ML17264A760

Proposed Annual TS Bases

ML17264A760
Person / Time
Site:	Ginna
Issue date:	12/16/1996
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A761	List: ML17264A760 ML17309A608
References
NUDOCS 9612200087
Download: ML17264A760 (67)

Text

TABLE OF CONTENTS 2.0 2.1.1 2.1.2 8 3.0 8 3.0 8 3.1 8 3.1.1 8 3.1.2 8 3.1.3 8 3.1.4 8 3.1.5 8 3.1.6 8 3.1.7 8 3.1.8 8 3.2 8 3.2.1 8 3.2.2 8 3.2.3 8 3.2.4 SAFETY I IMITS{SLs')Reactor Core SLs........Reactor Coolant System (RCS)Pressure SL.....LIMITING CONDITION FOR OPERATION{LCO)APPLICABILITY

.SURVEILLANCE REQUIREMENT (SR)APPLICABILITY REACTIVITY CONTROL SYSTEMS.....'.

SHUTDOWN MARGIN (SDM)Cove Reactivity

.Moderator Temperature Coefficient (MTC)Rod Group Alignment Limits.......Shutdown Bank Insertion Limit Control Bank Insertion Limits Rod Position Indication PHYSICS TESTS Exceptions-MODE 2....POWER DISTRIBUTION LIMITS Heat Flux Hot Channel Factor (F<(Z))Nuclear Eqthalpy Rise Hot Channel N Facto'F~)o~~~~~~~~~~~AXIAL FLUX DIFFERENCE (AFD)QUADRANT POWER TILT RATIO (QPTR)8 2.0-1 8 2.0-1 8 2.0-8 3.0-1 3.0-12~8 3.1-"1 8 3.1-1 8 3.1-8 8 3.1-15 8 3.1-22 8 3.1-34 8 3.1-41 8 3.1-49 8 3.1-57 8 3.2-1 8 3.2-1 8 3.2-8 8 3.2-17 8 3.2-29 8 3.3 8 3.3.1 8 3.3.2 8 3.3.3 8 3.3.4 8 3.3.5 8 3.3.6 8 3.4 8 3.4.1 8 3.4.2 8 3.4.3 8 3.4.4 8 3.4.5 8 3.4.6 8 3.4.7 8 3.4.8 8 3.4.9 8 3.4.10 INSTRUMENTATION Reactor Trip System (RTS)Instrumentation Engineered Safety Feature Actuation System (ESFAS)Instrumentation Post Accident Monitoring

{PAN)Instrumentation Loss of Power (LOP)Diesel Generator (DG)Start Instrumentation

...........Containment Ventilation Isolation Instrumentati Control Room Emergency Air Treatment System (CREATS)Actuation Instrumentation REACTOR COOLANT SYSTEM (RCS)RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)Limits RCS Minimum Temperature for Criticality.

RCS Pressure and Temperature (P/T)Limits RCS Loops-MODE 1>8.5%RTP.RCS Loops-NODES 1 s 8.5/RTP, 2, and 3 RCS Loops-MODE 4 RCS Loops-MODE 5, loops Filled RCS Loops-MODE 5, Loops Not Filled Pressurizer Pressurizer Safety Valves on 8 3.3-1 8 3.3-1 8 3.3-64 8 3.3-108 8 3.3-130 8 3.3-,138 8 3.3-146 8 3.4-.1 8 3.4-1 8 3.4-8 8 3.4-12 8 3.4-20 8 3.4-24 8 3.4-31 8 3.4-37 8 3.4-43 8 3.4-47 8 3.4-5396i2200087 96i2i6 I'OR ADOCK 05000244 P PDR (continued)

R.E.Ginna Nuclear Power Plant iv Revision 1 TABLE OF CONTENTS 3.4 8 3.4.11 8 3.4.12 8 3.4.13 8 3.4.14 8 3.4.15 8 3.4.16 8 3.5 8 3.5.1 8 3.5.2 8 3.5.3 8 3.5.4 REACTOR COOLANT SYSTEM (RCS)(continued)

Pressurizer Power Operated Relief Valves (PORVs)Low Temperature Overpressure Protection (LTOP)S ystem RCS Operational LEAKAGE RCS Pressure Isolation Valve (PIV)Leakage..RCS Leakage Detection Instrumentation RCS Specific Activity EHERGENCY CORE COOLING SYSTEMS (ECCS)Accumulators 0~~~~~~~~~~~~ECCS-NODES'1, 2, and 3 ECCS-MODE 4.Refueling Water Storage Tank (RWST)~~~~~~~~~~8 3.4-58 8 3.4-68 8 3.4-85 8 3.4-92 8 3.4-100 8 3.4-108 8 3.5-1 8 3.5-1 8 3.5-10 8 3.5-25 8 3.5-29 8 3.6 8 3.6.1 8 3.6.2 8 3.6.3 8 3.6.4 8 3.6.5 8 3.6.6 3.6.7 8 3.6-1 8 3.6-1 8 3.6-8 8 3.6-18 8 3.6-38 8 3.6-42 CONTAINMENT SYSTEMS Containment Containment Air Locks Containment Isolation Boundaries

.........Containment Pressure Containment Air Temperature Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), NaOH, and Containment Post-Accident Charcoal Systems...............8 3.6'-46 Hydrogen Recombiners

..............8 3.6-66 8 3.7 8 3.7.1 8 3.7.2 3.7.3 8 3.7.4 8 3.7.5 8 3.7.6 8 3.7.7 8 3.7.8 8 3.7.9 8 3.7.10 8 3.7.11 8 3.7.12 8 3.7.13 8 3.7.14 PLANT SYSTEMS Hain Steam Safety Valves (HSSVs)....Hain Steam Isolation Valves (MSIVs)and Non-Return Check Valves.....Hain Feedwater Regulating Vhlves (HFRVs), Associated Bypass Valves, and Main Feedwater Pump Discharge Valves (MFPDVs)Atmospheric Relief Valves (ARVs)..Auxiliary Feedwater (AFW)System........Condensate Storage Tanks (CSTs)Component Cooling Water (CCW)System......Service Water (SW)System~Control Room Emergency Air Treatment System (CREATS)...Auxiliary Building Ventilation System (ABVS)Spent Fuel Pool (SFP)Water Level Spent Fuel Pool (SFP)Boron Concentration Spent Fuel Pool (SFP)Storage Secondary Specific Activity 3.7-1 3.7-1 8 3.7-6 8 3.7-13 8 3.7-22 8 3.7-27 8 3.7-42 8 3.7-46 8 3.7-55 8 3.7-65 8 3.7-75 8 3.7-82 8 3.7-86 8 3.7-90 8 3.7-97 (continued)

R.E.Ginna Nuclear Power Plant Revision TABLE OF CONTENTS 8 3.8 8 3.8.1 8 3.8.2 8 3.8.3 8 3.8.4 8 3.8.5 8 3.8.6 8 3.8.7 8 3.8.8 8 3.8.9 8 3.8.10 8 3.9 8 3.9.1 8 3.9.2 l 8 3.9 3 8 3.9.4 8 3.9.5 8 3.9.6 ELECTRICAL POWER SYSTEMS.AC Sources-MODES 1, 2, 3, and 4...AC Sources-MODES 5 and 6 Diesel Fuel Oil DC Sources-MODES 1, 2, 3, and 4....DC Sources-MODES 5 and 6 Battery Cell Parameters AC Instrument Bus Sources-HODES 1, 2, 3, AC Instrument Bus Sources-MODES 5 and 6 Distribution Systems-MODES 1, 2, 3, and Distribution Systems-MODES 5 and 6 REFUELING OPERATIONS

.Boron Concentration Nuclear Instrumentation Containment Penetrations

.......-.Residual Heat Removal (RHR)and Coolant Circulation-Mater Level a 23 Ft'esidual Heat Removal (RHR)and Coolant Circulation-Water Level<23 Ft Refueling Cavity Water Level~~0~~and 4 4 8 3.8-1 8 3.8-1 8 3.8-24 8 3.8-31 8 3.8-36 8 3.8-46 8 3.8-52 8 3.8-57 8 3.8-64 8 3.8-70 8 3.8-83 8 3.9-1 8 3.9-1 8 3.9-6 8 3.9-10 8 3.9-16 8 3.9-21 8 3.9-25R.E.Ginna Nuclear Power Plant vi Revision 1 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS B.2 B.3 B.4 B.5 and 8.6 (continued)

Verifying that'Fo(Z) and F~are within the required limits (i.e., SR 3.2.1.1 and SR 3.2.2.1)ensures that current operation't z 75%RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power.The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate Fo(Z)and F~.Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Accident for the duration of operation under these conditions.

As a m'inimum, the following accident analyses shall be re-evaluated:

a~b.C.d.e.f.g.Rod insertion characteristics; Rod misalignment; Small break loss of coolant accidents (LOCAs);Rod withdrawal at full power;Large break LOCAs;Main steamline break;and Rod ejection.A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.C.1 When Required Actions of Condition B cannot be completed within their Completion Time, the plant must be brought to a MODE or Condition in which the LCO requirements are not applicable.

To achieve this status, the plant must be brought to at least MODE 2 with K,<<<1.0 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which obviates concerns about the development of undesirable xenon or power distributions.

The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching MODE 2 with K,<<<1.0 from full power conditions in an orderl'y manner and without challenging plant systems.(continued)

R.E.Ginna Nuclear Power Plant 8 3.I-30 Revision I BASES RTS Instrumentation B 3.3.1 ACTIONS (continued)

U.l and U.2 Condition U applies to the RTB Undervoltage and Shunt Trip Mechanisms (i.e., diverse trip features)in MODES 1 and 2.Condition U applies on a RTB basis.This allows one diverse trip feature to be inoperable on each RTB.However, with two diverse.trip features inoperable (i.e., one on each of two different RTBs), at least one diverse trip feature must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable considering the low probability of an event occurring during this time interval.With one trip mechanism for one RTB inoperable, it must be restored to an OPERABLE status within 4S hours.The affected RTB shall not be bypassed while one of the diverse trip features is inoperable except for the time required to perform maintenance to one of the diverse trip features.The allowable time for performing maintenance of the diverse trip features'is 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the reasons stated under Condition T.The Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for Required Action U.2 is reasonable considering that in this Condition there is one remaining diverse trip feature for the affected RTB, and one OPERABLE RTB capable of performing the safety function and given the low probability of an event occurring during this interval.V.1 If the Required Action and Associated Completion Time of Condition R, S, T, or U is not met, the plant must be placed in a NODE where the Functions are no longer required to be OPERABLE.To achieve this status, the plant must be placed in NODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner without challenging plant systems.It should be noted that for inoperable channels of Functions 16a, 16b, 16c, and 16d, the MODE of Applicability will be exited before Required Action V.1 is completed.

Therefore, the plant shutdown may be stopped upon exiting the NODE of Applicability per LCO 3.0.2.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-50 Revision 1 0 0 BASES RTS Instrumentation B 3.3.1 ACTIONS (continued)

M.l and W.2 Condition M applies to the following reactor trip Functions in MODE 3, 4, or 5 with the CRD System capable of rod withdrawal or all rods not fully inserted:~RTBs;~RTB Undervoltage and Shunt Trip Mechanisms; and~Automatic Trip Logic.Mith two trip mechanisms irioperable, at least one trip mechanism must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is reasonable considering the low probability of an event occurring during this time ihterval;Mith one trip mechanism or train inoperable, the inoperable trip mechanism or train must be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.For the trip mechanisms, Condition M applies on a RTB basis.This allows one diverse trip feature to be inoperable on each RTB.However, with two diverse trip features inoperable (i.e., one on each of two different RTBs), at least one diverse trip feature must be restored to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.The Completion Time is reasonable considering that in this Condition, the remaining OPERABLE train is adequate to perform the safety function, and given the low probability of an event occurring during this interval.X.l and X.2 If the Required Action and Associated Completion Time of Condition M is not met, the plant must be placed in a NODE where the Functions are no longer required.To achieve this status, action be must initiated immediately to fully insert all rods and the CRD System must be incapable of rod withdrawal within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.These Completion Times are reasonable, based on operating experience to exit the MODE of Applicability in an orderly manner.R.E.Ginna Nuclear Power Plant B 3.3-51 (continued)

~Revision 1 ESFAS Instrumentation B 3.3.2 BASESAPPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) e.Auxiliar Feedwater-Undervolta e-Bus llA and llB The Undervoltage-Bus llA and 11B Functioh must be OPERABLE in MODES 1, 2, and 3 to ensure that the SGs remain the heat sink for the reactor.In MODE 4, AFW actuation is not required to be OPERABLE because either AFW or RHR will already be in operation to remove decay heat or sufficient time is available to manually place either system in operation.

This Function is not required to be OPERABLE in MODES 5 and 6 because there is not enough heat being generated in the reactor to require'he SGs as a heat sink.A loss of power to 4160 V Bus llA and 11B will be acc'ompanied by a loss of power to both MFW pumps and the subsequent need for some method of decay heat removal.The loss of offsite power is'etected by a voltage drop on each bus.Loss of power to both buses will start the turbine driven AFW pump to ensure that at least one SG contains enough water to serve as the heat sink for reactor decay heat and sensible heat removal following the reactor trip.Each bus is considered a separate Function for the purpose of this LCO.I Auxiliar Feedwater-Tri Of Both Hain Feedwater~Pum s A Trip of both HFW pumps is an indication of a loss of HFW and the subsequent need for some method of decay heat and sensible heat removal.The HFW pumps are equipped with a breaker position sensing device.An open supply breaker indicates that the pump.is not running.Two OPERABLE channels per HFW pump satisfy redundancy requirements with two-out-of-two logic.Each HFW pump is considered a Separate Function for the purpose of this LCO.A trip of both HFW pumps starts both motor driven AFW (HDAFW)pumps to ensure that at least one SG is available with water to act as the heat sink for the reactor.However, this actuation of the HDAFW pumps i's not credited in the mitigation of any accident.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-92 Revision 4 ESFAS Instrumentation 8 3.3.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY f.Auxiliar Feedwater-Tri Of Both Hain Feedwater~Pun s (continued)

During HODES 1 and 2, the AFW pumps may be providing for removal of decay heat with the HFW pumps removed from service.To prevent an unnecessary actuation of both HDAFW pumps under these conditions, a HFW pump breaker may be placed in the test position provided it is capable of being tripped on undervoltage and overcurrent conditions on the associated 4160 Y bus.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-92a Revision 4

ESFAS Instrumentation B 3.3.2 BASES ACTIONS (continued)

If the Required Actions and Completion Times of Condition L are not met, the plant must be brought to a MODE in which the LCO does not apply.To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and pressurizer pressure reduced to<2000 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.The allowed Completion Times are reasonable, based on.operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.N.l Condition N applies if a AFM Manual Initiation channel is inoperable.

If a manual initiation switch is inoperable, the associated AFM or SAFM pump must be declared inoperable and the applicable Conditions of LCO 3.7.5,"Auxiliary Feedwater (AFM)System" must be enter ed immediately.

Each AFM manual initiation switch controls one AFM or SAFW pump.Declaring the associated pump inoperable ensures that appropriate action is taken in LCO 3.7.5 based on the number and type of pumps involved.SURVEILLANCE RE(UIREHENTS The SRs for each ESFAS Function are identified by the SRs column of Table 3.3.2-1.~Each channel of process protection supplies both trains of the ESFAS.When testing Channel 1, Train A and Train 8 must be examined.Similarly, Train A and Train B must be examined when testing Channel 2, Channel 3, and Channel 4 (if applicable).

The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which.SRs apply to which ESFAS Functions.

{continued)

R.E.Ginna Nuclear Power Plant 8 3.3-100 Revision 4 PAH Instrumentation B 3.3.3 BASES LCO 19, 20.AFM Flow (continued)

The AFW System provides decay heat removal via the SGs and is comprised of the preferred AFM System and the Standby AFM (SAFM)System.The use of the preferred AFM or SAFW System to provide this decay heat removal.function is.dependent upon the type of accident.AFW flow indication is required from the three pump trains which comprise the preferred AFW System since these pumps automatically start on various actuation signals.The failure of the preferred AFW System (e.g., due to a high energy line break (HELB)in the Intermediate Building)'s detected by AFM flow indication.

At this point, the SAFM System is manually aligned to provide the decay heat removal function.SAFM flow can also be used to verify that AFW flow is being delivered to the SGs.However, the primary indication of this is provided by SG water level.Therefore, flow indication from the SAFW pumps is not required.Each of the three preferred AFW pump trains has two redundant transmitters; however, only the flow transmitter supplied power from the same electrical train as the AFM pump is required for this LCO.Therefore, flow transmitters FT-2001{HCB indicator FI-202lA)and FT-2006 (HCB indicator FI-2023A)comprise the two required channels for SG A and FT-2002 (HCB indicator FI-'2022A) and FT-2007 (HCB indicator FI-2024A)comprise the two required channels for SG B.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-12l Revision I

LOP DG Start Instrumentation B 3.3.4 BASES APPLICABLE SAFETY ANALYSES The LOP DG start instrumentation is required for the ESF Systems to function in any accident with a loss of offsite power.Its design basis is that of the ESF Actuation System (ESFAS).Undervoltage conditions which occur independent of any accident conditions result in the start and bus connection of the associated DG, but no automatic loading occurs.Accident"analyses credit the loading of the DG based on the*loss of offsite power during a Design Basis Accident (DBA).The most limiting DBA of concern is the large break loss of coolant accident (LOCA)which requires ESF Systems in'rder to maintain containment int'egrity and protect fuel contained within the reactor vessel (Ref.2).The detection and processing of an undervoltage condition, and subsequent DG loading, has been included in the delay time assumed for each ESF component requiring DG supplied power following a DBA and loss of offsite power.The loss of offsite power has been assumed to occur either coincident with the DBA or at a later period (40 to 90 seconds following the reactor trip)due to a grid disturbance caused by the turbine generator trip.If the loss of offsite power occurs at the same time as the safety injection (SI)signal parameters are reached, the accident analyses assumes the SI signal will actuate the DG within 2 seconds and that the DG will connect to the affected safeguards bus within an additional 10 seconds (12 seconds total time).If the loss of offsite power occurs before the SI signal parameters are reached, the accident analyses assumes the LOP DG start instrumentation will actuate the DG within 2.75 seconds and that the DG will connect to.the affected safeguards bus within an additional 10 seconds (12:75 seconds total time).If the loss of offsite power occurs after the SI signal parameters are reached (grid disturbance), the accident analyses assumes the DG will connect to the bus within 1.5 seconds after the feeder breaker to the bus i.s opened (DG was'actuated by SI signal).The grid disturbance has been evaluated based on a 140'F peak clad temperature penalty during a LOCA and demonstrated to result in acceptable consequences.

(continued)

R.E.Ginna Nuclear Power Plant'3.3-131 Revision 1 Containment Ventilation Isolation Instrumentation B 3.3.5 BASES ACTIONS (continued)

A Note has been added to the ACTIONS to clarify the application of.Completion Time rules.The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.5-1.The Completion Time(s)of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.l Condition A applies to the failure of one containment ventilation isolation radiation monitor channel.Since the two containment radiation monitors measure different parameters, failure of a single channel may result in loss of the radiation monitoring Function for certain events.Consequently, the failed channel must be restored to OPERABLE status.The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allowed to restore the affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events.8.1 Condition B applies to all Containment Ventilation Isolation Functions and addresses the train orientation of the system and the master and slave relays for these Functions.

It also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A.l.If a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue-as long as the Required Action for the applicable Conditions of LCO 3.6.3 is met for each valve made inoperable by failure of isolation instrumentation.

A Note is added stating that Condition B is only applicable in MOOE I, 2, 3, or 4.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-142 Revision I

Containment Ventilation Isolation Instrumentation 8 3.3.5 BASES ACTIONS (continued)

C.l and C.2 Condition C applies to all Containment Ventilation Isolation Functions and addresses the train orientation of the system and the master and slave relays for these Functions.

It also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A.l.If a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action to place each valve in its closed position or the applicable Conditions of LCO 3.9.3,"Containment Penetrations," are met for each valve made inoperable by failure of isolation instrumentation.

The Completion Time for these Required'Actions is Immediately.

A Note states that Condition C is applicable during CORE ALTERATIONS and during movement of irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS A Note has been added to the SR Table to clarify that Table 3.3.5-1 determines which SRs apply to which Containment Ventilation Isolation Functions.

SR 3.3.5.1 Performance of the CHANNEL CHECK once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensures that a.gross failure of instrumentation has not occurred and the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

The CHANNEL CHECK agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability.

If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-143 Revision I Containment Ventilation Isolation Instrumentation B 3.3.5 BASES SURVEILLANCE REqUIRENENTS SR 3.3.5.1 (continued)

The Frequency is based on operating experience that demonstrates channel failure is rare.The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels..

SR 3.3.5.2A COT is performed every 92 days on each required channel to ensure the entire channel will perform the intended Function.The Frequency is based on the staff recommendation for increasing the availability of radiation monitors according to NUREG-1366 (Ref.2).This test verifies the capability of the instrumentation to provide the containment ventilation system isolation.

The setpoint shall be left consistent with the current plant specific calibration procedure tolerance.

SR 3.3.5.3 This SR is the performance of an ACTUATION LOGIC TEST.All possible logic combinations, with and without applicable permissives, are tested for each protection function.In addition, the master relay is tested for continuity.

This verifies that the logic modules are OPERABLE and there is an.intact voltage signal path, to the master relay coils.This test is performed'very 24 months.The Surveillance interval is acceptable based on instrument reliability and industry operating experience.(continued)

R.E.Ginna Nuclear Power Plant B 3.3-144 Revision I

Containment Ventilation Isolation Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.5.4 A CHANNEL CALIBRATION is performed every 24 months, or approximately at every refueling.

CHANNEL'ALIBRATION is a complete check of the instrument loop, including the sensor.The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.REFERENCES 1.10 CFR 100.11.2.NUREG-1366.

R.E.Ginna Nuclear Power Plant B 3.3-145 Revision 1

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES (continued)

APPLICABLE SAFETY ANALYSES The requirements of this LCO represent the initial conditions for DNB limited transients analyzed in the plant safety analyses (Ref.1).The safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the DNB design criterion.

This is the acceptance limit for the RCS DNB parameters.,Changes to the plant that could impact these parameters must be assessed for their impact on the DNB design criterion.

The transients analyzed include loss of coolant flow events and dropped or stuck rod events.A key assumption for the analysis of.these events is that the core power distribution is within the limits of LCO 3:1.6,"Control Bank Insertion Limits";LCO 3.2.3,"AXIAL FLUX DIFFERENCE (AFD)";and LCO 3.2.4,"QUADRANT POWER TILT RATIO (QPTR)." The limit for pressurizer pressure is based on a+30 psig instrument uncertainty.

The accident analyses assume that nominal pressure is maintained at 2235 psig.By Reference 2, minor fluctuations are acceptable provided that the time averaged pressure is 2235 psig.The RCS coolant average temperature limit is based on a+4'F instrument uncertainty which includes a+1.5 F deadband.It is assumed that nominal T., is maintained within+1.5 F of the nominal T., specified in the COLR.By Reference 2, minor fluctuations are acceptable provided that the time averaged temperature is within 1.5 F of nominal.The limit for RCS flow rate is based on the nominal T.and SG plugging criteria limit.Additional margin of approximately 3%is then added for conservatism.

The RCS DNB parameters satisfy Criterion 2 of the NRC Policy Statement.

LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses.Opet ating within these limits will result in meeting the DNB design criterion in the event of a DNB limited transient.(continued)

'.E.Ginna Nuclear Power Plant B 3.4-3 Revision 4 RCS Loops-MODE 5, Loops Filled B 3.4.7 BASES (continued)

APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.The RCS loops are considered filled until the isolation valves are opened to facilitate draining of the RCS.The loops are also considered filled following the completion of filling and venting the RCS.However, in both cases, loops filled is based on the ability to use a SG as a backup.To be able to take credit for the use of one SG the ability to pressurize to 50 psig and control pre'ssure in the RCS must be available.

This is to prevent flashing and void formation at the top of the SG tubes which may degrade or interrupt the natural circulation flow path (Ref.2).One loop of RHR provides sufficient ci'rculation for these purposes.However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least one SG is required to be a 16%.Operation I LC0,3.4.4, LCO 3.4.5, LCO 3.4.6, LCO 3.4.8, LCO 3.9.4, LCO 3.9.5, in other'MODES is covered by: "RCS Loops-MODE 1>8.5%RTP";"RCS Loops-'MODES 1 s 8.5%RTPy 2y AND 3"RCS Loops-MODE 4";"RCS Loops-MODE 5, Loops Not Filled";"Residual Heat Removal (RHR)and Coolant Circulation-Water Level~23 Ft" (MODE 6);and"Residual Heat Removal (RHR)and Coolant Circulation-Mater Level<23 Ft" (MODE 6).ACTIONS A.l and A.2 If one RHR loop is inoperable and both SGs have secondary side water levels<16%, redundancy for heat removal is lost.Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore at least one SG secondary side water level.Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths.The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.The action to restore must continue until an RHR loop is restored to OPERABLE status or SG secondary side water level is restored.(continued)

R.E.Ginna Nuclear Power Plant B 3.4-40 Revision 1 RCS Loops-NODE 5, Loops Filled 8 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.7.3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the standby RHR pump.If secondary side water level is z 16%in at least one SG, this;Surveillance is not needed.The Frequency of 7 days is considered reasonable in view of other administrative controls available and'has been shown to be acceptable by operating experience.

REFERENCES 1.UFSAR, Section 14.6.1.2.6 2.NRC Information Notice 95-35.R.E.Ginna Nuclear Power Plant 8 3.4-42 Revision 1

CS, CRFC, NaOH, and Containment'Post-Accident Charcoal Systems B 3.6.6 BASES APPLICABLE SAFETY ANALYSIS (continued)

The analysis and evaluation show that under the worst case scenario, the highest peak containment pressure is 59.8 psig and the peak containment temperature is 374 F (both experienced during an SLB).Both results meet the intent of the design basis.(See the Bases for LCO 3.6.4,"Containment Pressure," and LCO 3.6.5," Containment Temperature," for a detailed discussion.)

The analyses and evaluations assume a plant specific power level of 102%, one CS train and one containment cooling train operating, and initial (pre-accident) containment conditions of 120 F and 1.0 psig..The analyses also assume a response time delayed initiation to provide conservative peak calculated containment pressure and temperature responses.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative.

In particular, the effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with incr easing containment backpressure.

For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with 10 CFR 50, Appendix K (Ref.7).The effect of an inadvertent CS actuation is not considered since there is no single failure, including the loss of offsite power, which results in a spurious CS actuation.

The modeled CS System actuation for the containment analysis's based on a response time associated with exceeding the containment Hi-Hi pressure setpoint to achieving full flow through the CS nozzles.To increase the response of the CS System, the injection lines to the spray headers are maintained filled with water.The CS System total response time is 28.5 seconds for one pump to the upper spray header and 26.5 seconds for.two pumps (average time between upper , and lower spray headers).These total response times (assuming the containment Hi-Hi pressure is reached at time zero)includes opening of the motor operated isolation valves, containment spray pump startup, and spray line filling (Ref.8).(continued)

R.E.Ginna Nuclear Power Plant B 3.6-51 Revision 1 CS, CRFC, NaOH, and Containment Post-Accident Charcoal Systems B 3.6.6 Iy Ill RUNS 1 ty 010 LcScndt Tltc RtVST and anociatcd cotnmon linc b sddtcatcd by tA30 33'S Pump Train Naon System-Not addrcslcd by LCD 3.6.6 CVCS I I Q 4044 or I ultratlon on RNR I cy I tI I I I I I I I N~ON I~ot II 011~fk", I I Sl I I 450A CS fteote A I I I~elo ecto I I I Sdottot aa I I atyA~I Io tlat 4410 I VIOee Ooedooeet Steer lyeta IIO Notetoet et to Choetod rdtot A~Cheteoel Iutet 4~+410A Qayca 44lc 4114 Cootoeooeet aeter ICea Q9 Itot decl CS Iteea 0Figure B 3.6.6-1 Containment Spray and NaOH Systems R.E.Ginna Nuclear Power Plant B 3.6-64 Revision 1' CS, CRFC, NaOH, and Containment Post-Accident Charcoal Systems B 3.6.6//Containment Recirculating Fan Cooling Unit A/->QP/Containment Recirculating Fan Cooling Unit B/Containmcnt Recirculating Fan Cooling Unit C 58 I 5873 (FO)(FC)5875 (FC)58 6 (FO)Post Accid Charcoal Filter Unit A 587 (FO)5874 (FO)Post Accident Charcoal Filter Unit B/Containmcn Recirculating ,Fan Cooling Unit D 5877 (FO)V Various Supply Points For illustration only Notes: 1.Dampers 5871 and 5872 are associated with Post Accident Charcoal Filter Unit A 2.Dampers 5874 and 5876 are associated with Post Accident Charcoal Filter Unit B 3.Damper 5873 is assoicated with both CRFC Unit A and Post Accident Charcoal Filter Unit A 4.Damper 5875 is associated with both CRFC Unit C and Post Accident Charcoal Filter Unit B Figure B 3.6.6-2 CRFC and Containment Post-Accident Charcoal Systems R.E.Ginna Nuclear Power Plant B 3.6-65 Revision 1 WFRY Bypass Yalvc 421 l-39t 399l 3995A 3993 SGA Stputpo.PA 3973 Fccdeatcr Heater SA MptCY 39~39SSA 3NSA 3Ãt 39Stg SN9 3933A 3933 O e2 tt)CS2 O s C23 CB CU th Caa Ch~t tts Ch Ftora Coadeosatc Booster bmps MFWLcadlog Ed Sc Ttaosdoccr g 3NO 3980 3913 3932A 3N2 hWV Potap B 39F4 MFPDY 3926 Fccdwctcr Heater SB 4.LCO 3.7.3 Condition 8 entered when any eombinalion of valve inopcrabilities results in an uniso!able ftowpath from lhe condensate booster pumps to onc or more SGs.Notes: 1.LCO 3.7.3 Condition A entered when MFPDV 3976 and/or 3977 is inoperable.

2.LCO 3.7.3 Condiuon B entered cvhen MFlCV 4269 and/or 4270 is inoperable.

3.LCO 3.7.3 Condition C entered when MFRV Bypass Valve 4271 and/or 4272 is inoperable.

MAY Bypass Yalw 4222 39SS S9 3 4 MHCY 3N6 3992 4220 39S4A 39'Sdh or ustra ono Y 3994 l SOB O cn C5.(cn Ch C5 DD Cay~NO~(Cay Ch

AFW System B 3.7.5 B 3.7 PLANT SYSTEHS B 3.7.5 Auxiliary Feedwater (AFW)System BASES BACKGROUND The AFW System supplies feedwater to the steam generators (SGs)to remove decay heat from the Reactor Coolant System (RCS)upon the.loss of normal feedwater.supply.The SGs function as a heat sink for core decay heat.The heat load is dissipated by releasing steam to the atmosphere from the SGs via the main steam safety valves (HSSVs)or atmospheric relief valves (ARVs).If the main condenser is available, steam may be released via the steam dump valves.The AFW System is comprised of two'separate systems, a preferred AFM System and a Standby AFW (SAFM)System (Ref.1).~AFM S stem The preferred AFW System consists of two, motor driven AFM (HDAFW)pumps and one turbine driven AFW (TDAFM)pump configured into three separate trains which are all located in.the Intermediate Building (see Figure B 3.7.5-1).Each HDAFM train provides 100%of AFM flow capacity, and the TDAFW pump~provides 200%of, the required capacity to the SGs, as assumed in the accident analysis.The pumps are equipped with independent recirculation lines to the condensate storage tanks (CSTs).Each HDAFW train is power ed from an i.ndependent Class lE power supply and feeds one SG, although each pump has the capability to be realigned from the control room to feed the other.SG via cross-tie lines containing normally closed motor operate'd valves (4000A and 4000B).The two HDAFM trains will actuate automatically on a low-low level signal in either SG, opening of the main feedwater (HFW)pump breakers, a safety injection (SI)signal, or the ATWS mitigation system actuati'on circuitry (AHSAC).The pumps can.also be manually started from the control room.(continued)

R.E.Ginna Nuclear Power Plant B 3.7-27 Revision 5

AFM System B 3.7.5 BASES BACKGROUND (continued)

The SAFW Pump Building environment is controlled by room coolers which are supplied by the same SW header as the pump trains.These coolers are required when the outside air tempe}ature is a 80 F to ensure the SAFM Pump Building remains s 120 F during accident conditions.

The AFM System is designed to supply sufficient water to the SG(s)to remove decay heat with SG pressure at the lowest HSSV set pressure plus l%%d.Subsequently, the AFW System supplies sufficient water to cool the plant to RHR entry conditions, with steam released through the ARVs.APPLICABLE SAFETY ANALYSES The design basis of the AFM System is, to supply water to the SG(s)to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the SGs at pressures corresponding to the lowest HSSV set pressure plus 1/.The AFM System mitigates the consequences'f.any event with the loss of normal feedwater.

The limiting Design Basis Accidents (DBAs)and transients for the AFW System are as follows (Ref.2): a.Feedwater Line Break (FWLB);b.Loss of HFM (with and without offsite power);c.Steam Line Break (SLB);d.Small break loss of coolant accident (LOCA);e.Steam generator tube rupture (SGTR);and f.External events (tornados and seismic events).AFM is also used to mitigate the effects of an ATWS event which is a beyond design basis event not addressed by this LCO.(continued)

R.E.Ginna Nuclear Power Plant B 3.7-29 Revision 5 AFW System 8 3.7.5 , BASES APPLICABLE SAFETY ANALYSES (continued)

The AFM System design is such that any of the above OBAs can be mitigated using the preferred AFM System or SAFM System.For the FWLB, SLB, and external events OBAs{items a, c, and f), the worst case scenario is the loss of all three preferred AFW trains due to a HELB in the Intermediate or Turbine Building, or a failure of the Intermediate Building block walls.For these three events, the use of the SAFW System within 10 minutes is assumed by the accident analyses.Since a single failure must also be assumed in addition to the HELB or external event, the capability of the SAFW System to supply flow to an intact SG, could be compromised if the SAFW cross-tie is not,available.

For HELBs within containment, use of either the SAFM System or the AFM System to the intact SG is assumed within 10 minutes.(For the SGTR events (item e), the accident analyses assume that one AFW train is available upon a SI signal or low-low SG level signal.Additional inventory is being added to the ruptured SG as a result of the SGTR such that AFW flow is not a critical feature for this OBA.The loss of MFW'(item b)is a Condition 2 event (Ref.3)which places limits on the response of the RCS from the transient (e.g., no challenge to the pressurizer power operated relief valves is allowed).This analysis has been performed assuming no AFM flow is available until 10 minutes with acceptable results.The most limiting small break LOCA (item d)analysis has also been performed assuming no AFW flow with no adverse impact on peak cladding temperature.

In summary, all limiting OBAs and transients have been analyzed assuming a 10 minute delay for actuation of flow.(continued)

R.E.Ginna Nuclear Power Plant B 3.7-30 Revision 5 AFW System B 3.7.5 BASES APPLICABLE SAFETY ANALYSES (continued)

I In addition to its accident mitigation function, the energy and mass addition capability of the AFW System is also consider ed with respect to HELBs within containment.

For SLBs and FWLBs within containment, maximum pump flow from all three AFW pumps is assumed for 10 minutes until operations can isolate the flow by tripping the AFM pumps or by closing the respective pump discharge flow path(s).Therefore, the motor oper ated discharge isolation valves for the motor HDAFM pump trains (4007 and 4008)are designed to limit flow to z 230 gpm to limit the energy and mass addition so that containment remains within design limits for items a and c.The TDAFM train is assumed to be at runout conditions (i.e., 600 gpm).The AFW System satisfies the requirements of Criterion 3 of the NRC Policy Statement.

LCO This LCO provides assurance that the AFW System will perform its design safety function to mitigate the consequences of accidents that could result in overpressurization of the reactor coolant pressure boundary or containment.

The AFW System is comprised of two systems which are configured into five trains.The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the SGs are OPERABLE (see Figures B 3.7.5-1 and 3.7.5-2).This requires that the following be OPERABLE: a.Two'DAFW trains taking suction from the CSTs as required by LCO 3.7.6 (and capable of taking suction from the SW system within 10 minutes), and capable of supplying their respective SG with a 200 gpm within 10 minutes and s 230 gpm total flow upon AFM actuation; b.The TDAFM train taking suction from the CSTs as required by LCO 3.7.6 (and capable of taking suction from the SW system within'10 minutes), provided steam is available from both main steam lines upstream of the HSIVs, and capable of supplying both SGs with a 200 gpm each within 10 minutes;and(continued)

R.E.Ginna Nuclear Power Plant B 3.7-31 Revision 5 AFW System 8 3.7.5 BASES LCO (continued) c.Two motor driven SAFW trains capable of being initiated either locally or from the control room within 10 minutes, taking suction from the SW System, and supplying their respective SG and the opposite SG through the SAFW cross-tie line with z 200 gpm.The piping, valves, instrumentation, and controls in the required flow paths are also required to be OPERABLE.The TDAFW train is comprised of a common pump and two flow paths.A TDAFW train flow path is defined as the steam supply line and the SG injection line from/to the same SG.The failure of the pump or both flow paths renders the TDAFW train inoperable.

The cross-tie line for the preferred HDAFM pumps is not required for this LCO.However, since the accident analyses have been performed assuming a 10 minute delay for AFM, and there are two separate systems, the use of this cross-tie line is allowed in MODES 1, 2, and 3.Also, provided that the AFW and SAFW discharge valves are set to provide the minimum required flow, the.recirculation lines for the preferred AFM system and SAFW system pumps are not credited in the accident analysis.The recirculation lines are also not required to be OPERABLE for this LCO since the HSSYs maintain the SG pressure below the pump's shutoff head.The SAFW Pump Building room coolers are required to be OPERABLE when the outside air temperature is z 80 F.If one room cooler is inoperable, the associated SAFW train is inoperable.

APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in the event that it is called upon to function when the HFW System is lost.In addition, the AFW System is required to supply enough makeup water'to replace the lost SG secondary inventory as the plant cools to HODE 4 conditions.

In HODE 4, 5, or 6, the SGs are not normally used for heat removal, and the AFW System is not required.(continued)

R.E.Ginna Nuclear Power Plant B 3.7-32 Revision 5 m 44S 2001 O ID 5 I I I I T 2022 TD TD 5 tl tD CCI 5 5 5 ID 63.CQ rl 4A5 tab~C Vl V)I tD CST A 4025 L I-----------I l I I a-IX~I 4015 4014 For illustration only 4Ols 4016 Note-t.I'-200l, t I'-2002, Fl'-2006 andFf-2007 also addressed by LCO 3.3.3.4344 4026 Serrice Water Sctvicc Waicr 4013 Sctaricc Waict Labe Oil Cooler 3652 9SISB tube Oin Cooler MDAFIY B 4291 r I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I 4031 4032 4310<<5 lani M 8 Steam Gcnctamt A 4352 4000A Q 4000B 4356 SIILm Gcactamt B 3505A I 3505 B Cp I--Mt-++--=-;I I 7 I To MSlV I 3512 I I I To<<MSW I SSI 6 I I CP I 3504 B 3504A LEGEND:-Flow path not required for LCO----Addressed in LCO 3.7.6------TDAFWflovvpath AFW Train(N<<ia

-TDAFW train includes both steam and both injection Qovvpaths)

Lube Oil Coolct 01 4431 443 CQ~th~ID Ul Q ScM 9626A 6 A 9629A SAFW Pump C 01A 4084 A 9702A 970 A 9706A Steam Genera'or A O tD SV O cD 8 to Ch M t<<2<<5 CZ'1 M tD~4J cn m lh Vl I cD SAFW Pump Room Cooling Unit 1B SAFW Pump Room Cooling Unit 1B 9708A I f 9707A I I I I 9728 I I Cond entete Tett Tank 9710A 9703B Service Water 1I622B 9627 B 9629B$9707B I I I 9708B I SAFW Pump D 9710B 9701B 4085¹6 9704B 9702B 9705B 9706B Stcam Generator B cD O Legend:-----Flow path not'retluired for LCO SAFW Train For illustration Gnl Tl CQ<i'~\h~rD Ul B CCW System 8 3.7.7 BASES BACKGROUND (continued)

The principal safety related function of the CCW System is'the removal of decay heat from the reactor via the Residual Heat Removal (RHR)System.Since the removal of decay heat via the RHR System is only performed during the recirculation phase of an accident, the CCW pumps do not receive an automatic start signal.Following the generation of a safety injection signal, the normally operating CCW pump will remain in service unless an undervoltage signal is'.present on either Class lE electrical Bus 14 or Bus 16 at which time the pump is stripped from its respective bus.A CCW pump can then be manually placed into service prior to switching to recirculation operations which would not be required until a minimum of 22.4 minutes following an accident.APPLICABLE SAFETY ANALYSES The design basis of the CCW System is for one CCW train and one CCW heat exchanger to remove the loss of coolant accident (LOCA)heat load from the containment sump during the recirculation phase.The Emergency Core Cooling System (ECCS)and containment models for a LOCA each consider the minimum performance of the CCW System.The normal temperature of the CCW is s 100 F, and, during LOCA conditions, a maximum temperature of 120 F is assumed.This prevents the CCW System from exceeding its design temperature limit of 200 F, and provides for a gradual reduction in the temperature of containment sump fluid as it is recirculated to the Reactor Coolant System (RCS)by the ECCS pumps.The CCW System is designed to perform its function with a single failure of any active component, assuming a coincident loss of offsite power.The CCW trains, heat exchangers, and loop headers are manually placed into service prior to the recirculation phase of an accident (i.e., 22.4 minutes following a large break LOCA).(continued)

R.E.Ginna Nuclear Power Plant B 3.7-47 Revision 1 717L 777F Sl PUMP A Sl PUMP 8 777M 7775 il tQ hC h g Lrl CQ trr 4rJ rD I Return Line From Non.Accident Loads I I I I I, I I I I~$728 I 122A I'I I I CCW'2&A Pump A CCW Pump 8 I A I I I I I For iHustration only LEGEND I I I I I I I I I I I CCW Loop Header I I CCW Train I I CCW heat exchanger,'

I I I I I I I I To Non Aeeldant y Loads Z 725 CCW HX 8 724A I I I 7338 I I 7348 I I I I I I I L J J~OO L I I CCW HX A I I I I-'l0-I rR-'248 133A I I 134A I~QQ i%%&7778 777J 777N 777H Sl PUMP C 777K 777R 777G 771P 777C CS PUMP 8 777D 164C I I I I I I I I I I I I 764D I I I I I 73&A HR LOOP A 780A 741A AHR LOOP A RHR LOOP 8 RHR t.OOP S 7388 7078 817 707A RHR PUMP 18 RHR PUMP 1A 7808 1418 769 7088 708A 14&A 750A ACP A 15&A 62A 7498 7508 742A 743 813 RCP 8 Excess~tdown HX Rx Support Cool~rs 7598 7628 qFC 745 7428 814 815 A CS PVMP A 7778 777A'I I I I I I I I I I I I I I I I I I I I I I I I I I I I I h G7 trl trJ LL~th~rD SW System B 3.7.8 BASES APPLICABLE SAFETY ANALYSES (continued)

The S'W trains and loop header are assumed to supply to following components following an accident: a.The CRFCs, DGs and safety injection pump bearing housing coolers immediately following a safety injection signal (i.e., after the loop header becomes refilled);

b.The preferred AFW and SAFW pumps within 10 minutes following receipt of a low SG level signal;and c.The CCW heat exchangers within 22.4 minutes following a safety injection signal.The SW system, in conjunction with the CCM System, can also cool the plant from residual heat removal (RHR)entry conditions (T.,<350 F)to MODE 5 (T.,<200 F)during normal operations.

The time required to cool from 350 F,to 200 F is a function of the number of CCW and RHR System trains.that are operating.

Since SW is comprised of a large loop header, a.passive failure can be postulated during this cooldown period which results in failing the SW System to potentially multiple safety related functions.

The SW system has been evaluated to demonstrate the capability to meet cooling needs with an assumed 500 gal leak.The SM System is also vulnerable to external events such as tornados.The plant has been evaluated for the loss of SW under these conditions with the use of alternate cooling mechanisms (e.g., providing for natural circulation using the atmospheric relief valves and the AFM Systems)with acceptable results (Ref.I).The temperature of the fluid supplied by the SW System is also a.consideration in the accident analyses.If the cooling water supply to the containment recirculation fan coolers and CCW heat exchangers is too warm, the accident analyses with respect to containment pressure response following a SLB and the containment sump fluid temperature following a LOCA may no longer be bounding.If the cooling water supply is too cold, the containment heat removal systems may be more efficient than assumed in the accident analysis.This causes the backpressure in containment to be reduced which potentially results in increased peak clad temperatures.

(continued)

R.E.Ginna Nuclear Power Plant B 3.7-57 Revision I rr7 To Cireuiatmg SVatcr Pumps And Travelling Screens 3 rD t/)Cc3~GJ trt J~4609 StV Pump A~4730 9 r SW Pump B 4602 4606 I I X4612 461 3 StV Pump C I pr I 202S I S'tV Pump D 604 460 Legend:~StV Pump Train(one pump l'rom each clcctricat vain faired)--~SW Loop Header To Sl pumps (LCD 3')and Safety Related Pump Room Coolers 4623 4139 4139 4640 t To Motor&van AFW Pumps (LCD 3.78)I I 4'733 4790 To Sl pumps(LCO382) and Safety Related Pump Room Coolers I I 4663 P~To Non Safety s~610 i~l~ReiatedLoads I (Station Air)L I To Direct M&0 I h 4667 (LCD 3.g.l)I I I$4559 I I I I To Diesel c j~W Ocncrator B (LCO 3.S.I)466SB To SAFW Pump C To CCWHXA and SAFWRootn (LCO 3.7.7)and Cooler A (LCO 3.7S)Spent Fuel Pool HXA A 4 N27h~4133 I N26A I I s I I I I h.A 4616 I I I I I t$4670 To SAFW pump D and.SAFW Room Coo!ca B 9d27B 4779 9626B I s 0-P'4734 46IS To CCtV HX B (LCD 3.7.7)and Spent Foci Pool HX B To Hon Safety Rclatcd Loads (th Com>>store)

I I To TDAFW Pump (LCD 3.73)I I I I I To CRFC Unit A 0.CO 3.tL6)I I I I To CRFC Unit B (LCO 35.6)"------.Mw I I I I I I g 4736 I I I I g 4639 I I To CRFC s Unit C (LCO 3 6 6)I I I To CRFC s Unit D(LCD 3.6.6)I r-M I I I I I I I I 4663 e I DP 4733 y To Xon Safety Rcbtcd Loads (Killers)CQ Lh tas<~th~rD CO 6 For illustration only I I AC Sources-HODES.1, 2, 3, and 4 B 3.8.1 BASES APPLICABLE SAFETY ANALYSES (continued)

DG Load DG A Time 480V safeguards buses and CS pumps 10 SI pump A and'B 10 SI pump C 15 Residual heat removal pump 20 Selected service water pump 25 First containment recirculatio'n fan cooler 30 Second containment recirculation fan cooler 35 Hotor'riven auxiliary feedwater pump 40 The pumps and fans are assumed to be running within 5 seconds following breaker closure.DG B Time 10 10 17 22 27 32 37 42 Since the DGs must start and begin loading within 10 seconds, only one air start must be available in the air receivers as assumed in the accident analyses.The long term operation of the DGs (until offsite power is restored)is discussed in LCO 3.8.3,"Diesel Fuel Oil." The AC sources satisfy Criterion 3 of NRC Policy Statement.

LCO One qualified independent offsite power circuit connected between the offsite transmission network and the onsite 480 V safeguards buses and separate and independent DGs for each train ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an AOO or a postulated DBA.An OPERABLE qualified independent offsite power circuit is one that is capable of maintaining rated voltage, and accepting required loads during an accident, while connected to the,480 V safeguards buses required by LCO 3.8.9,"Distribution Subsystems-HODES 1, 2, 3, and 4." Power from either offsite power circuit 751 or 767 satisfies this requirement.(continued)

R.E.Ginna Nuclear Power Plant B 3.8-7 Revision 1 BASES AC Sources&ODES 1, 2, 3, and 4 8 3.8.1 LCO'continued)

A DG is considered OPERABLE when: 'a~The DG is capable of starting, accelerating to rated speed and voltage, and connecting to its respective 480 V safeguards buses on detection of bus undervoltage within 10 seconds;(c'ontinued)

R.E.Ginna Nuclear Power Plant 8 3.8-7a Revision 1 AC Sources-NODES 1, 2, 3, and 4 B 3.8.1 BASES LCO (continued) b.All loads on each 480 V safeguards bus except for the safety r elated motor control centers, CCW pump, and CS pump are capable of being tripped on an undervoltage signal (CCW pump must be capable of being tripped on coincident SI and undervoltage signal);C.The DG is capable of accepting required loads both manually and within the assumed loading sequence intervals following a coincident SI and undervoltage signal, and continue to operate until offsite power can be restored to the safeguards bus (i.e., 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />);d.The DG day tank is available to provide fuel oil for a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 110/design loads;e.The fuel oil transfer pump from the fuel oil storage tank to the associated day tank is OPERABLE including all required piping, valves, and instrumentation (long-term fuel oil supplies are addressed by LCO 3.8.3,"Diesel Fuel Oil");and f.A ventilation train consisting of at least one of two fans and the associated ductwork and dampers is OPERABLE.g.The service water (SW)~p through the diesel generator heat exchangers is<31 psid with two SW pumps operating and<44 psid with three SW pumps operating.

The AC sources in one train must be separate and independent of the AC sources in the other train.For the DGs, separation and independence must be complete assuming a single active failure.For the independent offsite power source, separation and independence are to the extent practical (i.e., oper ation is preferred in the 50/50 mode, but may also exist in the 100/0 or 0/100 mode).APPLICABILITY The AC sources are required to be OPERABLE in NODES 1, 2, 3, and 4 to ensure that: (continued)

R.E.Ginna Nuclear Power Plant B 3.8-8 Revision 1 i AC Sources-NODES 1, 2, 3, and 4 B 3.8.1 BASES APPLICABILITY a.Acceptable fuel design limits and r eactor coolant (continued) pressure boundary, limits are not exceeded as a result of AOOs or abnormal transients; and(continued)

R.E.Ginna Nuclear Power Plant B 3.8-8a Revision 1

\

AC Sources-NODES 5 and 6 B 3.8.2 BASES LCO (continued)

A DG is considered OPERABLE when: a~b.C.d.e.The DG is capable of starting, accelerating to rated speed and voltage, and connecting to its respective 480 V safeguards buses on detection of bus undervoltage within 10 seconds;All loads on each 480 V safeguards bus except for the safety related'motor control centers, component cooling water (CCW)pump, and containment spray (CS)pump are capable of being tr ipped on an undervoltage signal (CCW pump must be capable of being tripped on coincident safety inje'ction (SI)and undervoltage signal);The DG is capable of accepting required loads manually.Since most equipment which receives a SI signal are isolated in these MODES due to maintenance or low temperature over pressure protection concerns, and the DBA of concern (i.e., a fuel handling accident)would not generate a SI signal, manual loading of the DGs will most likely be required.These loads must be capable of being added to the OPERABLE DG within 10 minutes;The DG day tank is available to provide fuel oil for z 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 110%design loads;The fuel oil transfer pump from the fuel oil storage tank to the associated day tank is OPERABLE including all required piping, valves, and instrumentation (long-term fuel oil supplies are addressed by LCO 3.8;3,"Diesel Fuel Oil");and A ventilation train consisting of at least one of two fans and the associated ductwork and dampers is OPERABLE.g, The service water (SW)~p through the diesel generator heat exchanger is<31 psid with two SW pumps operating and<44 psid with three SW pumps operating.

R.E.Ginna Nuclear Power Plant B 3.8-27 (continued)

Revision 1

Til 4100V BUS 12A 4180V BUS 12 B)STATION SERViCE TRANSFORMEA NO.Ia STATION SERVICE'TRANSFORMER NO.18 480V BUS Ia DQ A T.S.C.VITAL BAiTEAY DQ B BUS 18 Tl CO (D CO 4J CJ1 OD I I BATTERY i ii CHARGE I IAI I MCC C BATIEAY CHARGER IA DIST.PANEL A INVERT.A 125.V BATTERY A RISE CABINET A A BAiT.OISCON.i SWITCH i B BAiT.T.S.C.DISCOH.BATTEAY DISCONNECT SWITCH T.S.C.125 V VITAL BATT.BATTERY MANUAL B THAOWOVER SWITCH FUSE CABINET B BATTERY CHARGEA 1B DIST.PANEL B INVERTS B I BATTERY CHARGER 1S1 I MCC B MCC A EM&.INSTIL TRANSFORMEA 78 KVA 120 VOLT AUTO STATIC TRANSFER A 7.5 KVA 110 VOLT CONST.VOLTAGE 1RINSFDAMEII A)AUTO L STAilC TRANSFER B 7S KVA 110 VOLT CONST.VOLTAGE TRANSFORMER B ID O INSTR.BUS A+NORMAILYOPEH WHEN Tavy>>200 F INSTR.BUS e DC SOUACE t DC DIST.SYSTEMS g a'DC ELEC.POWER SOURCES INSTR.BUS C INSTTL BUS D INST>>BUS POWER SOURCES GJ gu CO CL Distribution Systems-MODES 5 and 6 B 3.8.10 BASES (continued)

LCO Various combinations of AC, DC, and AC instrument bus electrical power distribution subsystems, trains within these subsystems, and equipment and components within.these trains are required OPERABLE by other LCOs, depending on the specific plant condition.

Implicit in those requirements is the required OPERABILITY of necessary support featu}es.This LCO explicitly requires energization of the portions of.the electrical distribution system necessary to support OPERABILITY of required systems, equipment, and components-all specifically addressed in each LCO and implicitly required via the definition of OPERABILITY.

The LCOs which apply when the Reactor Coolant System is s 200'F and which may require a source of electrical power are: LCO 3.1.1 LCO 3.3.1 LCO 3.3.4 LCO 3.3.5 LCO 3.3.6 LCO 3.4.7 LCO 3.4.8 LCO 3.4.12 LCO 3.7.9 LCO 3.9.2 LCO 3.9.4 LCO 3.9.5 SHUTDOWN MARGIN (SDM)Reactor Trip System (RTS)Instrumentation Loss of Power (LOP)Diesel Generator (DG)Start Instrumentation Containment Ventilation Isolation Instrumentation Control Room Emer'gency Air Treatment System (CREATS)Actuation RCS Loops-MODE 5, Loops filled RCS Loops-MODE 5, Loops Not Filled Low Temperature Overpressure Protection (LTOP)System Control Room Emergency Air Treatment System (CREATS)Nuclear Instrumentation Residual Heat Removal (RHR)and Coolant Circulation

-Water Level z 23 Ft Residual Heat Removal (RHR)and Coolant Circulation

-Water Level.<23 Ft Maintaining the necessary trains of the AC, DC, and AC instrument bus electrical power distribution subsystems energized ensures the availability of sufficient power to operate the plant in a safe manner to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents).(continued)

R.E.Ginna Nuclear Power Plant B 3.8-86 Revision 3 l 1 Distribution Systems-MODES 5 and 6 B 3.8.10 BASES LCO (continued)

Bus-tie breakers required to be open during MODES 1, 2, 3, and 4 per SR 3.8.9.1 may be closed during MODES 5 and 6 provided that the distribution system alignment continues.to'-support systems necessary to mitigate the postulated events assuming either a loss of all offsite power, loss of all onsite DG power, or a worst case single failure.The postulated events during MODES 5 and 6 include a boron dilution event and fuel handling accident.Examples of allowed configurations are as follows (note that other configurations are acceptable provided that they meet the above criteria):

'a~b.Bus-Tie Breakers 16-15.and 14-13 (and their associated"dummy" breakers on non-safeguards Buses 13 and 15)provide the capability to cross-tie the safeguards and non-safeguards 480 V buses.Closure of these bus-ties is allowed provided that the OPERABLE DG per LCO 3.8.2 can accept all loads which would be automatically loaded from the safeguards and non-safeguards buses, and accept those loads which must be manually loaded to mitigate the accident.Bus-Tie Breakers 14-16, 16-14, and 17-18 provide the capability to cross-tie the two safeguard electrical trains.Closure of these bus-ties is allowed provided that the OPERABLE DG per LCO 3.8.2 can accept all loads which would be automatically loaded, and accept those loads which must be manually loaded to mitigate the accident.In addition, the automatic trip logic of the bus-ties due to an undervoltage signal from either of the two cross-tied buses must be OPERABLE.This trip logic ensures that upon a fault of either 480 V safeguards bus as the single failure, the redundant bus is capable of mitigating the accident using either the DG or offsite power.R.E.Ginna Nuclear Power Plant B 3.8-87 (continued)

Revision 3 Distribution Systems-MODES 5 and 6 B 3.8.10 BASES (continued)

APPLICABILITY The AC, DC, and AC instrument bus electrical power distribution subsystems required to be OPERABLE in MODES 5 and 6 provide assurance that systems required to mitigate the effects of a postulated event and maintain the plant in the cold shutdown or refueling condition are available.

The AC, DC, and AC instrument bus electrical power distribution subsystems requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.9,"Distribution Systems-MODES 1, 2, 3, and 4." ACTIONS A.l Although redundant required features may require redundant trains of electrical power distribution subsystems to be OPERABLE, one OPERABLE distribution subsystem train may be capable of supporting sufficient required featur es to allow continuation of CORE ALTERATIONS and operations involving positive reactivity additions.

By allowing the option to declare required features associated with an inoperable distribution subsystem or train inoperable, appropriate restrictions are implemented in accordance with the LCO ACTIONS of the affected required features.A.2.1 A.2.2 A.2.3 A.2.4 and A.2.5 With one or more required electrical power distribution subsystems or trains inoperable, the option exists to declare all required features inoperable per Required Action A.l.Since this option may involve undesired administrative efforts, the allowance for sufficiently conservative actions is made.Therefore, immediate suspension of CORE ALTERATIONS, movement of irradiated fuel assemblies, and operations involving p'ositive reactivity additions is an acceptable option to Required Action A.l.Performance of Required Actions A.2.1, A.2.2, and A.2.3 shall not preclude completion of movement of a component to a safe position or'ormal co'oldown of the coolant volume for the purpose of system temperature control within established procedures.(continued)

'.E.Ginna Nuclear Power Plant B 3.8-88 Revision 3 Distribution Systems-MODES 5 and 6 B 3.8.10 BASES ACTIONS A.2.1 A.2.2 A.2.3 A.2.4 and A.2.5 (continued)

It is further required to immediately initiate action to restore the required AC, OC, and AC instrument bus electrical power distribution subsystems and to continue this action until restoration is accomplished in order to provide the necessary power to the plant safety systems.In addition to performance of the above conservative Required Actions, a required residual heat removal (RHR)loop may be inoperable.

In this case, Required Actions A.2.1, A.2.2, A.2 3, and A.2.4 do not adequately address the concerns relating to coolant circulation and heat removal.Pursuant to LCO 3.0.6, the RHR ACTIONS would not be entered.Therefore, Required Action A.2.5 requires declaring RHR inoperable, which results in taking the appropriate RHR actions.The Completion Time of immediately is consistent with the required times for actions requiring prompt attention.

The restoration of the required distribution subsystems should be completed as quickly as possible in order to minimize the time the plant safety systems may be without power.R.E.Ginna Nuclear Power Plant B 3.8-89 (continued)

Revision 3

'i0 Distribution Systems-NODES 5 and 6 B 3.8.10 BASES (continued)

SURVEILLANCE RE(UIREHENTS SR 3.8.10.1 This Surveillance verifies that the electrical power distribution trains are functioning properly, with all the required power sour'ce circuit breakers closed, required tie-breakers open, and the required buses energized from their allowable power sources.Required voltage for the AC power distribution electrical subsystem is z 420 VAC, for the DC power distribution electrical subsystem a 108.6 VDC, and for AC instrument bus power distribution electrical subsystem is between 113 VAC and 123 VAC.Required voltage for the twinco panels supplied by the 120 VAC instrument buses is between 115.6 VAC and 120.4.VAC.The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses.The Frequency of 7 days takes into account the capability of the AC, DC, and AC instrument bus electrical

.power distribution subsystems, and other indications available in the control room that alert the operator to subsystem malfunctions.

REFERENCES None.R.E.Ginna Nuclear Power Plant B 3.8-90 Revision 3 Nuclear Instrumentation B 3.9.2 BASES (continued)

LCO This LCO requires two source range neutron flux monitors be OPERABLE to ensure redundant monitoring capability is available to detect changes in core reactivity.

To be'PERABLE, each monitor must provide visual indication and at least one of the two monitors must provide an audible count rate function in the control room.Mith the discharge of fuel from core positions adjacent to source range detector locations, counts decreasing to zero is the expected response.Based on this indication alone, source range detection should not be considered inoperable.

Following a full core discharge, source range response is verified with the initial fuel assemblies reloaded.APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity.

There are no other direct means available to check core reactivity conditions in this MODE.In MODES 2, 3, 4, and 5, these same installed source'range detectors and circuitry are also required to be OPERABLE by LCO 3.3.1,"Reactor Trip System (RTS)Instrumentation." ACTIONS A.l and A.2 Mith only one source range neutron flux monitor OPERABLE, redundancy has been lost.Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and positive reactivity additions must be suspended immediately.

Performance of Required Actions A.l and A.2 shall not preclude completion of movement of a component to a safe position (i.e., other than normal cooldown of the coolant volume for the purpose of system temperature control within established procedures).(continued)

R.E.Ginna Nuclear Power Plant 8 3.9-7 Revision 1

\\0 i ep Nuclear Instrumentation B 3.9.2 BASES ACTIONS (continued)

B.l and B.2 Mith no source range neutron flux monitor OPERABLE there are no direct means of detecting changes in cove reactivity.

Therefore, actions to restore a monitor to OPERABLE status shall be initiated immedi'ately and continue until a source range neutron flux monitor is restored to OPERABLE status.(continued)

R.E.Ginna Nuclear Power Plant B 3.9-7a Revision I Nuclear Instrumentation 8 3.9.2 BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.9.2.1 Thi s SR is the performance of a CHANNEL CHECK, which i s a comparison of the parameter indicated on one monitor to a similar parameter on another monitor.It is based on the assumption that the two indication channels should be consistent with core conditions.

Changes in fuel loading and core geometry can result in significant differences between source range monitors, but each monitor should be consistent with its local conditions.

The inoperability of one source range neutron flux channel prevents performance of a CHANNEL CHECK for the operable channel.However, the Required Actions for the inoperable channel requires suspension of CORE ALTERATIONS and positive reactivity addition such that the CHANNEL CHECK of the operable channel can consist of ensuring consistency with known core conditions.

The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.1,"Reactor Trip System (RTS)Instrumentation." SR 3.9.2.2 This SR is the performance of a CHANNEL CALIBRATION every 24 months.This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.

The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to baseline data.The 24 month Frequency is based on the need to perform this Surveillance

.unde}the conditions that apply during a plant outage.Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency.

REFERENCES l.UFSAR,.Section 7.7.3.2.2.Atomic Industrial Forum (AIF)GDC 13 and 19, Issued for Comment July 10, 1967.R.E.Ginna Nuclear Power Plant B 3.9-9 F Revision 1 Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASESBACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be restricted from.escaping to the environment when the LCO requirements are met.In MODES I, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1,"Containment." In MODE 5, there are no accidents of concern which require containment.

-In MODE 6, the potential for containment pressurization as a result of an accident is not likely;therefore, requirements to isolate the containment from the outside atmosphere can be less stringent.

The LCO requirements are referred to as"containment closure" rather than"containment OPERABILITY." Containment closure means that all potential escape paths are closed or capable of being closed.Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained within the requirements of'10 CFR 100.Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a=means for moving large equipment and components into and out of containment.

During CORE ALTERATIONS or m'ovement of irradiated fuel assemblies within containment, the equipment hatch must be bolted in place.Good engineering practice dictates that.a minimum of 4 bolts be used to hold the equipment hatch in place and that the bolts be approximately equally spaced.As an alternative, the equipment hatch opening can be isolated by a closure plate that restricts air flow from containment or by an installed roll up door and enclosure building.(continued)

R.E.Ginna Nuclear Power Plant B 3.9-l0 Revision 2 Containment Penetrations 8 3.9.3 BASESBACKGROUND (continued)

The containment equipment and personnel air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES I, 2, 3, and 4 in accordance with LCO 3.6.2,"Containment Air Locks." Each air lock has a door at both ends.The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required.During periods of plant shutdown when containment closure is not required, the door inter lock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure is required;therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed in the personnel and equipment hatch (unless the equipment hatch is isolated by a closure plate or the roll up door and associated enclosure building).

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted from escaping to the environment.

The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident during refueling.

The Containment Purge and Exhaust System includes two subsystems.

The Shutdown Purge System includes a 36 inch purge penetration and a 36 inch exhaust penetration.

The second subsystem, a Mini-Purge System, includes a 6 inch purge penetration and a 6 inch exhaust penetration.

During MODES I, 2, 3, and 4, the shutdown purge and exhaust penetrations are isolated by a blind flange with two 0-rings that provide the necessary boundary.The two air operated valves in each of the two mini-purge penetrations can be opened intermittently, but are closed automatically by the Containment Ventilation Isolation Instrumentation System.Neither of the subsystems is subject to a Specification in MODE 5.(continued)

R.E.Ginna Nuclear Power Plant B 3.9-11 Revision 2 e

Containment Penetrations B 3.9.3 BASES (continued)

LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment.

The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and exhaust penetr ations.For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that at least one valve in each of these penetrations is'solable by the Containment Ventilation Isolation System.APPLICABILITY The containment penetration'equirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident.In MODES I, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1.In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment.

are not being conducted, the potential for a fuel handling accident does not exist.Therefore, under these conditions, no requirements are placed on containment penetration status.ACTIONS A.l and A.2 If the containment equipment hatch (or its closure plate or ro11 up door and associated enclosure building), air lock doors, or any'ontainment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Ventilation Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the plant must be placed in a condition where the isolation function is not needed.This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment.

Performance of these actions shall not preclude completion of movement of a component to a safe position.'.E.Ginna Nucleal Power Plant B 3.9-13 (continued)

Revision 2 C W'P I