ML14160A699

Extended Power Uprate License Amendment Request - Supplement 27 Supplemental Information

ML14160A699
Person / Time
Site:	Peach Bottom
Issue date:	06/05/2014
From:	Borton K Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML14160A699 (30)

Text

n Exeton Generation.

10 CFR 50.90 June 5, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Extended Power Uprate License Amendment Request - Supplement 27 Supplemental Information

Reference:

1.

Exelon letter to the NRC, "License Amendment Request -

Extended Power Uprate," dated September 28, 2012 (ADAMS Accession No. ML122860201)

2.

Exelon letter to the NRC, "Supplemental Information Supporting Request for License Amendment Request - Extended Power Uprate - Supplement 9," dated August 22, 2013 (ADAMS Accession No. ML13240A002)

In accordance with 10 CFR 50.90, Exelon Generation Company, LLC (EGC) requested amendments to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, respectively (Reference 1).

Specifically, the proposed changes would revise the Renewed Facility Operating Licenses (RFOLs) to implement an increase in rated thermal power from 3514 megawatts thermal (MWt) to 3951 MWt. EGC performed an impact review of RFOL Amendments issued since submittal of the proposed extended power uprate (EPU) license amendment (Reference 1). Several recently approved RFOL amendments changed Technical Specifications (TS) that affect the proposed EPU TS changes. Therefore, revised TS page markups are being provided to clearly delineate the EPU-required changes from the approved TS. This letter also provides minor corrections to values reported in the Power Uprate Safety Analysis Report (PUSAR) provided in Attachment 6 of Reference 1.

Additionally, this letter provides a revised response to SRXB-RAI-24 that was provided in Supplement 9 (Reference 2). These corrections do not impact the previously provided conclusions.

A description of the TS changes are included in Attachment 1. Revised markups of the affected TS pages are provided in Attachment 2. Revised markups of the affected TS Bases pages are provided for information only in Attachment 3.

A Co C

Supplemental Information June 5, 2014 Page 2 In addition, the cover to the EPU amendment request (Reference 1) described EGC's plan for implementation of the EPU amendment. The following statement amends the original implementation plan:

Once approved, the amendment will be implemented prior to startup from refueling outage P2R20 for Unit 2 and P3R20 for Unit 3.

EGC has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the U. S. Nuclear Regulatory Commission in Reference 1. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Further, the additional information provided in this submittal does not affect the bases for EGC concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the Commonwealth of Pennsylvania and the State of Maryland of this application by transmitting a copy of this letter with attachments to the designated State Officials.

There are no regulatory commitments contained in this letter.

Should you have any questions concerning this letter, please contact Mr. David Neff at (610) 765-5631.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 5 th day of June 2014.

jepctfful Kevin F Borton Sr. Manager, Licensing - Power Uprate Exelon Generation Company, LLC Attachments

1. Description of Supplemental Information and Technical Specifications Changes

2. Technical Specifications Marked Up Pages

3. Technical Specifications Bases Marked Up Pages cc:

USNRC Region I, Regional Administrator w/attachments USNRC Senior Resident Inspector, PBAPS w/attachments USNRC Project Manager, PBAPS w/attachments R. R. Janati, Commonwealth of Pennsylvania w/attachments S. T. Gray, State of Maryland w/attachments Peach Bottom Atomic Power Station Units 2 and 3 NRC Docket Nos. 50-277 and 50-278 DESCRIPTION OF SUPPLEMENTAL INFORMATION and TECHNICAL SPECIFICATIONS CHANGES

EPU LAR Supplement 27 Supplemental Information Page 1 of 5 DESCRIPTION OF SUPPLEMENTAL INFORMATION and TECHNICAL SPECIFICATIONS CHANGES

1.0 INTRODUCTION

Exelon Generation Company, LLC (EGC) submitted a License Amendment Request (LAR) for an Extended Power Uprate (EPU) for the Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3 on September 28, 2012 (Reference 1). This letter provides supplemental information regarding:

Two PBAPS Renewed Facility Operating License (RFOL) Amendments were issued following the EPU LAR submittal and included TS that affect the EPU-proposed changes.

Amendment Nos. 290 and 293, for Unit 2 and Unit 3, approving the implementation of a 3% lift setpoint tolerance for the Safety Relief Valves (SRVs) and Safety Valves (SVs) (Reference 2).

Amendment Nos. 288 and 291, for Unit 2 and Unit 3, approving the addition of the Drywell Spray function into the TS as Section 3.6.2.5 (Reference 3).

" Corrections to reflect the impact of a recent analysis revision to better account for reactor recirculation pump inertia in the transient analysis. These corrections have a small effect on results provided in the Power Uprate Safety Analysis Report (PUSAR) (Reference 1, Attachment 6). The conclusions regarding the capability of the plant to safely respond and mitigate the events are not affected.

Correction of the response to SRXB-RAI-24 contained in Supplement 9 (Reference 5) regarding Reactor Core Isolation Cooling (RCIC) system pump fluid temperature. The conclusion regarding the capability of the RCIC system to operate within design limits is not affected.

A detailed description and technical justification of each of these changes are provided in the following sections of this attachment. Revised markups of the affected TS pages are provided in Attachment 2. Revised markups of the affected TS Bases pages are provided for information only in Attachment 3.

2.0 SAFETY RELIEF VALVE SETPOINT TOLERANCE The NRC issued Amendments 290 (Unit 2) and 293 (Unit 3) on May 5, 2014 (Reference 2). These amendments revised the TSs to: (1) increase the allowable as-found safety relief valve (SRV) and safety valve (SV) lift setpoint tolerance from t 1% to t 3%; (2) increase the required number of operable SRVs and SVs from 11 to 12; and (3) increase the Standby Liquid Control System (SLCS) pump discharge pressure from 1255 psig to 1275 psig. These amendments included three revised TS pages for each unit:

TS page 3.1-23 was revised to reflect a higher SLCS pump discharge pressure of 1275 psig in TS Surveillance Requirement (SR) 3.1.7.8.

" TS page 3.4-8 was revised to reflect the revised Limiting Condition for Operation (LCO) to require the operability of 12 valves.

EPU LAR Supplement 27 Supplemental Information Page 2 of 5 TS page 3.4-9 was revised to reflect the new lift setpoint tolerances for the SRVs and SVs.

The EPU LAR, as described in Reference 1, Attachment 1, included EPU-required changes to each of these pages. The impact to the EPU proposed TS due to the approved amendments, is described below:

The EPU LAR had proposed the SLCS discharge pressure in TS SR 3.1.7.8 be revised from > 1255 psig to > 1265 psig. The value of 1275 psig approved in the recent amendments bounds the proposed EPU requirement. As such, EGC withdraws the proposed change to the SLCS pump discharge pressure value in TS SR 3.1.7.8. The revised markup of this page reflects there is no EPU change associated with this parameter.

The other EPU changes proposed to this page, or the analyses on which they were based, are not affected by the issued SRV/SV lift setpoint tolerance amendments.

The EPU LAR had proposed a change to TS 3.4.3.1 LCO to increase the total number of operable SRVs/SVs to 13. The revised markup of this page reflects the EPU required increase from 12 valves to 13 valves.

The original EPU LAR analyses also used the more conservative 3% lift setpoint tolerance value for each SRV and SV. There is no impact on the EPU LAR due to the 3% lift setpoint values established in Amendments 290/293. However, the lift setpoint values in TS SR 3.4.3.1 now reflect the 3% tolerance.

The other EPU change proposed to this page, or the analyses on which it was based, is not affected by the issued SRV/SV lift setpoint tolerance amendments.

Markups of the recently approved TS pages 3.1-23, 3.4-8 and 3.4-9 reflecting these EPU changes are included in Attachment 2. Markups of the affected TS Bases pages are included for information only in Attachment 3.

3.0 DRYWELL SPRAY TECHNICAL SPECIFICATION The NRC issued Amendments 288 (Unit 2) and 291 (Unit 3) on June 18, 2013 (Reference 3). These amendments added TS 3.6.2.5 to address requirements for the Drywell Spray function of the Residual Heat Removal System. Drywell Spray is a mode of the RHR system which is assumed to be initiated under post-accident conditions to reduce the temperature and pressure of the primary containment atmosphere. These amendments included two new TS pages for each unit: 3.6-30a and 3.6-30b.

As discussed in the EPU LAR (Reference 1) Attachment 1, PBAPS is implementing modifications to eliminate the need for containment accident pressure credit in assuring adequate net positive suction head (NPSH) is available for the Emergency Core Cooling Systems (ECCS) pumps. One of these modifications is the RHR cross-tie modification.

This change will enable the flow from a single RHR pump to be divided between two RHR heat exchangers.

EPU LAR Supplement 27 Supplemental Information Page 3 of 5 EPU safety analyses assume that, at one hour into the response to certain design basis events, the RHR cross-tie valve is opened to establish a containment cooling configuration. Control switches in the Control Room will provide the operator with the capability to transfer power for the motor-operated RHR cross-tie and RHR flow control valves from the normal source to the alternate source to assure the availability of this cooling function when needed.

10 CFR 50.36 (c)(3), Surveillance Requirements, states "Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." Because the PBAPS RHR system cross-tie modification relies on the capability for the selected valves to be powered from a normal source or an alternate source, it is appropriate to include a verification of this ability in the TS.

In EPU LAR Supplement 5 (Reference 4), PBAPS proposed the addition of two new TS SR, SR 3.6.2.3.3 for the RHR Suppression Pool Cooling TS and SR 3.6.2.4.3 for the RHR Suppression Pool Spray TS. These SRs provide for the verification of the power transfer capability for the RHR cross-tie and RHR flow control valves. With the implementation of the TS for Drywell Spray, a new TS SR is needed to support the same EPU design basis for this RHR system function. Specifically, the capability to transfer power for the motor-operated RHR cross-tie and RHR flow control valves for the containment cooling functions of the RHR system is relied upon in the safety analyses.

The following TS SR 3.6.2.5.3 is proposed to be conducted in accordance with the Surveillance Frequency Control Program:

Verify manual transfer capability of power supply for the RHR motor-operated flow control valve and the RHR cross-tie motor-operated valve from the normal to the alternate source.

In addition, similar to the proposed RHR suppression pool cooling and RHR suppression pool spray bases (TS bases sections 3.6.2.3 and 3.6.2.4, respectively), the RHR cross-tie description will be added to the Background and LCO sections of the RHR Drywell Spray TS bases.

A markup of the affected TS page 3.6-30b is provided in Attachment 2 and a markup of the affected TS Bases pages are provided for information only in Attachment 3.

4.0 REACTOR RECIRCULATION PUMP INERTIA A re-evaluation of the EPU ATWS analysis regarding the reactor recirculation pump-motor inertia resulted in a small reduction in the margin for the peak vessel bottom pressure. The results of the original EPU analysis are provided in PUSAR Section 2.8.5.7, Anticipated Transients Without Scram and Table 2.8-8, PBAPS Results for ATWS Analysis. Based on the revised analysis, the peak vessel bottom pressure in Table 2.8-8 increased slightly from 1458 to 1461 psig. This remains well below the ASME Service Level C limit of 1500 psig. The values for the other parameters in Table 2.8-8 (peak clad temperature (PCT), suppression pool temperature, containment pressure and cladding oxidation) conservatively bound the resulting values in the new analysis and are not being revised in this supplement.

EPU LAR Supplement 27 Supplemental Information Page 4 of 5 The above increase in the peak vessel bottom pressure results in a minor (i.e., 1 psi) impact on certain SLCS parameters described in the PUSAR. These impacts are detailed below:

PUSAR Original New PUSAR Section Parameter PUSAR Value Value Value Maximum SLCS pump 2.2.4.2 discharge pressure during 1265 psig 1266 psig ATWS 2.8.4.5.2 Peak vessel lower plenum 1190.3 psig / 1191.3 psig /

pressure at SLCS initiation 1205 psia 1206 psia 2.8.4.5.2 SLCS Relief Valve setpoint 185 psi 184 psi margin 2.8.5.7.1 Peak vessel lower plenum 1205 psia 1206 psia pressure at SLCS initiation 1

As part of the recently approved RFOL Amendments 290/293 discussed in Section 2.0 of this supplement, the TS SR 3.1.7.8 for the SLCS pump discharge pressure was increased to > 1275 psig. Since this bounds the maximum SLCS pump discharge pressure during an ATWS event at EPU conditions, there will be no EPU-related TS change proposed to this parameter.

The conclusions of the ATWS and SLCS evaluations in the PUSAR are unchanged regarding the acceptance criteria following implementation of the proposed EPU.

5.0 CORRECTION OF RESPONSE TO SRXB-RAI-24 During the review of the NRC draft Safety Evaluation Report for the PBAPS EPU LAR to verify the factual accuracy of the information, EGC identified that information provided in response to the SRXB-RAI-24 in Supplement 9 (Reference 5) was in error. The related information in the response to SRXB-RAI-30 has been confirmed to be correct.

Changes to the response to SRXB-RAI-24 is provided below with additions indicated by bolded characters and deletions indicated with

.',iketh.e.h markers. The NRC had asked:

Are there any instances under EPU conditions where the pump would be operating outside of this temperature range? If so, what are the conditions and how are they addressed for this EPU?

The EGC response is hereby corrected to read:

The only EPU analyses in which RCIC operation is credited are: Appendix R Method A (described in Reference 24-1 Section 2.5.1.4), Station Blackout (described in Reference 24-1 Section 2.3.5), ATWS (described in Reference 24-1 Section 2.8.5.7) and Loss of Feedwater Flow Event (described in Reference 24-1 Section 2.8.5.2.3.1). For the loss of feedwater flow event, there is no elevated suppression pool temperature. For the Appendix R, Station B!Gcko't and ATWS analyses, the RCIC pump suction source credited is exclusively from the

EPU LAR Supplement 27 Supplemental Information Page 5 of 5 condensate storage tank, which has a temperature range of 40 OF to 140 OF. For the Station Blackout event, RCIC is aligned to the suppression pool for approximately 0.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The suppression pool temperature peaks at 163

°F during RCIC operation. In order to mitigate certain events, such as SBO, system analysis has shown that RCIC can operate up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 180 OF.

Thoef or~e, thoro aro no safety analyoses foar EMU whore ROIC would operate outede +the docign temperature range of 40

+-

to 110 OF.

6.0 REFERENCES

1.

Exelon letter to the NRC, "License Amendment Request - Extended Power Uprate," dated September 28, 2012 (ADAMS Accession No. ML122860201)

2.

NRC letter to Exelon, "Issuance of Amendments Re: Safety Relief Valve and Safety Valve Lift Setpoint Tolerance," dated May 5, 2014 (ADAMS Accession No. ML14079A102)

3.

NRC letter to Exelon, "Issuance of Amendments Re: Revise Technical Specifications to Add Residual Heat Removal System Drywell Spray Function Requirements," dated June 18, 2013 (ADAMS Accession No. ML13109A463)

4.

Exelon letter to the NRC, "Supplemental Information Supporting Request for License Amendment Request - Extended Power Uprate - Supplement 5," dated June 27, 2013 (ADAMS Accession No. ML13182A025)

5.

Exelon letter to the NRC, "Supplemental Information Supporting Request for License Amendment Request - Extended Power Uprate - Supplement 9," dated August 22, 2013 (ADAMS Accession No. ML13240A002)

Peach Bottom Atomic Power Station Units 2 and 3 NRC Docket Nos. 50-277 and 50-278 TECHNICAL SPECIFICATIONS MARKED UP PAGES

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify the quantity of B 1. st.red 4n the in accrdance.

62C tank.- it 1t62 ibm..

Surveillance D eleted Control Programf.

SR 3.1.7.8 Verify each pump develops a flow rate In accordance gpm at a discharge pressure with the 25 psig.

Inservice Testing 49.1

-Program SR 3.1.7.9 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel, with the Surveillance Frequency Control Program.

SR 3.1.7.10 Verify sodium pentaborate atom percent B-1Q Ance within 8 enrichment is

.ithi the lims of hours after

'Table 3.1.7 1.

addition to eo SLC tank In accordance with the Surveillance Frequency Control Program AND PBAPS UNIT 2 3.1-23 Amendment No.

290

SRVs and SVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs)

LCO 3.4.3 The safety function of 4-valves (any combination of SRVs and SVs) shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required A.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SRVs or SVs inoperable.

AND A.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> PBAPS UNIT 2 3.4-8 Amendment No.

290

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the required SRVs and SVs are as follows:

In accordance with the Inservice Testing Program Number of SRVs 4

4 3

Number of Setpoint (psiQ) 1135 +/- 34.1 1145 +/- 34.4 1155 +/- 34.7 Setpoint SVs (psiq) 21260

+/- 37.8 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 Verify each required SRV actuator strokes In accordance when manually actuated in the with the depressurization mode.

Surveillance Frequency Control Program.

PBAPS UNIT 2 3.4-9 Amendment No.

290

RHR Drywell Spray 3.6.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.5.1 Verify each RHR drywell spray subsystem In accordance

manual, power operated, and automatic with the valve in the flow path that is not locked, Surveillance sealed, or otherwise secured in position Frequency is in the correct position or can be Control aligned to the correct position.

Program.

SR 3.6.2.5.2 Verify each drywell spray nozzle is In accordance unobstructed.

with the Surveillance Frequency Control Program.

SR 3.6.2.5.3 Verify manual transfer capability of power supply for the RHR motor-operated flow control valve and the RHR cross-tie motor-operated valve from the normal source to the alternate source.

PBAPS UNIT 2

3. 6-30b Amendment No.

288

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify the quantity of B 1Q stered in the In a..,rdance SLC tank is Ž! 162.7 i*bm.

with Surveillance Frequency

{Deleted PIgr a

-ffl.

SR 3.1.7.8 Verify each pump develops a flow rate In accordance

> 4-.- gpm at a discharge pressure with the 275 psig.

Inservice Testing 49.1 Program SR 3.1.7.9 Verify flow through one SLC subsystem from In accordance pump into reactor pressure vessel.

with the Surveillance Frequency Control Program.

SR 3.1.7.10 Verify sodium pentaborate atom percent B-1Q Once within 8 eenrichment is within the limits of

/

hours after Table 3.1.7 1.

addition to SLC tank enrichment to > 92.0.

In accordance with the Surveillance Frequency Control Program AND PBAPS UNIT 3 3.1-23 Amendment No.

293

SRVs and SVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety Relief Valves (SRVs) and Safety Valves (SVs)

LCO 3.4.3 The safety function of 4--2 valves (any combination of SRVs and SVs) shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required A.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SRVs or SVs inoperable.

AND A.2 Be in MODE 4.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> I

PBAPS UNIT 3 3.4-8 Amendment No.

293

SRVs and SVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the required SRVs and SVs are as with the follows:

Inservice Testing Program Number of Setpoint SRVs (psio) 4 1135 +/- 34.1 4

1145 +/- 34.4 3

1155 +/- 34.7 Number of Setpoint SVs Z

.=

(psio) 21260

+/- 37.8 Following testing, lift settings shall be within +/- 1%.

SR 3.4.3.2 Verify each required SRV actuator strokes In accordance when manually actuated in the with the depressurization mode.

Surveillance Frequency Control Program.

PBAPS UNIT 3 3.4-9 Amendment No.

293

RHR Drywell Spray 3.6.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.5.1 Verify each RHR drywell spray subsystem In accordance

manual, power operated, and automatic with the valve in the flow path that is not locked, Surveillance sealed, or otherwise secured in position Frequency is in the correct position or can be Control aligned to the correct position.

Program.

SR 3.6.2.5.2 Verify each drywell spray nozzle is In accordance unobstructed.

with the Surveillance Frequency Control Program.

SR 3.6.2.5.3 Verify manual transfer capability of power supply for the RHR motor-operated flow control valve and the RHR cross-tie motor-operated valve from the normal source to the alternate source.

PBAPS UNIT 3

3. 6 -30b Amendment No.

291 Peach Bottom Atomic Power Station Units 2 and 3 NRC Docket Nos. 50-277 and 50-278 TECHNICAL SPECIFICATIONS BASES MARKED UP PAGES FOR INFORMATION ONLY

SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. The rate of negative reactivity insertion is increased by using highly enriched boron in the SLC System solution that increases the rate of Boron-lO injection and functions to shutdown the reactor core faster. This limits the heat generated that is transferred to the suppression pool during an ATWS event. Limiting the heat transferred to the suppression pool maintains the pool below design limits, which ensures adequate NPSH is available for the ECCS pumps without credit for containment accident pressure. This test confirms one point on the pump design curve and 3P 17S

( c.nt inu.ed)

T hp rpa W !rd p ::

..nt.in. an additirnal am,..unt of B 10 equal to 25 e f the min.......required amunt of B 10 ncccssary to ohutdawn the reactor, to aeeount fer potential leakage an~d imfper-feet mflm )

l*

1 g.

The SuryeillaHee Pequeney is eentre r under the C. -

IZ-'

r'--i~'

An 12.

inAI SR 3.1.7.8,

9 49.1 I

Dem rating that each SLC System pump develops a flow rate

.4 gpm at a discharge pressure Ž 1275 psig ensures that pump performance has not degraded below design values during the fuel cycle.

is indicative of overall performance.

S inservice inspections confirm component OPERABILITY,/ 4nd performance, and detect incipient failures by ndicating abnormal performance.

in additi-n,

...../ *o f T-abl-e 3.1...

subsy-,emf.

The Frequency of this Surveillance is in acc dance with the Inservice Testing Program.

SR 3.1.7.9 This Surveillance ensures that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve.

The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV.

An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

PBAPS UNIT 2 B 3.1-46 Revision No.

109

SRVs and SVs B 3.4.3 BASES (continued)

APPLICABLE SAFETY ANALYSES The overpressure protection system must accommodate the most severe pressurization transient.

Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs),

followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref.

1).

For the purpose of the analyses, 4-s.RVs and SVs are assumed to operate in the safety mode.T he analysis results demonstrate that the design SR and SV capacity is capable of maintaining reactor pressure blow the ASME Code limit of 110% of vessel design pressure (11 x 1250 psig =

1375 psig).

This LCO helps to ensure hat the acceptance limit of 1375 psig is met during the De

• gn Basis Event.

From an overpressure standpoint, the design asis events are bounded by the MSIV closure with flux scram e nt described above.

Reference 2 discusses additional events hat are expected to actuate the SRVs and SVsa SRVs and SVs satisfy Criterion 3 of the NRC Policy Statement.

=

Although not a design basis event, the ATWS analysis demonstrates that peak vessel bottom pressure is less than the ASME Service Level C limit of 1,500 psig.

LCO The safety function of any combination of 4 SRVs and SVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs.

1 and 2).

Regarding the SRVs, the requirements of this LCO are applicable only to their capability to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety mode).

I The SRV and SV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied.

The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions.

The transient evaluations in the UFSAR are based on these setpoints, but also include the additional uncertainties of

+ 3% of the nominal setpoint to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

PBAPS UNIT 2 B 3.4-16 Revision No.

109

RHR Drywell Spray B 3.6.2.5 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.5 Residual Heat Removal (RHR)

Drywell Spray BASES BACKGROUND Delete and replace with Insert 3.6.2.5 A Drywell Spray is a mode of the RHR system which may be initiated under post accident conditions to reduce the temperature and pressure of the primary containment atmosphere.

The Drywell Spray function is credited in design basis analyses to limit peak drywell temperature following a steam line break inside of the Drywell and may be used to mitigate other loss of coolant accidents inside of the Drywell.

This function is provided by two redundant Drywell Spray subsystems.

The purpose of this LCO is to ensure that both subsystems are OPERABLE in applicable MODES.

The RHR System has twe leeps with each l..p consisting of tw.

m-tor driven pum..ps, two heat exchangers, and asso.iated piping and valv es.

There are two RHR Drywel1 spr.ay subsystemfs per RHR Systemf loop.

The four RHR drywell spray subsystemfs are mnanubally, initiated and indepenidently eent1l-ed-- The fur RHR drywell spray fe the drywe.l 1spray funtion by i reul ating water fro the suppression pool throuigh the RHR heat exchanger-s an discharging the ceoled suppressien peel water into the drywell air space through the drywell spray sparger and sp....

I........

T

.'I...

reduction through the combined effects of evapeoative and convective ceoling, depending on the drywell atmers-phe-..

44 V

F, y

11 V

preeess will ensue.

if the enyipenment in ýhe dr-ywell is saturated, tefflpepature and pressure will be redbieed via a eenveetive eeeling ppeeess.

Eaeh dr-ywell spray sparger line is eefflfflefl te the twe RR dpywell spFay subsysteffls in afl RHR Systeffl leap. 4*

reqbt4ped, a sFflall peFtien ef the spray ýIew ean be dir-eete4 te the suppressien peel spray spapgep and spPay nezzles.

High Ppessure Serviee Watep, eireulating thpeugh the tube side ef the heat e)ýehangers, exehanges heat with suppPessien peel watep en the she!! side ef the heat exehangeps and disehapges this heat te the extepnal heat sil9k.

(continued)

PBAPS UNIT 2 B 3.6-63a Revision No.

106 I

RHR Drywell Spray B 3.6.2.5 BASES (continued)

APPLICABLE ANALYSES Reference 2 contains the results of analyses used to SAFETY predict primary containment pressure and temperature response following a spectrum of small steam line break sizes.

Steam line breaks are the most limiting events for drywell temperature response, since steam has higher energy content than liquid.

These analyses, with primary focus on the drywell temperature response, take credit for containment sprays and structural heat sinks in the drywell and the suppression pool airspace.

These analyses demonstrate that, with credit for containment spray (drywell and suppression pool), drywell temperature is maintained within limits for Environmental Qualification (EQ) of equipment located in the drywell for the analyzed spectrum of small steam line breaks.

The RHR Drywell Spray System satisfies Criterion 3 of the NRC Policy Statement.

LCO Delete and replace with Insert 3.6.2.5 B In the event of a small steam line break in the dr.ywell, a mfinimfumf of one RHR drywell spray subsystemf is er-e ited in~

design analyses to miti.gate the rise in dr.ywell tem..perature and pressure

.aused.

by the steam line break, and to mfaintain the pr-imfar-y eontainmfent peak temfpe-ra-tbpe and pressure bel.w the design limnits (Ref. 2).

lte

.en.spe that these requirements are met, two RHR drywell spr" y subsystem.s (ne in ea.h l..p) must be OPERABLE with power fromf' two safety related independent powern supplies-.

two subsyst.ems mst be in separate loops sinee the drywell spray sparger line valves ape commofln to both subsystemfs in a lee-p.)

Therefore, in the event of an aeeident, at least one subsystemf is OPERABLE assumfing the worst ease single act~ive failuire.

An RHR-drzywell spr-ay suibsystem is.

OPERABLE when one of the pumfps, the associated hea4 exehanger, a HPSW4 Systemf pumfp capable of proeviding eaelng-to the heat exchanger and associated piping, valve-instrumfenltationl, and controls are OPERABLE-.

APPLICABILITY In MODES 1, 2, and 3, a steam line break in the drywell could cause a rise in primary containment temperature and pressure.

In MODES 4 and 5, the probability and consequences of steam line breaks are reduced due to the pressure and temperature limitations in these MODES.

Therefore, maintaining RHR drywell spray subsystems OPERABLE is not required in MODE 4 or 5.

(continued)

PBAPS UNIT 2 B 3.6-63b Revision No.

106 I

RHR Drywell Spray B 3.6.2.5 BASES (continued)

SURVEILLANCE SR 3.6.2.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR drywell spray mode flow path provides assurance that the proper flow paths will exist for system operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing.

A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis.

This is acceptable since the RHR drywell mode is manually initiated.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

NEDC-33566P - Safety Analysis Reportfor SR 3.6.2.5.2 Exelon Peach Bottom Station, This Surveillance is performed to verify that the spray Units 2 and 3, Constant nozzles are not obstructed and that flow will be provided Pressure Power Uprate.

when required.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1.

UFSAR, Sections 5.2 and 14.6.3.

2.

GE N4E 900 911 41.339 PrejOt Task R.p.r.t, Poech Bottom

..At..i. Pe......

ta unIts 2 and 3, SL 53 Eva! uati on.

SR 3.6.2.5.3 Verification of manual transfer between the normal and alternate power source (4kV emergency bus) for each RHR motor-operated flow control valve and each RHR cross-tie motor-operated valve demonstrates that AC power will be available to operate the required valves following loss of power to any single 4kV emergency bus. The ability to provide power to each RHR motor-operated flow control valve and each RHR cross-tie motor-operated valve from either of two independent 4kV emergency buses ensures that a single failure of a DG will not result in failure of the RHR motor-operated flow control valve and the RHR cross-tie motor-operated valve; therefore, failure of the manual transfer capability will result in inoperability of the associated RHR Drywell Spray subsystem. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

PBAPS UNIT 2 B 3.6-63d Revision No. 106 1

INSERT 3.6.2.5 A Each of the RHR drywell spray subsystems contains two motor driven pumps, two heat exchangers and a heat exchanger cross-tie line, which are manually initiated and independently controlled. The two RHR drywell spray subsystems perform the drywell spray function by circulating water from the suppression pool through the RHR heat exchangers and discharging the cooled suppression pool water into the drywell air space through the drywell spray sparger and spray nozzles. The spray then effects a temperature and pressure reduction through the combined effects of evaporative and convective cooling, depending on the drywell atmosphere. If the atmosphere is superheated, a rapid evaporative cooling process will ensue. If the environment in the drywell is saturated, temperature and pressure will be reduced via a convective cooling process.

Each drywell spray sparger line is supplied by one independent RHR drywell spray subsystem. If required, a small portion of the spray flow can be directed to the suppression pool spray sparger and spray nozzles. High Pressure Service Water, circulating through the tube side of the heat exchangers, exchanges heat with the suppression pool water on the shell side of the heat exchangers and discharges this heat to the external heat sink.

Each drywell spray subsystem is equipped with a RHR heat exchanger cross-tie line, located downstream of each RHR pump discharge and upstream of each heat exchanger inlet, which allows one RHR pump to be aligned to supply both RHR heat exchangers in the same subsystem to provide additional containment heat removal capability when only one RHR pump is available. The RHR heat exchanger cross-tie is normally closed, and is assumed in the design basis analyses to be placed in service one hour following a design basis accident or transient when insufficient electric power is available to operate two RHR pumps in a subsystem.

INSERT 3.6.2.5 B In the event of a small steam line break in the drywell, a minimum of one RHR drywell spray subsystem is credited in the design analyses to mitigate the rise in drywell temperature and pressure caused by the steam line break, and to maintain the primary containment peak temperature and pressure below the design limits (Ref. 2). To ensure that these requirements are met, two RHR drywell spray subsystems (one in each loop) must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR drywell spray subsystem is OPERABLE when one of the pumps, two heat exchangers in the same subsystem, the associated RHR heat exchanger cross-tie line, two HPSW System pumps capable of providing cooling to the two heat exchangers and associated piping, valves, instrumentation, and controls are OPERABLE.

SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. The rate of negative reactivity insertion is increased by using highly enriched boron in the SLC System solution that increases the rate of Boron-10 injection and functions to shutdown the reactor core faster.

This limits the heat generated that is transferred to the suppression pool during an ATWS event. Limiting the heat transferred to the suppression pool maintains the pool below design limits, which ensures adequate NPSH is available for the ECCS pumps without credit for containment accident pressure. This test confirms one point on the pump design curve and

.1.7

.13 7

(

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)

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~+~~+g~-The Surveillanec

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4- - -1 1 -4 A -

4-k -

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.J,'V*1 U. IIUI, p,' uen.eI*JI;G,.f" l,

U H 4,

i iJrJI.4..i; SR 3.1.7.8 I49.1 Demons ting that each SLC System pump develops a flow rate

Ž 4-3.-

gpm at a discharge pressure Ž 1275 psig ensures that pump performance has not degraded below design values during the fuel cycle. 'This test is indicative of overall performance.

uch inservice inspections confirm component OPERABIL

, trend performance, and detect incipient fa by indicating abnormal performance.

in additin-,

4-L 14.4--

.t T-k1-7 1

7 1

4-4

..,.I z..

I-C I cr subacstcorda.n accordance The Frequency of this Surveillance is in with the Inservice Testing Program.

SR 3.1.7.9 This Surveillance ensures that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve.

The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV.

An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

(continued)

PBAPS UNIT 3 B 3.1-46 Revision No.

109

SRVs and SVs B 3.4.3 BASES (continued)

APPLICABLE SAFETY ANALYSES The overpressure protection system must accommodate the most severe pressurization transient.

Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs),

followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref.

1).

For the purpose of the analyses, 4rg SRVs and SVs are assumed to operate in the safety mode.

he analysis results demonstrate that the design SR and SV capacity is capable of maintaining reactor pressure b low the ASME Code limit of 110% of vessel design pressure (11 x 1250 psig =

1375 psig).

This LCO helps to ensure hat the acceptance limit of 1375 psig is met during the De *gn Basis Event.

From an overpressure standpoint, the design asis events are bounded by the MSIV closure with flux scram e nt described above.

Reference 2 discusses additional events at are expected to actuate the SRVs and SVs SRVs and SVs satisfy Criterion 3 of the NRC Policy Statement.

Although not a design basis event, the ATWS analysis demonstrates that peak vessel bottom pressure is less than the ASME Service Level C limit of 1,500 psig.

LCO The safety function of any combination of 4-SRVs and SVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs.

1 and 2).

Regarding the SRVs, the requirements of this LCO are applicable only to their capability to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety mode).

The SRV and SV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied.

The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions.

The transient evaluations in the UFSAR are based on these setpoints, but also include the additional uncertainties of

+ 3% of the nominal setpoint to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

PBAPS UNIT 3 B 3.4-16 Revision No.

109

RHR Drywell Spray B 3.6.2.5 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.5 Residual Heat Removal (RHR)

Drywell Spray BASES BACKGROUND Delete and replace with Insert 3.6.2.5 A Drywell Spray is a mode of the RHR system which may be initiated under post accident conditions to reduce the temperature and pressure of the primary containment atmosphere.

The Drywell Spray function is credited in design basis analyses to limit peak drywell temperature following a steam line break inside of the Drywell and may be used to mitigate other loss of coolant accidents inside of the Drywell.

This function is provided by two redundant Drywell Spray subsystems.

The purpose of this LCO is to ensure that both subsystems are OPERABLE in applicable MODES.

The RHR System. has tw. loops with each lo.p cnsisting

.f two motor driven pumps, tw" heat exchangers, and asso.iat piping and valves.

There arc two RHR Drywell spra-y subsystems per RHR System loo.

The four R..R drywell spray subsystem..s are m

,anually, initiated and independ-e ntrolld.

The four. RHR drUywell spray subsyste-s perfoerm the drywel spray funtion by circu.lating water fro. the suppression pool through the RHR heat exhanger*

s nd discharging the cooled suppression pool water inte t-he dr ywell airi space through the dr1ywell spray sparger and sprIay I

I I

. J l 'l reduction through the combined effects of evaporative and convective cooling, depending en the drywell at es^pere.

q*

the atmolsphere is superheated, a rapil evaperativ-e e-ee-H-proces will ens.ue.

if the environm.ent in the drywell is saturated, tem..perature and pressure w4ill be redued via a convective cooling process.

Each dr-ywell spray sparger, line is commof,..n to the two RHR dr-ywell spray subsystemfs in an RHR Systemi loop.

4#f required, a small portion of the spray flow can be dipected to the suppression npeel spray sparger and spray nozzles.

High Pressurc Service Water, circublating through the tube side of the heat exchangers, exchanges heat with th suppression pool water en the shell side of the heat exchangers and discharges this heat to the external heat 5 i 19 1.

(continued)

PBAPS UNIT 3 B 3.6-63a Revision No.

106 I

RHR Drywell Spray B 3.6.2.5 BASES (continued)

APPLICABLE ANALYSES Reference 2 contains the results of analyses used to SAFETY predict primary containment pressure and temperature response following a spectrum of small steam line break sizes.

Steam line breaks are the most limiting events for drywell temperature response, since steam has higher energy content than liquid.

These analyses, with primary focus on the drywell temperature response, take credit for containment sprays and structural heat sinks in the drywell and the suppression pool airspace.

These analyses demonstrate that, with credit for containment spray (drywell and suppression pool),

drywell temperature is maintained within limits for Environmental Qualification (EQ) of equipment located in the drywell for the analyzed spectrum of small steam line breaks.

The RHR Drywell Spray System satisfies Criterion 3 of the NRC Policy Statement.

LCO Delete and replace with Insert 3,6.2.5 B in the event of a smiiall steam.

line brealk in the dr.ywell, a mfinimfumf ef one RU1R dr-ywell spray subsystemf is eredited in atopru and prtssure

.aus.d by the stam lire br.a.,

and to maintain the primary ontainmc.nt peak tem..peraturi.e and pr.ssur. be,,w the design limlts (Ref.

2).

T,... r-.e that these requieents are met, two RUR dr.yw.ll ra.y subsysteIs (one in eah lop) mflust be OPERABLE with power fromf two safety related independent power supplies--. 444e, tW8 subsystemfs mfust be in separate loops since the dr-ywell-spray sparger-line valves are commofln to both subsy-s-tems in al -eep)

Therefore, in the event of an a..id.nt, at least

.n. subsystem is OPERABLE assuming the wor.st ease singl aetive failure.

An IR

'Rdr-ywell spray subsyste-is OP[RABLE when one of the pumfps, the associated heat exehanger, a HPGW Systemf pumfp capable of providing eaelin to the heat cx.hanger and a seiat ed piping, va-lv-es-instr.um.entation, and controls ar.

e OPERAB APPLICABILITY In MODES 1, 2, and 3, a steam line break in the drywell could cause a rise in primary containment temperature and pressure.

In MODES 4 and 5, the probability and consequences of steam line breaks are reduced due to the pressure and temperature limitations in these MODES.

Therefore, maintaining RHR drywell spray subsystems OPERABLE is not required in MODE 4 or 5.

(continued)

PBAPS UNIT 3 B 3.6-63b Revision No.

106 1

RHR Drywell Spray B 3.6.2.5 BASES (continued)

SURVEILLANCE SR 3.6.2.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR drywell spray mode flow path provides assurance that the proper flow paths will exist for system operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing.

A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis.

This is acceptable since the RHR drywell mode is manually initiated.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

NEDC-33566P - Safety Analysis Reportfor SR 3.6.2.5.2 Exelon Peach BottomStation, This Surveillance is performed to verify that the spray Units 2 and 3, Constant nozzles are not obstructed and that flow will be provided Pressure PowerUprate.

when required.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES

1.

UFSAR, Sections 5.2 and 14.6.3.

N, Ato ic,"e

,,e e S-......., Units 2 a,,d 3, SIL 636 Eva uaIi eig.-

SR 3.6.2.5.3 Verification of manual transfer between the normal and alternate power source (4kV emergency bus) for each RHR motor-operated flow control valve and each RHR cross-tie motor-operated valve demonstrates that AC power will be available to operate the required valves following loss of power to any single 4kV emergency bus. The ability to provide power to each RHR motor-operated flow control valve and each RHR cross-tie motor-operated valve from either of two independent 4kV emergency buses ensures that a single failure of a DG will not result in failure of the RHR motor-operated flow control valve and the RHR cross-tie motor-operated valve; therefore, failure of the manual transfer capability will result in inoperability of the associated RHR Drywell Spray subsystem. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

PBAPS UNIT 3 B 3.6-63d Revision No. 106 1

INSERT 3.6.2.5 A Each of the RHR drywell spray subsystems contains two motor driven pumps, two heat exchangers and a heat exchanger cross-tie line, which are manually initiated and independently controlled. The two RHR drywell spray subsystems perform the drywell spray function by circulating water from the suppression pool through the RHR heat exchangers and discharging the cooled suppression pool water into the drywell air space through the drywell spray sparger and spray nozzles. The spray then effects a temperature and pressure reduction through the combined effects of evaporative and convective cooling, depending on the drywell atmosphere. If the atmosphere is superheated, a rapid evaporative cooling process will ensue. If the environment in the drywell is saturated, temperature and pressure will be reduced via a convective cooling process.

Each drywell spray sparger line is supplied by one independent RHR drywell spray subsystem. If required, a small portion of the spray flow can be directed to the suppression pool spray sparger and spray nozzles. High Pressure Service Water, circulating through the tube side of the heat exchangers, exchanges heat with the suppression pool water on the shell side of the heat exchangers and discharges this heat to the external heat sink.

Each drywell spray subsystem is equipped with a RHR heat exchanger cross-tie line, located downstream of each RHR pump discharge and upstream of each heat exchanger inlet, which allows one RHR pump to be aligned to supply both RHR heat exchangers in the same subsystem to provide additional containment heat removal capability when only one RHR pump is available. The RHR heat exchanger cross-tie is normally closed, and is assumed in the design basis analyses to be placed in service one hour following a design basis accident or transient when insufficient electric power is available to operate two RHR pumps in a subsystem.

INSERT 3.6.2.5 B In the event of a small steam line break in the drywell, a minimum of one RHR drywell spray subsystem is credited in the design analyses to mitigate the rise in drywell temperature and pressure caused by the steam line break, and to maintain the primary containment peak temperature and pressure below the design limits (Ref. 2). To ensure that these requirements are met, two RHR drywell spray subsystems (one in each loop) must be OPERABLE with power from two safety related independent power supplies. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure. An RHR drywell spray subsystem is OPERABLE when one of the pumps, two heat exchangers in the same subsystem, the associated RHR heat exchanger cross-tie line, two HPSW System pumps capable of providing cooling to the two heat exchangers and associated piping, valves, instrumentation, and controls are OPERABLE.