />2.62545e-5 months <br /> I

I

-~

4

. /11ý 1'

24 30 0

6 12 Time (hours) 36 2-301

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Figure 2.6-2 EPU DW and WW Airspace Temperature Response to DBLOCA (CIC)

PBPS RSLB10 aC - EPUMP 00005216 0109~11 350 300 250 S200

[ISO 100 50 0o.

0.A 1.0 10.0 100.0 1,000.0 10,000.0 Log lle (Seconds) 100,000.0 Note: This figure is generated from the calculations designed to maximize containment pressures and SP temperature.

This figure shows a short duration DW temperature excursion within the first 20 to approximately 50 seconds of the LOCA.

This temperature excursion is attributed to the break flow modeling within the SHEX code which assumes steam-only break flow when the water level in the vessel falls below the break elevation and the break is uncovered. This modeling drives the DW to a non-physical superheated condition which lasts for approximately 30 seconds, until the water level recovers to the break elevation and liquid break flow is again injected to the DW. It is expected that for the DBLOCA RSLB, a two-phase break flow would exist, exhibiting a different DW temperature response than shown for this short time period of this figure.

The M3CPT code has a two-phase break flow model. With more robust, two-phase, modeling of the break flow, the DW temperature during this time following a DBLOCA is accurately represented by the temperature response shown in Figure 2.6-7 or Figure 2.6-9, obtained from the M3CPT/LAMB short-term DBLOCA analyses.

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Figure 2.6-2a EPU DW and WW Airspace Temperature Response to RSLB DBLOCA - Dual Unit Interaction (CIC)

PBPS RSLBU O ac - EPUMP 00o05S44 03/15/12 350 250 t.200 E is 50

'0 0.1 1.0 10.0 100,0 1,000.0 10,000.0 100,000.0 Time (seconds)

Note:

This figure is generated from the calculations designed to maximize containment pressures and SP temperature. This figure shows a short duration DW temperature excursion within the first 20 to approximately 50 seconds of the LOCA. This temperature excursion is attributed to the break flow modeling within the SHEX code which assumes steam-only break flow when the water level in the vessel falls below the break elevation and the break is uncovered. This modeling drives the DW to a non-physical superheated condition which lasts for approximately 30 seconds, until the water level recovers to the break elevation and liquid break flow is again injected to the DW. It is expected that for the DBLOCA RSLB, a two-phase break flow would exist, exhibiting a different DW temperature response than shown for this short time period of this figure. The M3CPT code has a two-phase break flow model. With more robust, two-phase, modeling of the break flow, the DW temperature during this time following a DBLOCA is accurately represented by the temperature response shown in Figure 2.6-7 or Figure 2.6-9, obtained from the M3CPT/LAMB short-term DBLOCA analyses.

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EPU Suppression Pool Temperature Response to Loss of Normal RHR Shutdown Cooling Event (CIC)

Figure 2.6-3 300 250 200 150 100 a.

50 0

PBPS ASDCIO MSIV CLOSURE WITH CIC AT EPU POWER - EPUMP Peak Bulk SP T mperature is 186'° @ 3.43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> 0

6 12 18 Time (hours) 24 30 36 2-304

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Figure 2.6-3a EPU Suppression Pool Temperature Response to Loss of Normal RIIR Shutdown Cooling Event - Dual Unit Interaction (CIC)

PBAPS ASDC10 MSIV CLOSURE WITH CIC AT EPU POWER - EPUMP 300 250 Peak Bulk S Tempetwure Is 18f.7? '

3.15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> IL E

C I.-

32 200 150 100 so 0

6 12 18 rime (hours) 24 30 36 2-305

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Figure 2.6-4 EPU Short-Term RSLB DBLOCA Containment Pressure Response PBAPS EPU CONT PRESS RESP RSL810 (102P/IOOF)

RSL8

'ORYIELL PRESSURE WEN~TELL PRESSU~RE 60.

i10.

1111111 I

LIi Ix 1Lii 20.

A I 0.

WfRv 2020lM 081910 199.0 10.

TIME 20.

SECONDS 30.

qO.

Pdw and Pww - Design Case 2-306

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Figure 2.6-5 EPU Short-Term RSLB DBLOCA Containment Temperature Response PBAPS EPU CONT TEtf RESP R&MBL (02P/100F)

RSLB I

RY.E.LL TEMPERAT W T1.EL TEMPERAT RE RE 350.

250.

1i..

cn uJ LU 0:

LU 150.

50.

0.

KwY 2020!769 (81910 lw.o

10.

20.

TIME SECONDS 30.

40.

Tdw and Tww - Design Case 2-307

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EPU Short-Term RSLB DBLOCA Containment Pressure Response Figure 2.6-6 PBAPS EPU CONT PRESS RESP RSLBIO (102P/100F)

I RYW.LL PRESStRE W

.TW-LL PRESSLIM RSLB 60.

90.

I I i L0 0-Ur) cx) 20.

0 1

0.

W.RV 2MIw 021%1 ASA15L

10.

20.

TIME SECONDS 3i0.

40.

Pdw and Pww - Bounding Case 2-308

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Figure 2.6-7 EPU Short-Term RSLB DBLOCA Containment Temperature Response PBAPS EPU.

CONT TEI'I RESP RSLBIO (i02P/100F) 1 OY1.LL TEMPERATVRE

.WWLL TEMPIERATIJRE RSLB 350.

250.

Li..

LU LJ 0::

C)

LUJ M

LU I

I lilt I

III 1s0.

50.

HUNY 2S 021sti A1 96125&r

10.

20.

TIME SECONDS 30.

  • 0.

Tdw and Tww - Bounding Case 2-309

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Figure 2.6-8 EPU Short-Term RSLB DBLOCA Containment Pressure Response Pdw and Pww - Reference Case 2-310

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Figure 2.6-9 EPU Short-Term RSLB DBLOCA Containment Temperature Response PRAPS LPU CONT TEMP RESP RSLBIO (102P/100r)

ODRYELL TEMPERAT JRE 2

WET-ELL TEMPERATURE RSLB 350.

250.

Li.

LUi LU C)

LU

D LUj 0-I j

I 2

2 2

m I

I ISO.

  • A 0.

IERV M9IMSI 02tst PIPCi7

10.

20.

TIME SECONDS 30.

o0.

Tdw and Tww - Reference Case 2-311

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Figure 2.6-10 EPU Long-Term Small Steam Break LOCA Drywell Temperature Response PBAPS Drywell Ter Environmental 0 A.

A,

A_ý I "

Time (wond~s 2-312

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Figure 2.6-11 EPU Long-Term Small Steam Break LOCA Suppression Pool Temperature Response PBPS EQDWI6 0.01 SQFT SSLB-EPUMP 300 Peak Bulk SP Temperature Is 187"F@ 3.51 hours5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> 250 200

.0

ýs 150 100 50 0

6 12 18 24 30 36 Time (hours) 2-313

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Figure 2.6-11a EPU Long-Term Small Steam Break LOCA Suppression Pool Temperature Response - Dual Unit Interaction 300

__250 E

i9150 50 0

0 P000"MlOA1T #3)x@

PCCSIMS WE1 03UNO1 5

10 15 20 25 30 35 Time (hr) 2-314

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Figure 2.6-12a DBLOCA RHR Pump NPSH vs. Time RHR Pump NPSH - DBA-LOCA Short Term 42.00 40.00 3&00 36.00 34.00~

32.00 NPSHA 30.00.

2800 0.00 100.00 200D.00 300.00 Time (seconds) 400.00 50M0O.

600.00 RHR Pump NPSH - DBA-LOCA Long Term z

35 0

0.

23.00 29.00

+/-.

19.00 4-

. ~... NPSHA 25.00.

+/-

23 0

i.

0.00 10000.00 20000.00 30000.00 40000.00 50000.00 60000.00 70000.00 80000.00 90000.00 Time (seconds) 2-315

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Figure 2.6-12a DBLOCA RHR Pump NPSH vs. Time (continued)

RHR Pump NPSH - DBA-LOCA Long Term Dual Unit Interaction 35.00 33.00 31t00 29.00 NPSHA 21.00 19.00-17.00-600 10,600 20,600 30,600 40,600 50,600 60,600 70,600 80,600 Time (seconds) 2-316

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Figure 2.6-12b DBA-LOCA RHR Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - DBA-LOCA Short Term 1.70 1.60 1.50 1.40 1.30

~--------------------

1.20 1.10.

1.00.

0.00 100.00 200.00 300.00 Time (seonds) 400.00 500.00 600.00 RHR Pump NPSH Margin Ratio - DBA-LOCA Long Term 2.20 T

2.00 1.80 U 60 4 140r 1.20 100 4

0 10,000 i I-----

4 4 4

20,000 30,000 40,000 50,000 60,000 70,000 lime (seconds) 80,000 90,000 2-317

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Figure 2.6-12b (continued)

DBLOCA RHR Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - DBA-LOCA Long Term Dual Unit Interaction 2.20 2.00 I

1.60 1.40 1.00 -

600 10,6(00 20,6W 30,600 40,600 50,600 60,600 70,600 80,600 lime (seconds 2-318

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Figure 2.6-13a DBLOCA CS Pump NPSH vs. Time CS Pump NPSH - DBA-LOCA Short Term 45.00 40.00

ýNP HMA 35.00 30.00 UPS*4ReF 25.00 20.00 0.00 100.00 200.00 300.00 400.00 500.00 600.00 Time (seconds)

CS Pump NPSH - DBA-LOCA Long Term 3&00 36.00 34.00 32.00 30.00 2&00 2&00 24.00 22.00 -

9-.

NPSHA

-t p

-- -+

t_..

+

+r......

p...

10000 20000 30000 40000 50000 60000 70000 80000 Time (seconds) 90DOO 2-319

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Figure 2.6-13a DBLOCA CS Pump NPSH vs. Time (continued)

CS Pump NPSH - DBA-LOCA Long Term Dual Unit Interaction 3&00 g

2 2

36.00 34.00 32.00 3000 2800 z6&00 24.00 22.00 601 NPISHRUT 10601 20601 30601 40601 50601 lime (seconds) 60601 70601 80601 2-320

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Figure 2.6-13b DBLOCA CS Pump NPSH Margin Ratio CS Pump NPSH Margin Ratio - DBA-LOCA Short Term 200 1.90 1.70 1.60 - -----------

1.50 1.40 z

1.3D 1.10 1.00

0.

10.00 200,00 300.00 400.00 500.00 600 Tb- (secnds

'.00 CS Pump NPSH Margin Ratio - DBA-LOCA Long Term 1.80 1.60 4------.--,-.*

1.70

,t

,*rr?**

1.50*

I largin Rat Z 1.40 S1.30 1.20 ~

1.1 0 1.00.

0 10000 20000 30000 40000 50000 60000 70000 80000 90000 lime (seconds) 2-321

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Figure 2.6-13b DBLOCA CS Pump NPSH Margin Ratio (continued)

CS Pump NPSH Margin Ratio - DBA-LOCA Long Term Dual Unit interaction 1.80 1.70 1.10 6001 10sw 206W 306M 406W 50600 60600 706W 8060D 71me Isecofds) 2-322

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Figure 2.6-14a Small Steam Line Break LOCA RHR and CS Pump NPSH vs. Time RHR Pump NPSH - Small Steam Line Break Long Term "600 41.00-36.00

!31.00 26.00 21.00 I-16.00 4--

0.00 20000.00 40000.00 60000.00 80000.00 1000O0.00 Time (seconds) 120000.00 CS Pump NPSH - Small Steam Une Break Long Term 4700 42.00 37.00 32.00 NPSHA 27.00.

NPSHR",

22.00 0

20000 40000 60000 80000 100000 120000 Time (seconds) 2-323

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Figure 2.6-14a Small Steam Line Break LOCA RHR and CS Pump NPSH vs. Time (continued)

CS Pump NPSH - Small Steam Une Break Dual Unit Interaction 47.00 4 2.0 0

+.

3200 27.00..

22.00 6,500 26,500 46,500 60,500 86,500 106,500 126,500 Tlmo fsecofids) 2-324

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Figure 2.6-14b Small Steam Line Break LOCA RHR and CS Pump Margin Ratio F

RHR Pump NPSH Margin Ratio - Small Steam Line Break 2.60 2.40 2.20 Margn Rai~o 2.00 1.40 1.00

+

600 26,600 40,60 60,600 80,600 100,600 120,600 Time (seconds)

CS Pump NPSH Margin Ratio - Small Steam Une Break 2.20 200 1.80 1,60-------------------------------------------------------------------------

z 1,40 I.+41 1.00 120600 600 20600 40600 60600 80600 100600 120600 line (seconds) 2-325

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Figure 2.6-14b Small Steam Line Break LOCA RHR and CS Pump Margin Ratio (continued)

RHR Pump NPSH Margin Ratio - Small Steam Une Break Dual Unit Interaction 2.60 2.40 22 i0 Z160 1.00 L..

1.20 0

26,5. 0 4

6-Wo 26,50D 46,5o 66,50D 86,500 106-900 126,500 Time (seconds)

CS Pump NPSH Margin Ratio - Small Steam Line Break Dual Unit Interaction 2.20 2.004 Margin Ratio 1.20 1.00 6500 26500 46500 66500 86500 106500 126500 Time (seconds) 2-326

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Figure 2.6-15a Loss of RHR NSDC Event -

RHR and CS Pump NPSH vs. Time RHR Pump NPSH - ASDC 44.00 3900 3400 g

Z 2900 z

NPSIA 24.00 19.00 14.00 0.00

NPSHRm, 20000.00 40000.00 60000.00 Tim (seconds) 80000.00 100000.00 120000.00 CS Pump NPSH - ASDC 43.00 38A00 NPSHA 33.00 2L&00 23.00
NPSHfw, 1&800 0

60000 Thu. (Kecond)

NOW looooo 120000 2-327

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Figure 2.6-15a Loss of RHR NSDC Event -

RHR and CS Pump NPSH vs. Time (continued)

RHR Pump NPSH - ASDC Dual Unit Interaction 44.00 39.00 34.00 NPSIA 29.00 24.00 19.00 14.00 6410 NPSMRtff 20.600 40.600 60600 Time (seconds) 80,600 1001600 120,600 CS Pump NPSH - ASDC Dual Unit Interaction 43.00 3&00 33.00 2M00 NPSM 23.00 NPSHR~,W 13O00 60O 404600 6060 lime (Seconds) 8$0600 120AW0 2-328

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Figure 2.6-15b Loss of RHR NSDC Event -

RHR and CS Pump Margin Ratio RHR Pump NPSH Margin Ratio - ASDC 2.80 2.60 2.40 2.20 2.00 1.80 1s.

1.60 1.40 1.20 1.0 00 0

20,000 40,000 640,O Tim. I(euxds) 80,000 100,000 120,000 CS Pump NPSH Margin Ratio - ASDC 2.2D 2.00

'J 1.60 lAG 120-......................................................................................

0 2000o Thn (Sewnds) 100000 1200 2-329

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Figure 2.6-15b Loss of RHR NSDC Event -

RHR and CS Pump Margin Ratio jontinued)

RHR Pump NPSH Margin Ratio - ASDC Dual Unit Interaction 280 260 2.40 2.20 2.00 --

1.6o 1200 600 20,600 40.600 60.600 80600 1W.60 120,00 ma I.s-Orn CS Pump NPSH Margin Ratio - ASDC Dual Unit Interaction 2.20 20w 1.0 600 M0600 44,600 60.600

$0D0 100.600 120,600 Urn. (seOnds) 2-330

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Figure 2.6-16a SORV with RPV Isolation Event RHR and CS Pump NPSH vs. Time RHR Pump NPSH - SORV with RPV Isolation 39.00 34.00 29.00 z

NPSHA 24.00 19.00.

NPSHIRm 14.00 600.00 5600.00 10600.00 15600.00 20600.00 Time (seconds)

CS Pump NPSH - SORV with RPV Isolation 38.00 33.0O S0RVdcoses on Low RPV Pfessure.

i 28.00 NPSHA' 23.00 NPSHRm 18.00 600 5600 10600 15600 20600 Tine (seconds) 2-331

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Figure 2.6-16b SORV with RPV Isolation Event RHR and CS Pump Margin Ratio RHR Pump NPSH Margin Ratio - SORV with RPV Isolation 280 2.60 2A0 1.40 -

600 5,600 10,600 15,600 20,600 cs Pump NPSH Margin Ratio - SORV with RPV Isolation 220D 1.80 IAO Zi 1.40 1,20------------------------------------------------------------------

1.00~

600 51600 10600 15600 20,600 rime (seconds) 2-332

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Figure 2.6-17a Appendix R Case Al RHR Pump NPSH vs. Time RHR Pump NPSH - Appendix R Event Case Al 40.00 35W00 30.00 25.00 20.00 15.00 10.00 I

~~Tk"(8" 2-333

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Figure 2.6-17b Appendix R Case Al RHR Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - Appendix R Case Al 2AtO 2.40-2200 2M Maetin Rado 1,40 1.20 1.00 10000 20D00 30000 40000 50000 60D00 7000 80000 Thu. (mon~

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Figure 2.6-18a Appendix R Case C1A RHR and CS Pump NPSH vs. Time RHR Pump NPSH -Appendix R Event Case CIA 35.00 33.00 31.00 29.00 27.00 25.00 23.00 21.00 19.00 17.00 15.00 9.1

NPSMR,

)00 58.000 109,000 159,000 209.000 259,000 Time (oc.s)

CS Pump NPSH - Appendix R Event Case CiA 40.00 35.00 30.00 N____

25.00 20,00 15.00 1_

j 3.000 28,000 53.000 78,000 103,000 128,000 153,000 178,000 203.000 228,000 253,000 Time (seconds) 2-335

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Figure 2.6-18b Appendix R Case C1A RHR and CS Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - Appendix R Case CIA 2.20 2.00 1.80 j1.60 1.40 1.20---------------------------------------------------------

1.00 91000 591000 109,000 159.00 209.000 259.00 lime (secands CS Pump NPSH Margin Ratio - Appendix R Case CIA 2.00 1.90 1.80 1.70 1.6 Pum NPS Magi

-ai Apeni Cas A

1.60 1,40 1.20 1.120 1.00 8

3.000 28.000 533000 78000 103.000 128,000 153.000 178,000 203,000 228,000 253.000 lime (second) 2-336

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Figure 2.6-19a Appendix R Case CIB RHR Pump NPSH vs. Time RHR Pump NPSH - Appendix R Event Case CiB 40.00 35.00 I

21.00__

NPSMA 20,00

-50 3,000 13.000 23.000 33,000 43,000 Time (seo) 53.000 63.000 73,000 83.000 2-337

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Figure 2.6-19b Appendix R Case CIB RHR Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - Appendix R Case CIB 2.60 2.40 2.20 2.00 11.80 Margin Ratio 1.40 1.00

  • 3,000 13,000 23,000 33,000 43,000 53,000 63.000 73,000 83,000 Tnw (wns I

2-338

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Figure 2.6-20a SBO Event RHR Pump NPSH vs. Time RHR Pump NPSH - SBO Event 30.00 35.00 3000 NPSHA 25.00 20,00 NPS~ia 1500 10.00 3600 6100 8=0 11100 13600 16100 18600 21100 23600 26100 28600 Twos (uSwor" 2-339

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Figure 2.6-20b SBO Event RHR Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - S10 Event 2280-2.60 240 2.20-z2.00-120 1.00-36DO 6100 8410 1110i 13600 16100 rww Iscccwh) 18600 21100 23600 26100 28600 2-340

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Figure 2.6-21a ATWS Event RHR Pump NPSH vs. Time RHR Pump NPSH - ATWS Event 40.00 35,00 30.00-NP 25.00 20.00 15.00 NPSH 10.00 680 1860 2660 3060 4680 5060 6660 7660 Tine (seconds) 2-341

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Figure 2.6-21b ATWS Event RHR Pump NPSH Margin Ratio RHR Pump NPSH Margin Ratio - ATWS Event 2.40 2.20 Morglin Raio 2.00 1.60---------------------------------

I 2

1.40 1.20 -

1.00 660 1660 2660 3660 4660 5660 6660 7660 Eu'.- (mconds 2-342

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Figure 2.6-22a RHR Pump NPSH vs. Time - Second Unit Safe Shutdown RHR Pump NPSH - Second PBAPS Unit Safe Shutdown 39.00 34,00 29.00 NPSHA 24.00 19.00 14.00 3.600 23.600 43,600 63.600 893600 103,600 123,600 Time (seconds) 2-343

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Figure 2.6-22b RHR Pump NPSH Margin Ratio - Second Unit Safe Shutdown RHR Pump NPSH Margin Ratio - Second PBAPS Unit Safe Shutdown 240 2.20 2.00 Margin Ratbo 1.80 1.40 1.00 3,600 23,600 43,600 63,600 83,600 103,600 123,600 Th-fs.conds) 2-344

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.7 Habitability, Filtration, and Ventilation 2.7.1 Control Room Habitability System Regulatory Evaluation Exelon reviewed the CR habitability system and control building layout and structures to ensure that plant operators are adequately protected from the effects of accidental releases of toxic and radioactive gases. A further objective of the review was to ensure that the CR can be maintained as the backup center from which technical support center personnel can safely operate in the case of an accident. The review focused on the effects of the proposed EPU on radiation doses, toxic gas concentrations, and estimates of dispersion of airborne contamination.

The regulatory acceptance criteria for the CR habitability system are based on: (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with postulated accidents, including the effects of the release of toxic gases; and (2) GDC-19, insofar as it requires that adequate radiation protection be provided to permit access and occupancy of the CR under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC-4 listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-11, Draft GDC-40, Draft GDC-42, Draft GDC-62, Draft GDC-63, Draft GDC-64, and Draft GDC-65. AST was approved at PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 AST License Amendments 269 and 273 (Reference 50),

respectively.

Current GDC-19 in 10 CFR 50 2-345

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Appendix A states that holders of operating licenses using an AST shall meet the requirements of current GDC-19. Therefore, current GDC-19 is applicable to PBAPS.

The CR Habitability System is described in PBAPS UFSAR Sections 7.12.5.6.3, "Control Room Ventilation Intake Alarm and Bypass," 7.12.5.5.3, "Control Room Ventilation Intake Radiation Monitor," 10.13, "Main Control Room Air Conditioning," and 14.9, "Evaluations Using AEC Method."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10).

The license renewal evaluation associated with the CR Habitability System is documented in NUREG-1769, Section 2.3.3.8. Management of aging effects on the CR Habitability System is documented in NUREG-1769, Section 3.3.8. Inspection and testing activities of the ventilation system are documented in NUREG-1769, Section 3.0.3.12.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 4.4 of the CLTR addresses the effect of CPPU on Main Control Room Atmosphere Control System.

The main control room emergency ventilation (MCREV) system functions during a DBA or an AOO to provide filtered air for personnel ventilation and pressurization of the CR envelope.

Redundant radiation detectors are provided at the outside air intakes to automatically initiate emergency flow through filtration units. With no change to detection and controls, the operation of the CR HVAC system is not affected.

PBAPS meets all CLTR dispositions. The topic addressed in this evaluation is:

4 ~

~

~

,~ 4 ~2 4

~~Disposition-

'"Result, Meets CLTR Iodine Intake

[Disposition The CLTR states that EPU increases the radioisotopes seen by the CR atmosphere, control system following an accident.

The radiological effect of EPU on the Control Room Emergency Ventilation (CREV) system (which includes HEPA filtration capability only) is due to an increase in the particulates, including particulate iodines, released during an accident. PBAPS has implemented the AST methodology which affects the DBA iodine release model. The EPU analyses were performed 2-346

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) for 102% of the EPU power level (i.e., 4030 MWt), and thus incorporate the increased EPU iodine release as well as the effects of the AST iodine release model. These analyses included the radiological consequences of the DBAs documented in Chapter 14 of the PBAPS UFSAR that potentially result in the most significant CR exposures. In all cases the CR doses were within regulatory limits.

The quantities and locations of gases and hazardous chemicals that could affect CR habitability are unaffected by EPU. Therefore, EPU has no effect on the potential toxic gas concentrations in the main CR.

Therefore, the iodine intake meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the ability of the CR habitability system to protect plant operators against the effects of accidental releases of toxic and radioactive gases have been reviewed. Exelon concludes the increase of toxic and radioactive gases that would result from the proposed EPU has been adequately evaluated. Exelon further concludes the CR habitability system will continue to provide the required protection following implementation of the proposed EPU. Based on this, Exelon concludes the CR habitability system will continue to meet the requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the CR habitability system.

2.7.2 Engineered Safety Feature Atmosphere Cleanup Regulatory Evaluation ESF atmosphere cleanup systems are designed for fission product removal in post-accident environments.

These systems generally include primary systems (e.g., in-containment recirculation) and secondary systems (e.g., standby gas treatment systems) for the fuel-handling building, CR, shield building, and areas containing ESF components. The review focused on the effects of the proposed EPU on system functional design, environmental design, and provisions to preclude temperatures in the adsorber section from exceeding design limits. The regulatory acceptance criteria for ESF atmosphere cleanup systems are based on: (1) GDC-19, insofar as it requires adequate radiation protection be provided to permit access and occupancy of the CR under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE (PBAPS has implemented the alternative source term), for the duration of the accident; (2) GDC-41, insofar as it requires systems to control fission products released into the reactor containment be provided to reduce the concentration and quality of fission products released to the environment following postulated accidents; (3) GDC-61, insofar as it requires systems that may contain radioactivity be designed to assure adequate safety under normal and postulated accident conditions; and (4) GDC-64, insofar as it requires means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including AOOs, and postulated accidents.

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Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967.

The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC-61 and GDC-64 listed in the Regulatory Evaluation above, with the exception of current GDC-41, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-1 1, Draft GDC-17, Draft GDC-69, and Draft GDC-70. There is no Draft GDC directly associated with current GDC-41. Current GDC-64 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 ODCM License Amendments 102 and 104 (Reference 35), respectively. Also, AST was approved at PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 AST License Amendments 269 and 273 (Reference 50), respectively. Current GDC-19 in 10 CFR 50 Appendix A states that holders of operating licenses using an AST shall meet the requirements of current GDC-19. Therefore, current GDC-19 is applicable to PBAPS.

The ESF atmosphere cleanup system at PBAPS is the SGTS. The SGTS is described in PBAPS UFSAR Sections 5.3.3, "Standby Gas Treatment System," and 14.9, "Evaluations Using AEC Method."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the SGTS is documented in NUREG-1769, Section 2.3.2.7. The management of the effects of aging on the SGTS is documented in NUREG-1769, Section 3.2.7.

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Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 4.5 of the CLTR addresses the effect of CPPU on the SGTS. The results of this evaluation are described below.

One of the two ESF atmosphere cleanup systems at PBAPS is the CR Ventilation System. The acceptability of this system under EPU conditions is addressed in Section 2.7.1. The second ESF atmosphere cleanup system is the SGTS.

The SGTS is designed to maintain secondary containment at a negative pressure and to provide an elevated release path for the exhaust air for removal of fission products potentially present during abnormal conditions. By providing for an elevated release path for airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA. The effect of a CPPU on the performance of the SGTS was evaluated in the CLTR based on two bounding analyses.

CLTR dispositions regarding the flow capacity and iodine removal capability of the SGTS have been addressed in Sections 2.6.6 and 2.5.2.1, respectively. However, it should be noted that the iodine (both particulate and gaseous) removal capabilities of the HEPA filters and the charcoal contained in the SGTS trains are not credited with respect to the post-LOCA accident scenario (based on the site AST SER - Reference 50).

Details regarding the SGTS evaluation based on post-LOCA operation after EPU implementation are described in Section 2.5.2.

The SGTS at PBAPS meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the SGTS have been reviewed and it was found the system design has adequately accounted for the increase of fission products and changes in expected environmental conditions that would result from the proposed EPU. Exelon concludes the SGTS will continue to provide adequate fission product removal in post-accident environments following implementation of the proposed EPU. Based on this, Exelon concludes the SGTS will continue to meet the requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the SGTS.

2.7.3 Control Room Area Ventilation System Regulatory Evaluation The function of the control room area ventilation system (CRAVS) is to provide a controlled environment for the comfort and safety of CR personnel and to support the operability of CR components during normal operation, AOOs, and DBA conditions. The review of the CRAVS focused on the effects that the proposed EPU will have on the functional performance of safety-related portions of the system. The review included: (1) the effects of radiation, combustion, and other toxic products; and (2) the expected environmental conditions in areas served by the CRAVS. The regulatory acceptance criteria for the CRAVS are based on: (1) GDC-4, insofar as 2-349

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (2) GDC-19, insofar as it requires adequate radiation protection be provided to permit access and occupancy of the CR under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE (PBAPS has implemented the alternative source term) for the duration of the accident; and (3) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967.

The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC-4 and GDC-60 listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-11, Draft GDC-40, Draft GDC-42, Draft GDC-62, Draft GDC-63, Draft GDC-64, Draft GDC-65, and Draft GDC-70. Current GDC-60 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 ODCM License Amendments 102 and 104 (Reference 35),

respectively. Also, AST was approved at PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 AST License Amendments 269 and 273 (Reference 50), respectively.

Current GDC-19 in 10 CFR 50 Appendix A states that holders of operating licenses using an AST shall meet the requirements of current GDC-19. Therefore, current GDC-19 is applicable to PBAPS.

The CRAVS is described in PBAPS UFSAR Sections 7.12.5.5.3, "Control Room Ventilation Intake Radiation Monitor,"

10.13, "Main Control Room Air Conditioning," 13.4.3.3.22, "Emergency Ventilation Air Supply to Control Room," and 14.9, "Evaluations Using AEC Method."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for 2-350

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the CRAVS is documented in NUREG-1769, Section 2.3.3.8. Management of aging effects on the CRAVS is documented in NUREG-1769, Section 3.3.8. Inspection and testing activities of the Ventilation system is documented in NUREG-1769, Section 3.0.3.12.

Technical Evaluation The HVAC systems discussed in the CLTR are only those that have power dependent heat loads.

Power dependent HVAC systems require ((

)) The CR HVAC System maintains temperature and humidity conditions suitable for personnel comfort and for equipment reliable operation inside the CR envelope.

The CR HVAC System also maintains the CR envelope at positive pressure to inhibit air infiltration (see Section 2.7.1). Heat loads for the CR area envelope include boundary transmission, lighting and equipment such as CR panels. These heat loads are not affected by the slightly higher process temperatures that may result from EPU, thus they are not power dependent. EPU does not add any electrical or electronic equipment to the CR with the potential exception of new control valve hand switches and instrumentation in support of the modification to increase SPC capability. The effect of these devices of CR HVAC will be evaluated as part of the standard design process for this modification. EPU may add some amperage for control and indication signals, but the resulting changes in temperature are considered negligible. The change of conductance of heat through the building structure to the CR is expected to be negligible. The heat load increase is expected to be insignificant in comparison with the total CR heat load. Therefore, the CR temperature increase is expected to be insignificant as a result of EPU implementation. Table 2.7-1 shows the impact of EPU on the Ventilation Systems.

There is no increase in toxic or asphyxiant gas release that may result from EPU. The control of the concentration of airborne radioactive material in the CR envelope during AQOs and after postulated accidents is accomplished by the CR HVAC system described in Section 2.7.1.

ASD will add some electrical or electronic equipment in the cable spreading room and instrumentation in support of the ASD modification (see PBAPS EPU LAR Attachment 9, Section 2, for additional information). No adverse impact is anticipated based on the preliminary design input, however, evaluation of the impact of these devices on cable spreading room and CR HVAC will be performed as part of the standard design process for this modification.

There is no change to the CR HVAC System configuration or system parameters as a result of EPU.

Conclusion The effects of the proposed EPU on the ability of the CRAVS to provide a controlled environment for the comfort and safety of CR personnel and to support operability of CR components have been reviewed. Exelon concludes that the review has adequately accounted for the increase of toxic and radioactive gases that would result from a DBA under the conditions of 2-351

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) the proposed EPU, and associated changes to parameters affecting environmental conditions for CR personnel and equipment. Accordingly, Exelon concludes the CRAVS will continue to provide an acceptable CR environment for safe operation of the plant following implementation of the proposed EPU. Exelon also concludes the system will continue to suitably control the release of gaseous radioactive effluents to the environment. Based on this, Exelon concludes the CRAVS will continue to meet the requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the CRAVS.

2.7.4 Spent Fuel Pool Area Ventilation System Regulatory Evaluation The function of the spent fuel pool area ventilation system (SFPAVS) is to maintain ventilation in the SFP equipment areas, permit personnel access, and control airborne radioactivity in the area during normal operation, AOOs, and following postulated fuel handling accidents (FHAs).

The review focused on the effects of the proposed EPU on the functional performance of the safety-related portions of the system. The regulatory acceptance criteria for the SFPAVS are based on: (1) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents; and (2) GDC-61, insofar as it requires systems which contain radioactivity be designed with appropriate confinement and containment.

Peach Bottom Current Licensing Basis The PBAPS design does not include a separate SFPAVS. Ventilation in this area is provided by the Reactor Building HVAC system under normal conditions. The Reactor Building Ventilation System is separated into two areas, the area above the refueling floor and the area below the refueling floor. The ventilation to the SFP area is provided by the Reactor Building Ventilation System in the area above the refuel floor.

The SFPAVS is described in PBAPS UFSAR Section 5.3.2, "Reactor Building Heating and Ventilating System."

Technical Evaluation The PBAPS design does not include a separate SFPAVS. As described above, during normal power operation the Reactor Building ventilation system provides ventilation from the Refueling Floor, i.e., the SFP Area. The SGTS performs this function during abnormal plant operations (accident conditions) and its EPU evaluation is described in Sections 2.5.2.1 and 2.6.6.

Conclusion The effects of the proposed EPU on the SFPAVS have been reviewed. Exelon concludes the review has adequately accounted for the effects of the proposed EPU on the system's capability to: (1) maintain ventilation in the SFP equipment areas; (2) permit personnel access; (3) control airborne radioactivity in the area; (4) control release of gaseous radioactive effluents to the environment; and (5) provide appropriate containment.

Based on this, Exelon concludes the SFPAVS will continue to meet the requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the SFPAVS.

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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.7.5 Reactor, Turbine, and Radwaste Building Ventilation Systems Regulatory Evaluation The function of the reactor building, the radwaste area and the turbine area ventilation systems is to maintain ventilation in the reactor building and radwaste equipment and turbine areas, permit personnel access, and control the concentration of airborne radioactive material in these areas during normal operation.

During AOOs and after postulated accidents, control of the concentration of airborne radioactive material in the reactor building is controlled by use of the SGTS. The review focused on the effects of the proposed EPU on the functional performance of the safety-related portions of these systems. The regulatory acceptance criteria for these ventilation systems are based on GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents.

Peach Bottom Current Licensing Basis Current GDC-60 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 ODCM License Amendments 102 and 104 (Reference 35), respectively. The Reactor Building, Radwaste Area, and Turbine Area Ventilation Systems are described in PBAPS UFSAR Sections 5.3.2, "Reactor Building Heating and Ventilation System" and 10.15, "Plant Heating Ventilating and Air Conditioning Systems."

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.6 of the CLTR addresses the effect of CPPU on CLTR Power Dependent HVAC. The results of this evaluation are described below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

Meets CLTR Power dependent HVAC performance

[et Dp ti Disposition The CLTR states that EPU results in slightly higher process temperatures and electrical loads on the HVAC system.

The turbine building, reactor building, DW, and radwaste building ventilation systems evaluated in the CLTR are only those that are power dependent. The power dependent HVAC systems consist mainly of heating, cooling supply, exhaust, and recirculation units in the Turbine Building, Reactor Building, Radwaste Building, and the DW. The control of the concentration of airborne radioactive material in the Reactor Building is controlled by the Reactor Building Ventilation system during normal operation.

During AOOs and after postulated accidents, 2-353

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) control of the concentration of airborne radioactive material in the reactor building is controlled by use of the SGTS described in Sections 2.5.2 and 2.6.6. Monitoring of the Radwaste Building exhaust, and the Turbine Building exhaust, is not affected by EPU; additionally, monitoring of the Turbine Gland Sealing System and the Mechanical Vacuum Pump System are not affected by EPU.

At PBAPS, the normal operating EPU process temperatures affecting the normal HVAC loads increase slightly from CLTP values. However, the increases in temperatures and heat loads will not have a significant effect on the HVAC system. The effect of condensate pump upgrades on HVAC systems will be evaluated as part of the normal modification process.

Currently, no modifications are planned to any HVAC or atmospheric clean-up system and there is no significant EPU effect on HVAC systems during normal operation or accident conditions.

During normal operation MS tunnel temperatures may increase slightly due to an increase in FW temperature.

However, the increase in temperature will be less than 0.5°F. Any heat load increases in the DW are not considered significant, and are within existing system margin. The HVAC systems serving the remaining areas served by these HVAC systems are unaffected by the EPU because the process temperatures and equipment heat loads are not power dependent and not affected by EPU. Table 2.7-1 shows the impact of EPU on the Ventilation Systems.

ASD installation (see PBAPS EPU LAR Attachment 9, Section 2, for additional information) will result in a reduced heat load on the MG Sets Room HVAC system and may increase heat load on turbine building HVAC system. The heat load increase with ASD does not result in significant temperature increase in the turbine area at EPU conditions.

The maximum temperature increases calculated are considered within the daily anticipated temperature fluctuation.

Based on a review of design basis calculations, the design of the HVAC is adequate for EPU.

Therefore, the power dependent HVAC performance meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the Reactor, Turbine, and Radwaste Building HVAC systems have been reviewed. Exelon concludes the review has adequately accounted for the effects of the proposed EPU on the capability of these systems: (1) to maintain ventilation in the auxiliary and radwaste equipment areas and in the turbine area; (2) permit personnel access; (3) control the concentration of airborne radioactive material in these areas; and (4) control release of gaseous radioactive effluents to the environment.

Based on this, Exelon concludes the Reactor, Turbine, and Radwaste Building HVAC systems will continue to meet the requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the Reactor, Turbine, and Radwaste Building HVAC systems.

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NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) 2.7.6 Engineered Safety Feature Ventilation System Regulatory Evaluation The function of the engineered safety feature ventilation system (ESFVS) is to provide a suitable and controlled environment for ESF components following certain anticipated transients and DBAs. The review for the ESFVS focused on the effects of the proposed EPU on the functional performance of the safety-related portions of the system. The review also covered: (1) the ability of the ESF equipment in the areas being serviced by the ventilation system to function under degraded ESFVS performance; (2) the capability of the ESFVS to circulate sufficient air to prevent accumulation of flammable or explosive gas or fuel-vapor mixtures from components (e.g., storage batteries and stored fuel); and (3) the capability of the ESFVS to control airborne particulate material (dust) accumulation. The regulatory acceptance criteria for the ESFVS are based on: (1) GDC-4, insofar as it requires SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (2) GDC-17, insofar as it requires onsite and offsite electric power systems be provided to permit functioning of SSCs important to safety; and (3) GDC-60, insofar as it requires the plant design include means to control the release of radioactive effluents.

Peach Bottom Current Licensing Basis The general design criteria listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable PBAPS principal design criteria predate these criteria. The PBAPS principal design criteria are listed in UFSAR Section 1.5, "Principal Design Criteria." In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32 FR 10213 (Reference 9), July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, Philadelphia Electric Company (PECO), the predecessor to Exelon, performed a comparative evaluation of the design basis of PBAPS Units 2 and 3 against the AEC proposed General Design Criteria of 1967. The PBAPS UFSAR, Appendix H, "Conformance to AEC (NRC)

Criteria," contains this comparative evaluation. UFSAR Appendix H provides a comparative evaluation with each of the groups of criteria set out in the July 1967 AEC release. As to each group of criteria, there is a statement of Exelon's understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.

Following a restatement of each of the proposed criteria is a list of references to locations in the PBAPS UFSAR where there is subject matter relating to the intent of that particular criteria.

For the current GDC listed in the Regulatory Evaluation above, the PBAPS comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as "Draft GDC") is contained in PBAPS UFSAR Appendix H: Draft GDC-24, Draft GDC-39, Draft GDC-40, Draft GDC-42, and Draft GDC-70. Current GDC-60 is applicable to PBAPS as described in the NRC SER for PBAPS Unit 2 and Unit 3 ODCM License Amendments 102 and 104 (Reference 35), respectively. Also, current GDC-17 listed the Regulatory Evaluation above 2-355

NEDO-33566 - REVISION 0 NON-PROPRIETARY INFORMATION - CLASS I (PUBLIC) is applicable to PBAPS as described in UFSAR Sections 8.4.8 and 8.5.6, "Compliance with Safety Guides."

The ESF ventilation system is discussed in PBAPS UFSAR Sections 5.3.2, "Reactor Building Heating and Ventilating System," 10.14, "Emergency Ventilating Systems," and 10.15, "Plant Heating Ventilating and Air Conditioning Systems."

In addition to the evaluations described in the PBAPS UFSAR, PBAPS's systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the "Safety Evaluation Report Related to the License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3," NUREG-1769, dated March 2003 (Reference 10). The license renewal evaluation associated with the ESF ventilation system is documented inNUREG-1769, Sections 2.3.3.9, 2.3.3.10, and 2.3.3.11. Management of aging effects on the ESFVS is documented in NUREG-1769, Sections 3.3.9, 3.3.10, and 3.3.11.

Technical Evaluation NEDC-33004P-A, Revision 4, "Constant Pressure Power Uprate," Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of CPPUs. Section 6.6 of the CLTR addresses the effect of CPPU on CLTR Power Dependent HVAC. The results of this evaluation are described below.

PBAPS meets all CLTR dispositions. The topics addressed in this evaluation are:

~CLTiR*

~PBA.PS~

~W~/~2'~Disposition '/'i esult Meets CLTR Power dependent HVAC Performance

[Disposition The CLTR states that slightly higher process temperatures and electrical loads occur as a result of EPU.

The ESF HVAC systems consist mainly of heating, cooling supply, exhaust, and recirculation units serving the Emergency Switchgear and Battery Rooms, the Standby Diesel Generator Rooms, the ESW / HPSW Compartments, and the ECCS Pump Rooms (RHR, HPCI, CS, and RCIC).

These systems do not function to control the concentration of airborne radioactive material in these areas during normal operation, during AOOs, and after postulated accidents.

The control of the concentration of airborne radioactive material in the secondary containment during normal operation, during AOOs, and after postulated accidents is accomplished using the SGTS described in Sections 2.5.2 and 2.6.6.

During normal operation, the HVAC systems serving these areas are unaffected by the EPU because the process temperatures remain bounded by CLTP conditions.

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The design basis post-LOCA Reactor Building temperatures will not increase with EPU. The suppression pool temperature will not increase with EPU due to the RHR heat exchanger cross-tie modification, and therefore, heat loads and temperatures in the ECCS pump rooms will not increase.

Additionally, there are no major equipment modifications in the Emergency Switchgear and Battery Rooms, and the ESW / HPSW compartments, and therefore, design heat loads in these rooms will not change with EPU. The operating heat load in the Standby Diesel Generator Rooms will increase slightly, but is bounded by the design heat load with enough margin available to maintain design temperatures. Table 2.7-1 shows the impact of EPU on the Ventilation Systems.

Therefore, the power dependent HVAC performance meets all CLTR dispositions.

Conclusion The effects of the proposed EPU on the ESFVS have been evaluated. Exelon has determined the evaluation adequately accounted for the effects of the proposed EPU on the ability of the ESFVS to provide a suitable and controlled environment for ESF components. Exelon further concludes the ESFVS will continue to assure a suitable environment for the ESF components following implementation of the proposed EPU.

Exelon also concludes the ESFVS will continue to suitably control the release of gaseous radioactive effluents to the environment following implementation of the proposed EPU. Based on this, Exelon concludes the ESFVS will continue to meet the requirements of the current licensing basis. Therefore, Exelon finds the proposed EPU acceptable with respect to the ESFVS.

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Table 2.7-1 EPU Impact on Ventilation Systems Turbine Building Ventilation System Increases in process temperatures results in slight temperature increases of less than 2°F.

The design of the Turbine Building HVAC system is adequate to handle the increase in heat load.

EPU does not result in significant temperature increases in areas of the Reactor Building.

The only area which is expected to increase in area temperature is the MS tunnel. However, the expected increase in this area is Reactor Building

< 0.5°F, which is not significant.

The design of the HVAC system is Ventilation System adequate for EPU.

Design basis post-LOCA temperatures will not increase with EPU. Post-LOCA suppression pool temperatures will not increase with EPU.

Drywell Ventilation EPU will not result in a significant increase in drywell heat load or area temperature increases (< 0.5°F). The drywell HVAC system is adequate to System handle the small increase in heat load.

Radwaste Building Negligible effect.

Ventilation System Ventilation Systems for Miscellaneous Rooms and Negligible effect due to EPU.

Buildings Control Room Negligible effect due to slight increase in control and indication signals. No VCoo Rheat-generating equipment added to the CR, with the exception of new hand HVAC switch controls and instrumentation.

Emergency Negligible effect due to EPU.

Some electrical operational loads may Ventilating Systems increase slightly, but will stay below design loads.

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