Text

PVNGS Insertion Instructions for Technical Specifications Bases Changed Pages, Revision 54
	ML110610762
Person / Time
Site:	Palo Verde
Issue date:	01/26/2011
From:	Stephenson C; Arizona Public Service Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML110610762 (73)
	v • d • e

Insertion Instructions for Technical Specifications Bases Revision54 Technical Specification Bases Revision 54 incorporates LDCRs 09-B016, 09-B020, 09-B021,08-B013 and 10-B013.

LDCR 09-B016 relates to License Amendment le 1 (Administrative Relocation of MSIV and MFIV Stroke Times).

LDCR 09-B020 relates to License Amendment 1 2 (RWT Level Setpoint Change).

LDCR 09-B021 relates to License Amendment 1 3 (Organization Title Changes and Deletion of Redundant Safety Limit Reportin I).

LDCR 08-B013 relates to License Amendment 1 4 (Replacement of ASME Section Xl Code with OM Code).

LDCR 10-B013 clarifies the design basis for the ;ondensate Storage Tank (CST).

REMOVE PAGES INSERT PAGES Cover page Cover p_Lge List of Effective Pages List of Elfective Pages 1/2 through 7/8 1/2 throLgh 7/8 B 2.1.1-5 / Blank B 2.1.1-_ / Blank B 2.1.2-3 / B 2.1.2-4 B 2.1.2-: / B 2.1.2-4 B 2.1.2-5 / Blank (No Rep acement Page)

B 3.3.5-9 / B 3.3.5-10 B 3.3.5-!/B 3.3.5-10 B 3.3.5-11 / B 3.3.5-12 B 3.3.5-' 1 / B 3.3.5-12 B 3.3.5-17 / B 3.3.5-18 B 3.3.5-'17/ B 3.3.5-18 B 3.3.5-19 / B 3.3.5-20 B 3.3.5-'19/ B 3.3.5-20 B 3.4.7-3 / B 3.4.7-4 B 3.4.7-: / B 3.4.7-4 B 3.4.8-1 / B 3.4.8-2 B 3.4.8- / B 3.4.8-2 B 3.4.10-3 / B 3.4.10-4 B 3.4.1G3 / B 3.4.10-4 B 3.4.11-5 / B 3.4.11-6 B 3.4.1 -5 / B 3.4.11-6 B 3.4.13-9 / B 3.4.13-10 B 3.4.1:-9 / B 3.4.13-10 B 3.4.13-11 / Blank B 3.4.12-11 / Blank B 3.4.15-5 / B 3.4.15-6 B 3.4.1_-5 / B 3.4.15-6 B 3.4.15-7 / Blank B 3.4.1 -7 / Blank Stephenson, Digitally signed by Stephenson, Carl J(Z05778)

DN: cn=Stephenson, Carl J(Z05778)

Carl J(Z05778)

Reason: I attest to the accuracy and integrity of this document Date: 2011.01.21 15:48:29 -07'00'

Insertion Instructions for Technical SF,ecificationsBases Revision 54 B 3.5.3-9 / B 3.5.3-10 B 3.5.3-! / B 3.5.3-10 B 3.5.5-1 / B 3.5.5-2 B 3.5.5-' / B 3.5.5-2 B 3.5.5-3 / B 3.5.5-4 B 3.5.5-,! / B 3.5.5-4 B 3.6.6-7 / B 3.6.6-8 B 3.6.6-} / B 3.6.6-8 B 3.6.6-9 / Blank B 3.6.6-,,./ Blank B 3.7.1-5 / B 3.7.1-6 B 3.7.1-_ / B 3.7.1-6 B 3.7.2-7 / B 3.7.2-8 B 3.7.2-} / B 3.7.2-8 B 3.7.2-9 / Blank B 3.7.2-C

, / Blank B 3.7.3-5 / Blank B 3.7.3-,,./ Blank B 3.7.5-9 / B 3.7.5-10 B 3.7.5-C./ B 3.7.5-10 B 3.7.5-11 / Blank B 3.7.5-11 / Blank B 3.7.6-1 / B 3.7.6-2 B 3.7.6-1 / B 3.7.6-2 B 3.7.6-3 / B 3.7.6-4 B 3.7.6-,_/ B 3.7.6-4 B 3.8.3-5 / B 3.8.3-6 B 3.8.3-,_ / B 3.8.3-6 B 3.8.3-9 / B 3.8.3-10 B 3.8.3-,c / B 3.8.3-10 B 3.9.4-1 / B 3.9.4-2 B 3.9.4-" / B 3.9.4-2 B 3.9.5-1 / B 3.9.5-2 B 3.9.5-' / B 3.9.5-2

PVNGS Palo Verde Nuclear Genera!ing Station Units 1, 2, and 3 Technic Specification Bases Revision 54 January 26, 2011 o,_,-C*e"

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DN:cn=Stephenson,CarlJ(Z0577_

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Reason:I attest to the accurao/an I integrity of Carl J(Z05778) Date:2011.01,2110:54:22

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PALO VERDE UNITS i, 2, AND 3 8 Revision 54 January 26, 2011

Reactor Core SLs B 2.1,1 BASES REFERENCES 1. 10 CFR50, Appendix A, GDC1[ 1988.

2. UFSAR,Sections 6 and 15.

PALOVERDEUNITS 1,2,3 B 2.1.1-5 REVISION54

This page intentionally blank RCSPressure SL B 2.1.2 BASES APPLICABILITY SL 2.1.2 applies in MODES1, 2, 3, 4, and 5 because this SL could be approached or exceededir these MODES due to overpressurization events. The SL is not applicable in MODE6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCScan be pressurized.

SAFETYLIMIT The following SL violation responses are applicable to the VIOLATIONS RCSpressure SLs.

2.2.2.1 If the RCSpressure SL is violatec when the reactor is in MODE1 or 2, the requirement is tc restore compliance and be in MODE3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With RCSpressure greater than the value specified in SL 2.1.2 in MODE1 or 2, the pressure must be reduced to below this value. A pressure greater that the value specified in SL 2.1.2 exceeds 110_ of the RCSdesign pressure and may challenge system integrity.

The allowed Completion Time of 1 _our provides the operator time to complete the necessary actions to reduce RCS pressure by terminating the cause of the pressure increase, removing mass or energy from the FCS, or a combination of these actions, and to establish MCDE3 conditions.

2.2.2.2 If the RCSpressure SL is exceedecin MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes.

Exceeding the RCSpressure SL in _ODE3, 4, or 5 is potentially more severe than exceeding this SL in MODEI or 2, since the reactor vessel tenDerature may be lower and the vessel material, consequently less ductile. As such, pressure must be reduced to less man the SL within 5 minutes. This action does not equire reducing MODES, since this would require reducing temperature, which would (continued)

PALOVERDEUNITS 1,2,3 B 2.1.2-3 REVISION0

RCSPressure SL B 2.1.2 BASES SAFETYLIMIT 2.2.2.2 (continued)

VIOLATIONS compoundthe problem by adding thermal gradient stresses to i the existing pressure stress.

REFERENCES 1. 10 CFR50, AppendixA, GDC14, GDC15, and GDC28.

2. ASME,Boiler and Pressure Vessel Code, Section Ill, Article NB-7000.

3. ASME,Boiler and Pressure Vessel Code,Section XI, Article IWX-5000.

4. 10 CFR100.

I 5. UFSAR,Section 7.

I PALOVERDEUNITS 1,2,3 B 2.1.2-4 REVISION54

ESFAS Instrumentation B 3.3,5 BASES APPLICABLE 4. Main Steam Isolation Signal continued)

SAFETYANALYSES (continued) high level condition or if a high containment pressure condition exists. This prevents an excessive rate of heat extraction and subsequent cooldown of the RCS during these events.

5. Recirculation Actuation Signal At the end of the injection _hase of a LOCA,the Refueling Water Tank (RWT)will be nearly empty Continued cooling must be provided by the ECCSto remove decay heat. The source of water for the ECCS pumps is automatically switched to the containment recirculation sump. Switchover from RWTto containment sump must occur before the RWTempties to prevent damageto the ECCSpamps and a loss of core cooling capability. For similar reasons, switchover must not occur before there is sufficient water in the containment sump to support pump suction. Furthermore, early switchover must not octur to ensure sufficient borated water is injected from the RWTto ensure the reactor remains shut down in the recirculation mode.

An RWTLevel - Low signal initiates the RAS. Once a RAShas occurred, timely operator action is required to close the RWTisolation valves (CH-531 and CH-530) to preclude air entrainment in the suction from the RWTduring switchover to recirculation. The volume remaining in the RWTafter the RAS provides enough time for this operator actian and closure of the valves.

6, 7. Auxiliary Feedwater Actuation Signal AFASconsists of two steam generator (SG) specific signals (AFAS-1 and AFAS-2). AFAS-1 initiates auxiliary feed to SG#1, and AFAS-2 initiates auxiliary feed to SG#2.

AFASmaintains a steam generator heat sink during a steam generator tube rupture event and an MSLBor FWLB event either inside or outside containment.

Low steam generator water l vel initiates auxiliary feed to the affected steam enerator, providing the generator is not identified (by the rupture detection circuitry) as faulted (a steam or FWLB).

(continued)

PALOVERDE UNITS1,2,3 B 3.3.5-9 REVISION 54

ESFASInstrumentati on B 3.3.5 BASES APPLICABLE 6, 7. Auxiliary FeedwaterActuation Signal SAFETY ANALYSES (continued) AFASlogic includes steam generator specific inputs from the SGPressure Difference - High (SG#1 > SG#2 or SG#2 > SG#i, bistable comparators) to determine if a fault in either generator has occurred.

Not feeding a faulted generator prevents containment overpressurization during the analyzed events.

The ESFASsatisfies Criterion 3 of 10 CFR50.36 (c)(2)(ii).

LCO The LCOrequires all channel componentsnecessary to provide an ESFASactuation to be OPERABLE.

The Basesfor the LCOson ESFASFunctions are:

1. Safety Injection Actuation Signal

a. ContainmentPressure - High This LCOrequires four channels of Containment Pressure - High to be OPERABLE in MODES 1, 2 and 3.

The Containment Pressure- High signal is shared amongthe SIAS (Function 1), CIAS (Function 3),

and MSIS(Function 4).

The Allowable Value for this trip is set high enoughto allow for small pressure increases in containment expected during normal operation (i.e., plant heatup) and is not indicative of an abnormal condition. The setting is low enoughto initiate the ESFFunctions when an abnormal condition is indicated. This allows the ESF systems to perform as expected in the accident analyses to mitigate the consequencesof the analyzed accidents.

(continued)

PALOVERDEUNITS 1,2,3 B 3.3.5-10 REVISION54

ESFASInstrumentati on B 3.3.5 BASES LCO b. Pressurizer Pressure-Low (continued)

This LCOrequires four channels of Pressurizer Pressure - Low to be OPERABLin MODES1, 2 and 3.

The Allowable Value for this trip is set low enough to prevent actuating the ESF Functions (SIAS and CIAS) during normal plant operation and pressurizer pressure transients. The setting is high enough that, with the s)ecified accidents, the ESF systems will actuate to perform as expected, mitigating the consequences of the accident.

The Pressurizer Pressure - Low trip setpoint, which provides SIAS, CIAS, and RPStrip, may be manually decreased to a floor value of 100 psia to allow for a controlled cooldown and depressurization of the RCS_ithout causing a reactor trip, CIAS, or SIAS. The margin between actual pressurizer pressure and the trip setpoint must be maintained less than or equal to the specified value (400 psia) to ensure a reactor trip, CIAS, and SIAS will occur if required during RCScooldown and depressurization. When the RCScold leg temperature is _ 485°F the setpoint must be _ 140 psia greater than the saturation pressure of the RCScold leg. This is required to ensure a SIAS prior to reactor vessel upper head void formation in the event of RCS depressurization caused by a steam line break.

From this reduced setting, the trip setpoint will increase automatically as pressurizer pressure increases, tracking actual RCSpressure until the trip setpoint is reached.

Whenthe trip setpoint has been lowered below the bypass permissive setpoint cf 400 psia, the Pressurizer Pressure - Low reactor trip, CIAS, and SIAS actuation may be manually bypassed in preparation for shutdown cocling. When RCS pressure rises above the by ass removal setpoint, the bypass is removed.

(conti nued)

PALOVERDEUNITS 1,2,3 B 3.3.5-11 REVISION54

ESFASInstrumentati on B 3.3.5 BASES LCO Bypass Removal (continued)

This LCOrequires four channels of operating bypass removal for Pressurizer Pressure-Low to be OPERABLE in MODES1, 2 and 3.

Each of the four channels enables and disables the operating bypass capability for a single channel. Therefore, this LCOapplies to the operating bypass removal feature only. If the bypass enable function is failed so as to prevent entering an operating bypass condition, operation I may continue.

Because the trip setpoint has a floor value of 100 psia, a channel trip will result if pressure is decreased below this setpoint without bypassing.

The operating bypass removal Allowable Value was chosen because MSLBevents originating from below this setpoint add less positive reactivity than that which can be compensated for by required SDM.

2. Containment Spray Actuation Signal

a. Containment Pressure - High High This LCOrequires four channels of Containment Pressure - High High to be OPERABLE in MODES1, 2, and 3.

The Allowable Value for this trip is set high enough to allow for small pressure increases in containment expected during normal operation (i.e. plant heatup) and is not indicative of an abnormal condition. The setting is low enough to initiate CSASin time to prevent containment pressure from exceeding design.

(continued)

PALOVERDE UNITS1,2,3 B 3.3.5-12 REVISION1

ESFASInstrumentati on B 3,3.5 BASES LCO conservative than the UFS_RTrip Setpoint. The (continued) general relationship amon_ the PVNGStrip setpoint terms is as follows. The _alculated limiting setpoint (LSp) is determined within the plant specific setpoint analysi and is based on the Analytical Limit and Tota Loop Uncertainty. The UFSARTrip Setpoint is eq al to or more conservative than the LSp and is specified in the UFSAR. The DSp is the field installed setting and is equal to or more conservative than th_ UFSARTrip Setpoint.

This relationship ensures that sufficient margin to the safety and/or analyti:al limit is maintained.

If the as-found instrument setting is found to be non-conservative with respect to the AV specified in the technical specificatims, or the as-left instrument setting cannot be returned to a setting within the ALT, or the instrument is not functioning as required; then the instrument channel shall be declared inoperable.

b. Containment Pressure - High This LCOrequires four channels of Containment Pressure - High to be OPERABLE in MODES1, 2 and

3. The Containment Pressure - High signal is shared amongthe SIAS (Function 1), CIAS (Function 3), and MSIS (Fmction 4).

The Allowable Value for t_is trip is set high enough to allow for small pressure increases in containment expected duri ig normal operation (i.e., plant heatup) and s not indicative of an abnormal condition. The ;etting is low enough to initiate the ESF Functions when an abnormal condition is indicated. This allows the ESF systems to perform as expected in the accident analyses to mitigate the =onsequences of the analyzed accidents.

c. Steam Generator Level-Hig_

This LCOrequires four channels of Steam Generator Level-High to b OPERABLE in MODES1, 2 and 3.

The allowable value for t is trip is set high enough to ensure it does _ot interfere with (continued)

PALOVERDEUNITS 1,2,3 B 3.3.5-17 REVISION35

ESFASInstrumentati on B 3.3.5 BASES LCO c. Steam Generator Level-High (continued) normal plant operation. The setting is low enough to prevent moisture damage to secondary plant components in the case of a steam generator overfill event.

5. Recirculation Actuation Signal

a. Refueling Water Tank Level - Low This LCOrequires four channels of RWTLevel - Low to be OPERABLE in MODES1, 2, and 3.

The upper limit on the Allowable Value for this trip is set low enough to ensure RASdoes not initiate before sufficient water is transferred to the containment sump.

Premature recirculation could impair the reactivity control function of safety injection by limiting the amount of boron injection.

Premature recirculation could also damage or disable the recirculation system if recirculation begins before the sump has enough water to prevent air entrainment in the suction.

The lower limit on the RWTLevel - Low trip Allowable Value is high enough to transfer suction to the containment sump prior to emptying the RWT. Once a RAShas occurred timely operator action is required to close the RWTisolation valves (CH-531 and CH-530) to preclude air entrainment in the suction from the RWTduring switchover to recirculation. The volume remaining in the RWTafter the RASprovides enough time for I this operator action and closure of the valves.

6, 7. Auxiliary Feedwater Actuation Signal SG#1 and SG #2 (AFAS-1 and AFAS-2)

AFAS-1 is initiated to SG #1 by either a low steam generator level coincident with no differential pressure trip present or by a low steam generator level coincident with a differential pressure between the two generators with the higher pressure in SG #1.

AFAS-2 is similarly configured to feed SG #2.

I (continued)

PALOVERDEUNITS 1,2,3 B 3.3.5-18 REVISION54

ESFASInstrumentati on B 3.3.5 BASES LCO 6, 7. Auxiliar 7 Feedwater Actuation _ignal SG #1 and SG#2 continued) (AFAS-1andAFAS-2)

The steam generator secondary lifferential pressure is used, as an input of the AFAS logic where it is used to determine if a generator is intact. The AFAS logic inhibits feeding a steam generator if the pressure in that steam generator is less tilan the pressure in the other steam generator by the S;eam Generator Pressure Difference (SGPD)- High setpoint.

The SGPDsetpoint is high enou,fh to allow for small pressure differences and norma instrumentation errors between the steam generator chLnnels during normal operati on.

The fol lowing LCOdescription _ppl i es to both AFAS signals.

a. Steam Generator Level - Low This LCO requires four channels of Steam Generator Level - Low to be OPERABLE for each AFAS in MODES1, 2, and 3.

The Steam Generator Level - Low AFAS input is shared with the SteamGenerator Level-Low RPS function. The Steam Generator Level-Low AFASand RPSuse separate bistables. This allows the AFAS setpoint to be set lower than the RPSsetpoint.

The allowable value is high enough to ensure the steam generator is available as a heat sink. The setting is low enough to prevent inadvertent AFAS actuations during plant transients. This setpoint provides allowarce that there will be sufficient inventory in the steam generator at the time of the RPStrip to provide a margin of at least 10 minutes before auxiliary feedwater is required to prevent degraded core cooling.

b. SG Pressure Difference - High (SG #1 > SG #2) or (SG #2 > SG#1)

This LCOrequires four c_annels of SG Pressure Difference - High to be PERABLEfor each AFAS in MODES1, 2, and 3.

(continued)

PALOVERDEUNITS 1,2,3 B 3.3.5-19 REVISION54

ESFASInstrumentati on B 3.3.5 BASES LCO b. SG Pressure Difference-High (SG #1 > SG#2) or (continued) (S'G#2 > SG _1)

The Allowable Value for this trip is high enough to allow for small pressure differences and normal instrumentation errors between the steam generator channels during normal operation without an actuation. The setting is low enough to detect and inhibit feeding of a faulted (MSLB or FWLB)steam generator in the event of an MSLB or FWLB,while permitting the feeding of the intact steam generator.

APPLICABILITY In MODES1, 2 and 3 there is sufficient energy in the primary and secondary systems to warrant automatic ESF System responses to:

Close the main steam isolation valves to preclude a positive reactivity addition;

Actuate auxiliary feedwater to preclude the loss of the steam generators as a heat sink (in the event the normal feedwater system is not available);

Actuate ESF systems to prevent or limit the release of fission product radioactivity to the environment by isolating containment and limiting the containment pressure from exceeding the containment design pressure during a design basis LOCAor MSLB; and

Actuate ESF systems to ensure sufficient borated water inventory to permit adequate core cooling and reactivity control during a design basis LOCAor MSLB accident.

In MODES4, 5 and 6 automatic actuation of these Functions is not required because adequate time is available to evaluate plant conditions and respond by manually operating the ESFcomponents if required, as addressed by LCO3.3.6.

Several trips have operating bypasses, discussed in the preceding LCOsection. The interlocks that allow these bypasses shall be OPERABLE whenever the RPSFunction they support is OPERABLE.

(continued)

PALOVERDEUNITS 1,2,3 B 3.3.5-20 REVISION54

RCSLoops- MODE 5, LoopsFilled B 3.4.7 BASES LCO in order to use the provisions of the Note allowing the (continued) pumpsto be de-energized. In this MODE, the SG(s)can be used as the backup for SDCheat re_oval. To ensure their availability, the RCSloop flow path is to be maintained with subcooled liquid.

In MODE5, it is sometimes necessary to stop all RCPor SDC forced circulation. This is permitted to change operation from one SDCtrain to the other, perform surveillance or startup testing, perform the transition to and from the SDC, or to avoid operation below the RCPminimum net positive suction head limit. The time period is acceptable because natural circulation is acceptable for decay heat removal the reactor coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected.

Note 2 allows one SDCtrain to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided that the other SDCtrain is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable train during the only time when such testing is safe and possible.

Note 3 requires that secondary side water temperature in each SG is < IO0°F above each of t_e RCScold leg temperatures before an RCPmay be started with any RCScold leg temperature less than or equal to the LTOPenable temperature specified in the PTLR Satisfying the above condition wi l preclude a low temperature overpressure event du to a thermal transient whenthe RCPis started.

Note 4 restricts RCPoperation to no more than 2 RCPswith RCScold leg temperature _ 200°F, and no more than 3 RCPs with RCScold leg temperature > 2[O°F but s 500°F.

Satisfying these conditions will naintain the analysis assumptions of the flow induced pressure correction factors due to RCPoperation (Ref. 3).

(continued)

PALOVERDEUNITS 1,2,3 B 3.4.7-3 REVISION52

RCSLoops - MODE 5, Loops Filled B 3.4.7 BASES LCO Note 5 provides for an orderly transition from MODE5 to (continued) MODE4 during a planned heatup by permitting removal of SDC trains from operation when at least one RCPis in operation.

This Note provides for the transition to MODE4 where an RCP is permitted to be in operation and replaces the RCS circulation function provided by the SDCtrains.

An OPERABLE SDCtrain is composedof an OPERABLE SDCpump (CS or LPSI) capable of providing flow to the SDCheat exchanger for heat removal.

SDCpumpsare OPERABLE if they are capable of being powered and are able to provide flow, if required. A SG can perform as a heat sink when it is OPERABLE and has the minimum water level specified in SR 3.4.7.2.

The RCSloops may not be considered filled until two conditions needed for operation of the steam generators are met. First, the RCSmust be intact. This means that all removable portions of the primary pressure boundary (e.g.,

manways, safety valves) are securely fastened. Nozzle dams are removed. All manual drain and vent valves are closed, and any open system penetrations (e.g., letdown, reactor head vents) are capable of remote closure from the control room. An intact primary allows the system to be pressurized as needed to achieve the subcooling margin necessary to establish natural circulation cooling. Whenthe RCSis not intact as described, a loss of SDCflow results in blowdown of coolant through boundary openings that also could prevent adequate natural circulation between the core and steam generators. Secondly, the concentration of dissolved or otherwise entrained gases in the coolant must be limited or other controls established so that gases coming out of solution in the SG U-tubes will not adversely affect natural circulation. With these conditions met, the SGs are a functional method of RCSheat removal upon loss of the operating SDCtrain. The ability to feed and steam SGs at all times is not required when RCStemperature is less than 210°F because significant loss of SG inventory through boiling will not occur during time anticipated to take corrective action. The required SG level provides sufficient time to either restore the SDCtrain or implement a method for feeding and steaming the SGs (using non-class components if necessary).

(continued)

PALOVERDE UNITS1,2,3 B 3.4.7-4 REVISION 54

RCSLoops - MODE5, Loops Not Filled B 3.4.8 B 3.4 REACTORCOOLANT SYSTEM(RCS)

B 3.4.8 RCSLoops - MODE5, Loops Not Filled BASES BACKGROUND In MODE5 with the RCSloops not flled, the primary function of the reactor coolant is the removal of decay heat and transfer of this heat to the Slutdown Cooling (SDC) heat exchangers. The Steam Generators (SGs) are not available as a heat sink when the loops are not filled. The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid.

In MODE5 with loops not filled, only the SDCSystem can be used for coolant circulation. The number of trains in operation can vary to suit the operational needs. The intent of this LCOis to provide forced flow from at least one SDCtrain for decay heat removal and transport and to require that two paths be available to provide redundancy for heat removal.

APPLICABLE In MODE5, RCScirculation is considered in determining SAFETYANALYSES the time available for mitigation of the accidental boron dilution event. The SDCtrains provide this circulation.

The flow provided by one SDCtrain is adequate for decay heat removal and for boron mixing.

RCSloops - MODE5 (loops not filled) have been identified in 10 CFR 50.36 (c)(2)(ii) as important contributors to risk reduction.

LCO The purpose of this LCOis to reqblre a minimumof two SDC trains be OPERABLE and one of these trains be in operation.

An OPERABLE train is one that is apable of transferring heat from the reactor coolant at controlled rate. Heat cannot be removedvia the SDCSys emunless forced flow is used. A minimumof one running SIC pumpmeets the LCO requirement for one train in operation. An additional SDC train is required to be OPERABLE o meet the single failure criterion.

(continued)

PALOVERDEUNITS 1,2,3 B 3.4.8-1 REVISION0

RCSLoops - MODE5, Loops Not Filled B 3.4.8 BASES LCO Note 1 permits all SDCpumpsto be de-energized _ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per (continued) 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The circumstances for stopping both SDCpumps are to be limited to situations whenthe outage time is short and the core outlet temperature is maintained > IO°F below saturation temperature. The 10 degrees F is considered the actual value of the necessary difference betweenRCScore outlet temperature and the saturation temperature associated with RCSpressure to be maintained during the time the pumps would be de-energized. The instrument error associated with determining this difference is less than 10 degrees F. (There are no special restrictions for instrumentation use.)

Therefore, the indicated value of the difference betweenRCS core outlet temperature and the saturation temperature associated with RCSpressure must be greater than or equal to 20 degrees F in order to use the provisions of the Note allowing the pumpsto be de-energized. The Note prohibits boron dilution or draining operations whenSDCforced flow is stopped.

Note 2 allows one SDCtrain to be inoperable for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided that the other train is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable train during the only time when these tests are safe and possible.

An OPERABLE SDCtrain is composedof an OPERABLE SDCpump (CS or LPSl) capable of providing flow to the SDCheat exchanger for heat removal. SDCpumpsare OPERABLE if they are capable of being poweredand are able to provide flow, I if required.

APPLICABILITY In MODE5 with loops not filled, this LCOrequires core heat removal and coolant circulation by the SDCSystem.

Operation in other MODESis covered by:

LCO3.4.4, "RCSLoops-MODES1 and 2";

LCO3.4.5, "RCSLoops - MODE3";

LCO3.4.6, "RCSLoops - MODE4";

LCO3.4.7, "RCSLoops - MODE5, Loops Filled";

LCO3.9.4, "Shutdown Cooling (SDC) and Coolant Circulation - High Water Level" (MODE6); and LCO3.9.5, "Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level" (MODE6).

(conti nued)

PALOVERDEUNITS1,2,3 B 3.4.8-2 REVISION54

Pressurizer Safety Valves-MODES1, 2, and 3 B 3,4.10 BASES APPLICABILITY The requirements for overpressure )rotection in other MODES (continued) are coveredby LCO3.4.11, "Pressurizer Safety Valves-MODE4," and LCO3.4.13, "LTOP SysLem."

The Note allows entry into MODES 3 and 4 with the lift settings outside the LCOlimits. [his permits testing and examination of the safety valves aL high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that Lhe valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months exception is based on 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months outage time for each of the four valves.

The 18 hour2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months period is derived from operating experience that hot testing can be performed withiq this timeframe.

ACTIONS A.I With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS overpressure protection system. An inoperable safety valve coincident with an RCSoverpressure event could challenge the integrity of the RCPB.

B.1 and B.2 If the Required Action cannot be _et within the required Completion Time or if two or more pressurizer safety valves are inoperable, the plant must be brought to a MODEin which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed is reasonable, based on operating experience, to reach MODE3 from full power without challenging plant systems.

Similarly, the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed is reasonable, based on operating experience, to reach MOEE4 without challenging plant systems.

(continued)

PALOVERDEUNITS 1,2,3 B 3.4.10-3 REVISION0

Pressurizer Safety Valves-MODES1, 2, and 3 B 3.4.10 BASES ACTIONS B.1 and B.2 (continued)

The change from MODE1, 2, or 3 to MODE4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by four pressurizer safety valves.

SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the Inservice Testing Program.

Pressurizer safety valves are to be tested in accordance I with the requirements of the ASMEOMCode (Ref. 3), which provides the activities and the Frequencynecessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valve setpoint is +3%, 1%for OPERABILITY; however, the valves are reset to +/- 1%during the Surveillance to allow for drift (Ref. 2). The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

I REFERENCES 1. ASME,Boiler and Pressure Vessel Code, Section Ill.

2. PVNGSOperating License AmendmentNos. 75, 61, and 47 for Units 1, 2, and 3, respectively, and associated NRCSafety Evaluation dated May 16, 1994.

3. ASMECode for Operation and Maintenance of Nuclear Power Plants.

PALOVERDEUNITS 1,2,3 B 3.4.10-4 REVISION54

PressLri zer Safety Valves-MODE 4 B 3.4.11 BASES (continued)

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS SRs are specified in the Inservice Testing Program.

Pressurizer safety valves are to be tested in accordance with the requirements of the ASMEOMCode (Ref. 2), which provides the activities and the Frequency necessary to satisfy the SRs. No additional requirements are specified.

The pressurizer safety valve setpcint is +3%, -1% for OPERABILITY however the Surveillance the valves to allow for dri.t_re (Ref.

reset 3).

to +/-The 1%during lift setting pressure shall correspond to ambient conditions of the valve at nominal operating te_)erature and pressure.

SR 3.4.11.2 SR 3.4.11.2 requires that the reqL red Shutdown Cooling System suction line relief valve OPERABLE by verifying its open pathway condition either

a. Once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a vave that is unlocked, not sealed, or otherwise not secllred open in the vent pathway, or

b. Once every 31 days for a val_e that is locked, sealed or otherwise secured open in the vent pathway.

The SR has been modified by a Not_ that requires performance only if a Shutdown Cooling System suction line relief valve is being used for overpressure pr(_tection. The Frequencies consider operating experience witlt mispositioning of unlocked and locked pathway vent 'alves.

SR 3.4.11.3 SRs are specified in the Inservic_ Testing Program.

Shutdown Cooling System suction l-ne relief valves are to be tested in accordance with the reqLHrements of the ASMEOM Code (Ref. 2), which provides the activities and the Frequency necessary to satisfy the SRs. The Shutdown Cooling System suction line relie valve setpoint is 467 psig.

(continued)

PALOVERDEUNITS1,2,3 B 3.4.11-5 REVISION54

Pressurizer Safety Valves-MODE 4 B 3.4.11 BASES (continued)

REFERENCES 1. ASME,Boiler and Pressure Vessel Code, Section Ill.

2. ASMECodefor Operations and Maintenanceof Nuclear PowerPlants.

3. PVNGS Operating License Amendment Nos. 75, 61, and 47 for Units 1, 2, and 3 respectively, and associated NRC Safety Evaluation dated May16, 1994.

PALOVERDEUNITS 1,2,3 B 3.4.11-6 REVISION54

LTOPSystem BASES B 3.4,13 ACTIONS B.1 (continued)

(continued)

Cooling System suction line relief valve failures without exposure to a lengthy period with only one Shutdown Cooling System suction line relief valve CPERABLE to protect against overpressure events.

C.1 If two required Shutdown Cooling System suction line relief valves are inoperable, or if a Required Action and the associated Completion Time of Condition A or B are not met, the RCSmust be depressurized and a vent established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The vent must be sized at least 16 square inches to ensure the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action protects the RCPBfrom a low temperature overpressure event and a possible brittle failure of the reactor vessel. For personnel safety considerations, the RCScold leg temperature must be reduced to less than 200°F prior to venting.

The Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to depressurize and vent the RCSis based on the time required to place the plant in this condition and the relatively low probability of an overpressure event during this tire period due to increased operator awareness of administrative control requirements.

SURVEILLANCE SR 3.4.13.1 and 3.4.13.2 REQUIREMENTS SR3.4.13.1 and SR3.4.13.2 require verifying that the RCS vent is open _ 16 square inches or that the ShutdownCooling Systemsuction line relief valves be aligned to provide overpressure protection for the RCSis proven OPERABLE by verifying its open pathway condition either:

ShutdownCoolin9 Systemsuction/line relief valves

a. Onceevery 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for a valve that is unlocked, not sealed, or otherwise not secLred open in the vent pathway, or

b. Onceevery 31 days for a val\e that is locked, sealed, or otherwise secured open in the vent pathway.

RCSVent

a. Onceevery 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for avert pathway that is unlocked, not sealed, or oth(rwise not secured open (continued)

PALOVERDEUNITS1,2,3 B 3.4.13-9 REVISION42

LTOPSystem B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 and 3.4.13.2 (continued)

REQUIREMENTS

b. Once every 31 days for a vent pathway that is locked, sealed, or otherwise secured open.

For an RCSvent to meet the specified flow capacity, it requires removing all pressurizer safety valves, or similarly establishing a vent by opening the pressurizer manway(Ref. 11). The vent path(s) must be above the level of reactor coolant, so as not to drain the RCSwhen open.

The passive vent arrangement must only be open (vent pathway exists) to be OPERABLE.These Surveillances need only be performed if the vent or the Shutdown Cooling System suction line relief valves are being used to satisfy the requirements of this LCO. The Frequencies consider operating experience with mispositioning of unlocked and locked pathway vent valves, and passive pathway obstructions.

SR 3.4.13.3 SRs are specified in the Inservice Testing Program.

I Shutdown tested in Cooling System accordance withsuction line relief ofvalves the requirements are to be the ASMEOM Code (Ref. 10), which provides the activities and the Frequency necessary to satisfy the SRs. The Shutdown Cooling System suction line relief valve set point is 467 psig.

REFERENCES 1 10 CFR50, Appendix G.

2 Generic Letter 88-11.

3 UFSAR,Section 15.

4 10 CFR50.46.

I 5 10 CFR50, Appendix K.

6 Generic Letter 90-06.

7 UFSAR,Section 5.2.

(continued)

I PALOVERDEUNITS 1,2,3 B 3.4.13-10 REVISION54

LTOPSystem BASES B 3.4.13 REFERENCES 8. Pressure Transient Analyses (continued) a, V-PSAC-O09(3876 MWtw/Original Steam Generators

[historical reference])

b. MN725-00118(Unit 2, 4070 _Wt w/Replacement Steam Generators)

c. MN725-00562(Units 31, 407 MWtw/Replacement Steam Generators)

9. Mass Input Pressure Transient in Water Solid RCS a, V-PSAC-010(3876 MWtw/Orilinal Steam Generators

[historical reference])

b. MN725-00117(Unit 2, 4070 MWtw/Replacement Steam Generators)

c. MN725-01495(Units 31,407( MWtw/Replacement Steam Generators)

10. ASMECode for Operation and _aintenance of Nuclear i Power Plants, I

11. 13-C00-93-016, Sensitivity Study on Pressurizer Vent Paths vs. Days Post Shutdown PALOVERDE UNITS1,2,3 B 3.4.13-11 REVISION54

This page intentionally blank RCSPIVLeakage B 3.4.15 BASES SURVEILLANCE SR 3.4.15.1 (continued)

REQUIREMENTS For the two PIVs in series, the l _kage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs a not individually leakage tested, one valve may have failed completely and not be detected if the other valve in _eries meets the leakage requirement. In this situation, t_e protection provided by redundant valves would be lost.

Testing is to be performed every 9 months, but may be extended up to 18 months, a typica refueling cycle, if the plant does not go into MODE5 for least 7 days. The 18 month Frequency is consistent th 10 CFR50.55a(g)

(Ref. 8), is within frequency all by the American I Society of Mechanical Engineers ) OMCode (Ref. 7), I and is based on the need to perfor the Surveillance under conditions that apply during a pla outage and the potential for an unplanned transie if the Surveillance were performed with the reactor at power.

In addition, testing must be perfo once after the valve has been opened by flow or exercis to ensure tight reseating. PIVs disturbed in the )erformance of this Surveillance should also be unless documentation shows that an infinite testing loo cannot practically be avoided. Testing must be performe within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limi for performing this test after opening or reseating a val The SDCPIVs excepted in two of th three FREQUENCIES are UV-651, UV-652, UV-653, and UV- due to position indication of the valves in the room.

Although not explicitly required SR 3.4.15.1, performance of leakage testing to verify l is below the specified limit must be performed prior to _turning a valve to service following maintenance, re r or replacement work on the valve in order to demonstrate )erability.

The leakage limit is to be met at RCSpressure associated with MODES1 and 2. s permits leakage testing at high differential pressures wi stable conditions not possible in the MODESwith lower ressures.

(continued)

PALOVERDEUNITS 1,2,3 B 3.4.15-5 REVISION54

RCSPIV Leakage B 3.4.15 BASES SURVEILLANCE SR 3.4.15.1 (continued)

REQUIREMENTS Entry into MODES3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. The Note that allows this provision is complimentary to the Frequency of prior to entry into MODE2 whenever the unit has been in MODE5 for 7 days or more, if leakage testing has not been performed in the previous 9 months. In addition, this Surveillance is not required to be performed on the SDC System when the SDCSystem is aligned to the RCSin the shutdown cooling mode of operation. PIVs contained in the SDCshutdown cooling flow path must be leakage rate tested after SDCis secured and stable unit conditions and the necessary differential pressures are established.

SR 3.4.15.2 Verifying that the SDCopen permissive interlocks are I OPERABLE,when tested as described in Reference 10, ensures that RCSpressure will not pressurize the SDCsystem beyond 125%of its design pressure of 485 psig. The interlock setpoint that prevents the valves from being opened is set so the actual RCSpressure must be <410 psia to open the valves. This setpoint ensures the SDCdesign pressure will not be exceeded and the SDCrelief valves (Reference 9) will not lift. The 18 month Frequency is based on the need to perform this Surveillance under conditions that apply during a plant outage. The 18 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment.

(continued)

PALOVERDEUNITS1,2,3 B 3.4.15-6 REVISION35

RCSPIV Leakage B3.4.15 BASES (continued)

REFERENCES 1 10 CFR50.2.

2 10 CFR50.55a(c).

3 10 CFR50, Appendix A, Section V, GDC55.

4 WASH-1400 NUREG-75/014), A_)endix V, October 1975.

5 NUREG-0677May 1980.

6 UFSAR,Section 3.9.6.2 7 ASMECode for Operation and Maintenance of Nuclear Power Plants.

8 10 CFR50.55a(g).

9 T.S. LCO3.4.13 (LTOP)

10. UFSARSection 7.6.2.2.1, (4. 0).

PALOVERDEUNITS 1,2,3 B 3.4.15-7 REVISION54

This page intentionally blank ECCS - Operating B 3.5.3 BASES SURVEILLANCE SR 3.5.3.3 REQUIREMENTS (continued) Periodic surveillance testing of ESCSpumpsto detect gross degradation caused by impeller strJctural damage or other hydraulic component problems is reTuired by the ASMEOM Code. This type of testing may be accomplished by measuring the pump developed head at only on? point of the pump characteristic curve. This verifies both that the measured performance is within an acceptabl? tolerance of the original pump baseline performance and that the performance at the test flow is greater than o equal to the performance assumed in the unit safety analysi SRs are specified in the Inservice Testing Program, whi h encompasses the ASMEOM Code (Ref. 7). The frequency of t_is SR is in accordance with the Inservice Testing Program SR 3.5.3.4, SR 3.5.3.5, and SR 5.3.6 These SRs demonstrate that each auLomatic ECCSvalve actuates to the required position Dn an actual or simulated SIAS and on an RAS, that each ECCSpump starts on receipt of an actual or simulated SIAS, and tqat the LPSI pumps stop on receipt of an actual or simulated _AS. This Surveillance is not required for valves that are IDcked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration Df the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the Engineered Safety Feature Actuation System (ESFAS) testing, and equipment performance is monitored as part of the Inservice Testing Program.

The following valve actuations must be verified at least once per 18 months:

on an actual or simulated re(irculation actuation signal, the containment sump isolation valves open, and the HPSI, LPSI and CS mirimum bypass recirculation flow line isolation valves arc combined SI mini flow valve close.

(continued)

PALOVERDEUNITS1,2,3 B 3.5.3-9 REVISION54

ECCS Operating B 3.5.3 BASES SURVEILLANCE SR 3.5.3.7 REQUI REMENTS (continued) Realignment of valves in the flow path on an SIAS is necessary for proper ECCSperformance. The safety injection valves have stops to position them properly so that flow is restricted to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimumflow. The 18 month Frequency is based on current industry practice.

These valves are also monitored in accordancewith the requirements of 10 CFR50.65 (Ref. 5).

SR 3.5.3.8 Periodic inspection of the containment sumpensures that it is unrestricted and stays in proper operating condition.

The 18 month Frequencyis based on the need to perform this Surveillance under the conditions that apply during an outage, on the need to have access to the location, and on the potential for unplannedtransients i f the SurveiI Iance were performed with the reactor at power. This Frequency is sufficient to detect abnormaldegradation and is confirmed by operating experience.

REFERENCES 1. 10 CFR50, Appendix A, GDC35.

2. 10 CFR50.46.

3. UFSAR,Chapter 6.

4. NRCMemorandum to V. Stello, Jr., from R. L. Baer, "Recommended Interim Revisions to LCOsfor ECCS Components,"December1, 1975.

5. 10 CFR50.65.

6. CombustionEngineering OwnersGroup Joint Applications Report for Low Pressure Safety Injection SystemAOT Extension, CE NPSD-995,dated May 1995, as submitted to NRCin APSletter no. 102-03392, dated June 13, 1995, with updates described in letter no. 102-04250 dated February 26, 1999. Also see TS amendment no. 124 dated February 1, 2000.

PowerPlants.

I 7. ASMECodefor Operation and Maintenance of Nuclear I

PALOVERDE UNITS1,2,3 B 3.5.3-10 REVISION 54

RWT B 3,5.5 B 3.5 EMERGENCY CORECOOLINGSYSTEMS (ECCS)

B 3.5.5 Refueling Water Tank (RWT)

BASES BACKGROUND The RWTsupports the ECCSand the ]ontainment Spray System by providing a source of borated w_ter for Engineered Safety Feature (ESF) pumpoperation.

The RWTsupplies two ECCStrains b/ separate, redundant supply headers. Each header also supplies one train of the Containment Spray System. A motor operated isolation valve is provided in each header to alloN the operator to isolate the usable volume of the RWTfrom the ECCSafter the ESF pump suction has been transferred to the containment sump following depletion of the RWTduring a Loss of Coolant Accident (LOCA). A separate header is used to supply the Chemical and Volume Control System (CVCS) from the RWT. Use of a single RWTto supply both trains of the ECCSis acceptable since the RWTis a passive component, and passive failures are not assumedto occur soincidently with the Design Basis Event during the injestion phase of an accident. Not all the water storeJ in the RWTis available for injection following a LOCA;the location of the ECCS suction piping in the RWTwill result in some portion of the stored volume being unavailable.

The High Pressure Safety Injection (HPSI), Low Pressure Safety Injection (LPSI), and containment spray pumps are provided with recirculation lines that ensure each pumpcan maintain minimum flow requirements when operating at shutoff head conditions. These lines discharge back to the RWT.

The RWTvents to the Fuel Building Ventilation System. When the suction for the HPSI and containment spray pumps is transferred to the containment sun), this flow path must be isolated to prevent a release of tle containment sump contents to the RWT. If not isolated, this flow path could result in a release of contaminants to the atmosphere and the eventual loss of suction head for the ESF pumps.

This LCOensures that:

a. The RWTcontains sufficient L_orated water to support the ECCSand Containment SpryLySystem during the injection phase; (continued)

PALOVERDEUNITS 1,2,3 B 3.5.5-1 REVISION54

RWT B3.5.5 I BASES BACKGROUND b. Sufficient water volume exists in the containment sump (continued) to support continued operation of the ESF pumps at the time of transfer to the recirculation mode of cooling; and

c. The reactor remains subcritical following a LOCA.

Insufficient water inventory in the RWTcould result in (1) insufficient cooling capacity of the ECCSand Containment Spray System, or (2) insufficient water level to support continued ESF pumpoperation when the transfer to the recirculation mode occurs. Improper boron concentrations could result in a reduction of SDMor excessive boric acid precipitation in the core following a LOCA, as well as excessive caustic stress corrosion of mechanical components and systems inside containment.

The RWTalso provides a source of borated water to the charging system for makeupto the RCSto compensate for contraction of the RCScoolant during plant cooldown while maintaining adequate shutdown margin. Although this charging system boration function is not required to be in a Technical Specification LCOper 10 CFR50.36(c)(2)(ii) criteria, the RWTvolume requirements of Figure 3.5.5-1 include this function in order to provide the plant operators with a single requirement for RWTvolume.

(continued)

PALOVERDEUNITS 1,2,3 B 3.5.5-2 REVISION54

RWT B 3.5.5 BASES BACKGROUND The table below provides the required RWTlevel at selected (continued) RCSaverage temperature values, corresponding to Figure 3.5.5-1. The RWTvolume is the tctal volume of water in the RWTabove the vortex breaker. This volume includes the volumes required to be transferred, as discussed below, an allowance for instrument uncertainty, and the volume that will remain in the RWTafter the s_itch over to the recirculation mode.

RWT RequiredLevelat RCS Temperatures RCSTemperature (°F) RWTRequired Level RWTVolume

average Indicated (Gallons)

(%)

Pre-RWTTS After-RWT TS Pre-RWTTS After-RWT Setpoint Setpoint Setpoint TS Setpoint Change Change Change Change 210 79.9 81.2 601,000 611,000 250 80.1 81.4 603,000 613,000 300 80.4 81.8 605,000 615,000 350 80.8 82.1 608,000 618,000 400 81.2 82.5 611,000 621,000 450 81.6 83.0 614,000 624,000 500 82.1 83.5 618,000 628,000 565 83.0 84.3 624,000 634,000 600 83.0 84.3 624,000 634,000

The volumes include instrument uncertainty and have been rounded up or down to the nearest 1,000 gallons.

(continued)

PALOVERDEUNITS 1,2,3 B 3.5.5-3 REVISION54

RWT B3.5.5 BASES APPLICABLE During accident conditions, the RWTprovides a source of SAFETYANALYSESborated water to the HPSl, LPSl and containment spray pumps.

As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 1). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of Bases B 3.5.3, "ECCS- Operating," and B 3.6.6, "Containment Spray." These analyses are used to assess changesto the RWTin order to evaluate their effects in relation to the acceptance limits.

The level limit of Figure 3.5.5-1 for the ESFfunction is based on the largest of the following four factors:

I a. A volume of borated via the water must be transferred to a low containment ESFpumpsprior to reaching level switchover to the containment sumpfor recirculation. This ESFReserveVolumeensures that the ESFpumpsuction will not be aligned to the containment sumpuntil the point at which 75%of the minimumdesign flow of one HPSl pumpis capable of meeting or exceeding the decay heat boil-off rate.

b. A volume of borated water must be transferred to the RCSand containment for flooding of sumpstrainers to prevent vortexing and to ensure adequatenet positive suction head to support continued ESFpumpoperation after the switchover to recirculation occurs.

I c. A volume of borated water must be available for ContainmentSpray Systemoperation as credited in the containment pressure and temperature analyses.

I d. A volume ofto borated functions ensure water is neededduring shut downmargin ECCS (SDM)is maintained. The volume required is similar to that neededfor the charging system function of compensatingfor contraction of the RCScoolant during plant cooldown. The volume required will vary depending upon the event and is boundedby the volume (continued)

PALOVERDE UNITS1,2,3 B 3.5.5-4 REVISION 54 l

ContainmentSpraySystem B 3.6.6 BASES SURVEILLANCE SR 3.6.6.2 REQUIREMENTS (continued) Verifying that the containment sp ay header piping is full of water to the 113 ft level minimizes the time required to fill the header. This ensures that spray flow will be admitted to the containment atmosphere within the time frame assumed in the containment analysis. The analyses shows that the header may be filled with unborated water which helps to reduce boron plate out due to evaporation. The 31 day Frequency is based on the static nature of the fill header and the low probability of a significant degradation of water level in the piping occurring between surveillances. The value of 113 ft is an indicated value which accounts for instrument unc_rtainty.

SR 3.6.6.3 Verifying that each containment s ray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pumpperformance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASMEOMCode (Ref. 6). Since the I containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow (either full flow or miniflow as conditions permit). This test is indicative of overall performance. Such inservice inspections confirm component OPER&BILITY,trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.

(conti nued)

PALOVERDEUNITS 1,2,3 B 3.6.6-7 REVISION54

ContainmentSpray System B 3.6.6 BASES SURVEILLANCE SR 3.6.6.4 and SR 3.6.6.5 (continued)

REQUIREMENTS These SRsverify that each automatic containment spray valve actuates to its correct position and that each containment spray pumpstarts upon receipt of an actual or simulated safety injection actuation signal, recirculation actuation signal and containment spray actuation signal as applicable.

This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for an unplannedtransient if the Surveillances were performed with the reactor at power.

Operating experience has shownthat these componentsusually pass the Surveillances when performed at the 18 month Frequency. Therefore, the Frequencywas concluded to be acceptable from a reliability standpoint.

The surveillance of containment sumpisolation valves is also required by SR3.5.3.5. A single surveillance maybe used to satisfy both requirements.

SR 3.6.6.6 Unobstructed flow headers and nozzles are determined by either flow testing or visual inspection.

With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. Performanceof this SR demonstrates that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment during an accident is not degraded. Dueto the passive design of the nozzle, a test at 10 year intervals is considered adequate to detect obstruction of the spray nozzles.

(continued)

PALOVERDE UNITS1,2,3 B 3.6.6-8 REVISION 48

Containment Spray System B 3.6.6 BASES REFERENCES 1. 10 CFR50, Appendix A, GDC3E GDC39, GDC40, GDC41, GDC42, and GDC43.

2. UFSAR,Section 6.2.

3. UFSAR,Section 6.5.

4. UFSAR,Section 7.3.

5. UFSAR,Section 3.1.34

6. ASMECode for Operation and aintenance of Nuclear Power Plants.

7. 10 CFR50.44.

8. Regulatory Guide 1.7, Revisi n O.

PALOVERDEUNITS 1,2,3 B 3.6.6-9 REVISION54

This page intentionally blank MSSVs B 3.7.1 BASES ACTIONS D.1 (continued)

Whenmore than eight required MSSVsper steam generator are inoperable, the unit must be placed in a MODEin which the LCOdoes not apply. To achieve this status the unit must be placed in at least MODE3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITYcf the MSSVsby the verification of each MSSVlift setpoints in accordance with the Inservice Testing Program. The ASMEOMCode (Ref. 4),

requires the following tests for PSSVs:

a. Visual examination;

b. Seat tightness determination

c. Setpoint pressure determinat on (lift setting);

d. Compliance with owner's seat tightness criteria; and

e. Verification of the balancin( device integrity on bal anced valves.

The ASMEOMCode requires that al valves be tested every 5 years, and a minimum of 20%of th valves tested every 24 months. The ASMEOMCode specifie the activities and frequencies necessary to satisfy the requirements.

Table 3.7.1-2 allows a +/- 3% setpoint tolerance for OPERABILITY;however, the valves re reset to +/- 1% during the Surveillance to allow for dri t.

PALOVERDEUNITS 1,2,3 B 3.7.1-5 REVISION54

MSSVs B 3.7.1 BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS (continued) This SR is modified by a Note that allows entry into and operation in MODE3 prior to performing the SR. This is to allow testing of the MSSVsat hot conditions. The MSSVsmay be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVsare not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

REFERENCES 1. UFSAR,Section 5.2.

2. ASME,Boiler and Pressure Vessel Code, Section Ill, Article NC-7000, Class 2 Components.

3. UFSAR,Section 15.2.

4. ASMECode for Operation and Maintenance of Nuclear I Power Plants.

I PALOVERDEUNITS 1,2,3 B 3.7.1-6 REVISION54

MSIVs B 3.7.2 BASES (continued)

ACTIONS E.1 (continued)

(continued) or more MSIVs inoperable while in _IODE1 requires entry into LCO3.0.3.

F.1 With one MSIV inoperable in MODE1, time is allowed to restore the component to OPERABLE status. Some repairs can be made to the MSIV with the unit qot. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, considering the probability of an accident occurring during th time period that would require closure of the MSIVs.

Condition F is entered when one MSV is inoperable in MODE1, including when both actuator trains for one MSIV are operable. Whenonly one actuator train is inoperable on one MSIV, Condition A applies.

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is consistent with that normally al I owed for containment i sol ation valves that i sol ate a closed system penetrating containment. These valves differ from other containment isolation valves in that the closed system provides an additional means for containment isolation.

G.I I If the MSIV cannot be restored to OPERABLE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the unit must be placed in a MODEin which the LCOdoes not apply. To achieve this status, the unit must be placed in MODE2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition H would be entered. The I Completion Time is reasonable, based on operating experience, to reach MODE2, and close the MSIVs in an orderly manner and without challenging unit systems.

H.1 and H.2 Condition H is modified by a Note indicating that separate Condition entry is allowed for each MSIV.

Since the MSIVs are required to be OPERABLE in MODES 2 and 3, the inoperable MSIVs may either be restored to OPERABLE status or closed. Whenclosed, the MSIVs are already in the position required by the assumptions in the safety analysis.

(continued)

PALOVERDEUNITS 1,2,3 B 3.7.2-7 REVISION40

MSIVs B 3.7.2 BASES(continued)

ACTIONS H.1 and H.2 (continued)

(continued)

The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is consistent with that allowed in Condition F.

Inoperable MSIVsthat cannot be restored to OPERABLE status within the specified Completion Time but are closed must be verified on a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, MSIVstatus indications available in the control room, and other administrative controls, to ensure these valves are in the closed position.

I.l and 1.2 If the MSIVs cannot be restored to OPERABLE status, or closed, within the associated Completion Time, the unit must be placed in a MODEin which the LCOdoes not apply. To achieve this status, the unit must be placed in at least MODE3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from MODE2 conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that the closure time of each MSIV is within the limit given in Reference 5 with each actuator train on an actual or simulated actuation signal and is within that assumed in the accident and containment analyses. This SR also verifies the valve closure time is in accordance with the Inservice Testing Program. This SR I is normally performed upon returning the unit to operation following a refueling outage. The MSIVs should not be full stroke tested at power.

The Frequency for this SR is in accordance with the Inservice Testing Program. This Frequency demonstrates the valve closure time at least once per refueling cycle.

(continued)

I PALOVERDEUNITS 1,2,3 B 3.7.2-8 REVISION54

MSIVs B 3.7.2 BASES (continued)

SURVEILLANCE SR 3.7.2.1 (continued)

REQUIREMENTS (continued) This test is conducted in MODE3, Nith the unit at operating temperature and pressure, as discussed in the Reference 6 exercising requirements. This SR is modified I l

by a Note that allows entry into and operation in MODE3 prior to performing the SR. This allows a delay of testing until MODE3, in order to establisl conditions consistent with those under which the acceptance criterion was generated.

REFERENCES 1. UFSAR,Section 10.3.

2. CESSAR,Section 6.2.

3. UFSAR,Section 15.1.5.

4. 10 CFR100.11.

5. UFSAR,Section 5.1.5

6. ASMECode for Operation and _aintenance of Nuclear Power Plants.

PALOVERDEUNITS1,2,3 B 3.7.2-9 REVISION54

This page intentionally blank MFIVs B 3.7.3 BASES (continued)

SURVEILLANCE SR 3.7.3.1 REQUIREMENTS This SR verifies that closure time of each MFIV is within the limit given in Reference 2 on an actual or simulated actuation signal and is within that assumed in the accident and containment analyses. This SR also verifies the valve closure time is in accordance with the Inservice Testing Program. This SR is normally perfDrmed upon returning the unit to operation following a refueling outage. The MFIVs should not be full stroke tested a power.

The Frequency is in accordance wit the Inservice Testing Program. The Frequency for valve losure time is based on the refueling cycle. Operating experience has shown that these components usually pass the SRwhen performed at the specified Frequency.

REFERENCES 1. UFSAR,Section 10.4.7.

2. UFSAR,Section 5.1.5.

PALOVERDEUNITS 1,2,3 B 3.7.3-5 REVISION54

This page intentional ly blank AFWSystem B 3.7.5 BASES SURVEILLANCE SR 3.7.5.2 (continued)

REQUIREMENTS normal tests of pumpperformance required by the ASMEOM Code (Ref. 2). Because it is undesirable to introduce cold AFWinto the steam generators whi e they are operating, this testing may be performed on recir, ulation flow. This test confirms one point on the pumpdesign curve and can be indicative of overall performance. Such inservice tests confirm component OPERABILITY,trend performance, and detect incipient failures by indicating abnormal performance.

Performance of inservice testing, discussed in the ASMEOM Code, (Ref. 2), at 3 month intervals satisfies this requirement.

This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established.

Normal operating pressure is established in the steam generators when RCStemperature reaches 532°F, this corresponds to a Psatof 900 psia. Fhis deferral is required because there is an insufficient steam pressure to perform the test.

SR 3.7.5.3 This SR ensures that AFWcan be delivered to the appropriate steam generator, in the event of any accident or transient that generates an AFASsignal, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal. This Surveillance is not required for valves that are locked sealed, or otherwise secured in the required position under administrative controls. This SR is not required for the non-essential train since there are no automatic valves which receive an AFAS. The 18 morth Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage ard the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 18 month Frequency is acceptable, based on the design reliability ard operating experience of the equipment.

This SR is modified by a Note indicating that the SR should be deferred until suitable test ccnditions have been established. Normal operating pressure is established in the steam generators when RCStemFerature reaches 532°F, this corresponds to a Psatof 900 psia. This deferral is required because there is an insufficient steam pressure to perform the test.

(continued)

PALOVERDEUNITS 1,2,3 B 3.7.5-9 REVISION54

AFWSystem B 3.7.5 BASES SURVEILLANCE SR 3.7.5.3 (continued)

REQUIREMENTS Also, this SRis modified by a Note that states the SR is not required in MODE 4. In MODE 4, the required AFWtrain is already aligned and operating.

SR 3.7.5.4 This SRensures that the essential AFWpumpswill start in the event of any accident or transient that generates an AFASsignal by demonstrating that each essential AFWpump starts automatically on an actual or simulated actuation signal. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The 18 month Frequencyis acceptable, based on the design reliability and operating experience of the equipment.

The non-essential AFWpumpdoes not automatically activate and is not subject to this SR.

This SR is modified by two Notes. Note 1 indicates that the SRbe deferred until suitable test conditions are established. Normaloperating pressure is established in the steam generators whenRCStemperature reaches 532°F, this corresponds to a Psatof 900 psi a. This deferral is required becausethere is insufficient steam pressure to perform the test. Note 2 states that the SR is not required in MODE 4. In MODE 4, the required pumpis already operating and the autostart function is not required.

SR 3.7.5.5 This SRensures that the AFWSystemis properly aligned by verifying the flow path from each essential AFWpumpto each steam generator prior to entering MODE 2 operation, after 30 days in MODE 5 or 6. OPERABILITY of essential AFWflow paths must be verified before sufficient core heat is generated that would require the operation of the AFWSystem during a subsequent shutdown. The Frequencyis reasonable, based on engineering judgment, and administrative controls to ensure that flow paths remain OPERABLE.

(continued)

I I

PALOVERDEUNITS1,2,3 B 3.7.5-10 REVISION9

AFWSystem B 3.7.5 BASES SURVEILLANCE SR 3.7.5.5 (continued)

REQUIREMENTS To further ensure AFWSystem aligrnent, the OPERABILITYof the essential AFWflow paths is verified following extended outages to determine that no misal gnment of valves has occurred. This SR ensures that th flow path from the CST to the steam generators is properl aligned by requiring a verification of minimum flow capac ty of 650 gpm at pressures corresponding to 1270 psia at the entrance to the steam generators. (This SR is not required for the non-essential AFWpump since it is normally used for startup and shutdown.)

REFERENCES 1. UFSAR,Section 10.4.9.

2. ASMECode for Operation and _aintenance of Nuclear Power Plants.

PALOVERDEUNITS 1,2,3 B 3.7.5-11 REVISION54

This page intentionally blank CST B 3.7.6 B 3.7 PLANTSYSTEMS B 3.7.6 CondensateStorage Tank (CST)

BASES BACKGROUND The CSTprovides a safety grade source of water to the steam generators for removing decay and sensible heat from the Reactor Coolant System (RCS). The CST is the primary source of water for the Auxiliary Feedwater (AFW) System (LCO3.7.5, "Auxiliary Feedwater (&FW) System"). The steam produced is released to the atmosphere by the Main Steam Safety Valves (MSSVs)or the atmos)heric dump valves.

When the main steam isolation valves are open, the preferred means of heat removal is to discharge steam to the condenser by the nonsafety grade path of the steam bypass control valves. The condensed steam is returned to the CSTby the condensate pump draw-off. This has the advantage of conserving condensate while minimizing releases to the environment.

Because the CSTis a principal component in removing residual heat from the RCS, it is designed to withstand earthquakes and other natural phenomena. The CST is designed to Seismic Category I requirements to ensure availability of the feedwater supply. Feedwater is also available from the Reactor Makeupdater Tank (RMWT).

A description of the CSTis found in the UFSAR, Section 9.2.6 (Ref. I).

APPLICABLE The CSThas sufficient volume to _aintain the plant for SAFETYANALYSES 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months at MODE3 followed by a symmetrical cooldown (two steam generators available) to shutdown cooling (SDC) entry conditions at the design cooldown rate in the event of main condenser unavailability.

The CSTinventory is demonstrated to be sufficient by satisfying the requirements of long-term cooling event which includes both LOCALong-Term Cooling and Reactor Systems Branch Technical Position 5-1 (RSB 5-1) "Design Requirements of the Residual Heat Removal System" (Ref. 4), scenario, described in UFSARAppendix 5C, atural Circulation Cooldown Analysis", is based on a natural circulation cooldown with both steam generators (SGs) available, using (continued)

PALOVERDE UNITS1,2,3 B 3.7.6-1 REVISION 54

CST B3.7.6 BASES APPLICABLE safety-grade equipment, assuminga loss of offsite power, a SAFETY ANALYSESlimiting single failure, and with minimal operator actions (continued) outside the control room, as approvedby the NRC. The RSB5-1 guidance requires 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months at hot standby prior to initiating cooldownand is analytically found to be the bounding event for CSTsizing.

Transients and accidents other than the RSB5-1 scenario and Long TermLOCAare evaluated deterministically in the UFSAR Chapter 15 analyses to demonstrate the ability to achieve hot standby conditions (Refs 2 and 3).

Cooldownscenarios to SDCentry conditions outside the "events" described here are outside the current Design Basis. The Licensing Basis for these scenarios is that there are no significant decay heat removal vulnerabilities when all available plant equipment and the EOPsare evaluated through the facility's probabilistic risk assessment, as documentedin the APSresolution of "Unresolved Safety Issue" (USl) A-45, "ShutdownDecayHeat RemovalRequirements"and response to GL88-20, "Individual Plant Examination for Severe Accident Vulnerabilities.

The CSTsatisfies Criterion 3 of 10 CFR50.36 (c)(2)(ii).

I LCO The CSTmust contain sufficient cooling water to remove decay heat for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following a reactor trip from 102%

RTP, and then cool downthe RCSto SDCentry conditions, assuminga coincident loss of offsite power and the most adverse single failure as required by RSB5-1.

I The CSTlevel required is a usable volume of _ 300,000 gallons, which is based on holding the unit in MODE 3 for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , followed by a cooldown to SDCentry conditions at I 75°F per hour. This basis is analytically boundedby the level required by the NRCStandard Review Plan Branch Technical Position, Reactor SystemsBranch 5-1 (Ref. 4).

OPERABILITY of the CSTis determined by maintaining the tank level at or above the minimumrequired level.

(continued)

I PALOVERDE UNITS1,2,3 B 3.7.6-2 REVISION 54

CST B 3.7.6 BASES APPLICABILITY In MODES1, 2, and 3, and in MODE4, when steam generator is being relied uponfor heat removal the CSTis required to be OPERABLE.

In MODES5 and 6, the CST is not r ,quired because the AFW System is not required.

ACTIONS A.1 and A.2 If the CSTlevel is not within the limit, the OPERABILITYof the backup water supply (RMWT)must be verified within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

OPERABILITYof the RMWT must include initial alignment and verification of the OPERABILITYof flow paths from the RMWT to the AFWpumps, and availability of 26 ft. (300,000 gal.)

of water in the RMWT. The CSTlevel must be returned to OPERABLE status within 7 days, as the RMWTmay be performing this function in addition to its normal functions. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, based on operating experience, to verify the OPERABIITY of the RMWT. The 7 day Completion Time is reasonable based on an OPERABLE RMWT being available, and the low prob bility of an event requiring the use of the water frcm the CSToccurring during this period.

(continued)

PALOVERDEUNITS 1,2,3 B 3.7.6-3 REVISION54

CST B3.7.6 BASES ACTIONS B.1 and B.2 (continued)

If the CSTcannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODEin which the LCOdoes not apply. To achieve this status, the unit must be placed in at least MODE3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE4, without reliance on steam generator for heat removal, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that the CSTcontains the required volume of cooling water. (This level _ 29.5 ft (300,000 gallons)).

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on operating experience, and the need for operator awareness of unit evolutions that may affect the CST inventory between checks. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to abnormal CSTlevel deviations.

REFERENCES 1. UFSAR,Section 9.2.6.

2. UFSAR,Chapter 6.

3. UFSAR,Chapter 15.

4. NRCStandard Review Plan Branch Technical Position I (BTP) RSB5-1.

PALOVERDEUNITS 1,2,3 B 3.7.6-4 REVISION54 I

Diesel Fuel Oil, LubeOil, andStarting Air B 3.8.3 BASES ACTIONS D.1 (continued)

With the new fuel oil properties defined in the Bases for SR 3.8.3.3 not within the required limits, a period of 30 days is allowed for restoring the stored fuel oil properties. This period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, or restore the stored fuel oil properties. This restoration may involve feed and bleed procedures, filtering, or combinations of these procedures. Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the DGwould still be capable of performing its intended function.

E.1 Each DGis OPERABLE with one air receiver capable of delivering an operating pressure of _ 230 psig indicated.

Although there are two independent and redundant starting air receivers per DG, only one starting air receiver is required for DGOPERABILITY. Each receiver is sized to accomplish 5 DGstarts from its ncrmal operating pressure of 250 psig, and each will start the DG in _ 10 seconds with a minimum pressure of 185 psig indicated. If the required starting indicated, airthereceiver starting is air

< 230 psigiand syste_ _ 185 psig s degraded and a period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable. This 48-hour period is acceptable based on the minimum starting air capacity (_ ]85 psig indicated), the fact that the DGstart must be accomplished on the first attempt (there are no sequential starts in emergency mode),

and the low probability of an event during this brief period. Calculation 13-JC-DG-203 (Ref. 8) supports the I proposed values for receiver pressures.

F.1 With a Required Action and associeted Completion Time not met, or one or more DGswith diesel fuel oil, lube oil, or starting air subsystem inoperable for reasons other than addressed by Conditions A through E, the associated DGmay be incapable of performing its intended function and must be immediately declared inoperable.

(continued)

PALOVERDEUNITS 1,2,3 B 3.8.3-5 REVISION54

Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES ACTIONS F.1 (continued)

A Note modifies condition F. Periodic starting of the Emergency Diesel Generator(s) requires isolation on one of the two normally aligned air start receivers. During the subsequent Diesel Generator start, the air pressure in the one remaining air receiver may momentarily drop below the I minimum normally required require pressure declaring ofthe185 psig now indicated.

running Diesel This would Generator inoperable, due to low pressure in the air start system.

This is not required, as the Diesel Generator would now be running following the successful start. Should the start not be successful, the DGwould be declared inoperable per the requirements of LCO3.8.1. As such, this Condition is modified by a Note stating that should the required starting air receiver pressure momentarily drop to <185 psig I indicated while starting the Diesel Generator on one air receiver only, then entry into Condition F is not required.

It is expected that this condition would be fairly short duration (approximately 8 minutes), as the air start compressors should quickly restore the air receiver pressure after the diesel start.

SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load. The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.

The 31 day Frequency is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and unit operators would be aware of any large uses of fuel oil during this period.

SR 3.8.3.2 This Surveillance ensures that sufficient lube oil inventory is available to support at least 7 days of full load operation for each DG. The 2.5 inches visible in the I sightglass requirement is based on the DGmanufacturer consumptionvalues for the run time of the DG. Implicit in this SRis the requirement to verify the capability to (continued)

PALOVERDEUNITS 1,2,3 B 3.8.3-6 REVISION51

Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.4 REQUIREMENTS (continued)

This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available. The system design requirements provide for a minimum of five engine start cycles without recharging. A start cycle is defined by the DGvendor, but usually is measured in terms of time (seconds or cranking) or engine cranking speed. The pressure specified in this SR is intended to reflect the lowest vaIJe at which the DGcan be considered OPERABLE.

The 31 day Frequency takes into account the capacity, capability, redundancy, and diversity of the AC sources and other indications available in the control room, including alarms, to alert the operator to below normal air start pressure.

SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil storage tanks once every 92 days eliminates the necessary environment for bacterial survival. This is the most effective meansof controllin_ microbiological fouling.

In addition, it eliminates the potential for water entrainment in the fuel oil during DGoperation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regardin_ the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1 137 (Ref. 10). This SR is for preventive maintenance. The resence of water does not necessarily represent failure of this SR provided the accumulated water is removed durin_ the performance of this Surveillance.

(continued)

PALOVERDEUNITS 1,2,3 B 3.8.3-9 REVISION41

Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES REFERENCES 1. FSAR, Section 9.5.4.2.

2. Regulatory Guide 1.137.

3. ANSl N195-1976, Appendix B.

4. FSAR, Chapter 6.

5. FSAR, Chapter 15.

6. ASTMStandards: D4057-81; D975-O7b; D976-91; D4737-90; D1796-83; D2276-89, Method A.

7. ASTMStandards, D975, Table 1.

I 8. "Emergency Systems Diesel Generator Instrumentation and Diesel Uncertainty Fuel Oil Calculation", 13-JC-DG-203, Parts 23 and 51 PALOVERDEUNITS 1,2,3 B 3.8.3-10 REVISION54

SDCandCoolantCirculation - High WaterLevel B 3.9,4 B 3.9 REFUELINGOPERATIONS B 3.9.4 Shutdown Cooling (SDC) and Coolant Circulat on - High Water Level BASES BACKGROUND The purposes of the SDCSystem in _ODE6 are to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC34, to p°ovide mixing of borated coolant, to provide sufficient coo ant circulation to minimize the effects of a boron di ution accident, and to prevent boron stratification (Ref. 1). Heat is removed from the RCSby circulating reactor coo ant through the SDCheat exchanger(s), where the heat is trmsferred to the Essential Cooling Water System via the SDCh_at exchanger(s). The coolant is then returned to the RC5via the RCScold leg(s).

Operation of the SDCSystem for normal cooldown or decay heat removal is manually accomplished from the control room.

The heat removal rate is adjusted Dy controlling the flow of reactor coolant through the SDChe_t exchanger(s) and bypassing the heat exchanger(s). _ixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the SDCSystem.

APPLICABLE If the reactor coolant temperature is not maintained below SAFETYANALYSES 200°F, boiling of the reactor coolant could result. This could lead to inadequate cooling of the reactor fuel due to a resulting loss of coolant in the reactor vessel.

Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of the boiling activity, and because of the possible addition of water to the reactor vessel with a lower boron concentration than is required to keep the reactor subcritical. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. One train of the SDCSystem is required to be operational in MODE6, with the water level 2 23 ft above the top of the reactor vessel flange, to prevent this challenge. The LCOdoes permit de-energizing of the SDC pump for short durations under the condition that the boron concentration is not diluted. This conditional de-energizing of the SDCpump does not result in a challenge to the fission product barrier. SDCand Coolant Circulation - High Water Level sat sfies Criterion 2 of 10 CFR50.36 (c)(2)(ii).

(continued)

PALOVERDEUNITS 1,2,3 B 3.9.4-1 REVISION0

SDCand Coolant Circulation - High Water Level B 3.9.4 BASES LCO Only one SDCloop is required for decay heat removal in MODE6, with water level _ 23 ft above the top of the reactor vessel flange. Only one SDCloop is required because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one SDCloop must be in operation to provide:

a. Removal of decay heat

b. Mixing of borated coolant to minimize the possibility of a criticality and

c. Indication of reactor coolant temperature.

An OPERABLE SDCtrain is composed of an OPERABLE SDCpump (LPSl or CS) capable of providing flow to the SDCheat exchanger for heat removal. SDCpumpsare OPERABLE if they are capable of being powered and are able to provide flow, I if required.

The LCOis modified by a Note that allows the required operating SDCloop to be removed from service for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in each 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no operations are permitted that would cause a reduction of the RCSboron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles, surveillance testing of ECCSpumps, and RCSto SDCisolation valve testing. During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.

APPLICABILITY OneSDCloop must be in operation in MODE 6, with the water level _ 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft level was selected because it corresponds to the 23 ft requirement established for fuel movementin LCO3.9.6, "Refueling Water Level Fuel Assemblies."

I (continued)

PALOVERDEUNITS 1,2,3 B 3.9.4-2 REVISION54

SDCand Coolant Circulation Low Water Level B 3.9.5 B 3.9 REFUELINGOPERATIONS B 3.9.5 Shutdown Cooling (SDC) and Coolant Circulation Low Water Level BASES BACKGROUND The purposes of the SDCSystem in _ODE6 are to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC34, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification (Ref. 1). Heat is removed from the RCSby circulating reactor coolant through the SDCheat exchanger(s), where the heat is transferred to the Essential Cooling Water System via the SDCheat exchanger(s). The coolant is then returned to the RCSvia the RCScold leg(s).

Operation of the SDCSystem for normal cooldown or decay heat removal is manually accomplished from the control room.

The heat removal rate is adjusted by controlling the flow of reactor coolant through the SDCheat exchanger(s) and bypassing the heat exchanger(s). Mixing of the reactor coolant is maintained by this contlnuous circulation of reactor coolant through the SDCSystem.

APPLICABLE If the reactor coolant temperatur_ is not maintained below SAFETYANALYSES 200°F, boiling of the reactor coolant could result. This could lead to inadequate cooling cf the reactor fuel due to the resulting loss of coolant in the reactor vessel.

Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components n_ar the areas of the boiling activity, and because of he possible addition of water to the reactor vessel with lower boron concentration than is required to keep the reac or subcritical. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. Two trains of the SDCSystem are required to be OPERABLE,and one train is required to be in operation in MODE6, with the water level < 23 ft above the top of the reactor vessel flange, to prevent this challenge.

SDCand Coolant Circulation - Low Water Level satisfies Criterion 2 of 10 CFR50.36 (c)(2)(ii).

(continued)

PALOVERDE UNITS1,2,3 B 3,9.5-1 REVISION 0

SDCand Coolant Circulation - Low Water Level B 3.9.5 BASES LCO In MODE6, with the water level < 23 ft above the top of the reactor vessel flange, both SDCloops must be OPERABLE.

Additionally, one loop of the SDCSystem must be in operation in order to provide:

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility of a criticality; and

c. Indication of reactor coolant temperature.

An OPERABLE SDCtrain is composedof an OPERABLE SDCpump (LPSl or CS) capable of providing flow to the SDCheat exchanger for heat removal. SDCpumps are OPERABLE if they are capable of being powered and are able to provide flow, if required.

Both SDCpumpsmay be aligned to the Refueling Water Tank (RWT) to support filling the refueling cavity or for performance of required testing.

The LCO is modified by a Note that allows a required operating SDCloop to be removed from service for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in each 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided no operations are permitted that would cause a reduction of the RCSboron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles, surveillance testing of ECCSpumps, and RCSto SDCisolation valve testing. During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.

This LCOis modified by a Note that allows one SDCloop to be inoperable for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided the other loop is OPERABLE and in operation. Prior to declaring the loop inoperable, consideration should be given to the existing plant configuration. This consideration should include that the core time to boil is not short, there is no draining operation to further reduce RCSwater level and that the capacity exists to inject borated water into the reactor vessel. This permits surveillance tests to be performed on the non-operating loop during a time when these tests are safe and possible.

(conti nued)

PALOVERDEUNITS 1,2,3 B 3.9.5-2 REVISION54 I

ML110610762

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