ML17264A471

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Proposed Tech Specs,Correcting Typos
ML17264A471
Person / Time
Site: Ginna Constellation icon.png
Issue date: 05/08/1996
From:
ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A469 List:
References
NUDOCS 9605100242
Download: ML17264A471 (31)
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Text

7 TABLE OF CONTENTS 1.0 USE ANO APPLICATION 1.1-1 1.1 Definitions ~ ~ ~ ~ ~ ~ 1.1-1 1.2 Logical Connectors . 1.2-1 1.3'.4 Completion Times . 1.3-1 Frequency 1.4-1 2.0 SAFETY LIMITS, (SLs) 2.0-1 2.1 S Ls ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2.0-1 2.2 SL Violations 2.d-l 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS . 3 '-1 3.1.1 SHUTDOWN MARGIN (SOM) 3.1-1 3.1.2 Core Reactivity '.1-2.

3.1.3 Moderator Temperature Coefficient (HTC) 3.1-4 3.1.4 Rod Group Alignment Limits . . 3.1-7 3.1.5 Shutdown Bank Insertion Limit 3.1-11 3.1.6 Control Bank Insertion Limits 3.1-13 3.1.7 Rod Position Indication 3.1-15 3.1.8 PHYSICS TESTS Exceptions -MODE 2 . 3.1-18 3.2 POWER DISTRIBUTION LIHITS 3.2-1

3. 2'. 1 Meat F1ux Hot Channel Factor (Fa(Z)) ~ ~ ~ ~ 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F"z ~) 3.2-4 3.2.3 AXIAL FLUX DIFFERENCE (AFO) ~ ~ ~ ~ 3.2-6 3.2.4 QUADRANT POWER TILT RATIO (gPTR) 3. 2- 1'1 3.3 INSTRUMENTATION 3.3-1 3.3.1 Reactor Trip System (RTS) Instrumentation 3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation 3.3-20 3.3.3 Post Accident Monitoring (PAH) Instrumentation . 3.3-28 3.3.4 Loss of Power (LOP) Diesel Generator {DG) Start Instrumentation 3.3-34 3.3;5 Containment Ventilation Isolation 3.3-36 3.3.6 Control Room Emergency Air Treatment System

{CREATS) Instrumentation Actuation . . ~ . . . . 3.3-9+o~zooo42 ~6ooos PDR ADOCK 05000244 P PDR (continued)

R.E. Ginna Nuclear Power Plant Amendment No. 61

TABLE OF CONTENTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.4-1 3.4.2 RCS Minimum Temperature for Criticality 3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4-4 3.4.4 RCS Loops -MODE 1 > 8.5% RTP . 3.4-6 3.4.5 RCS Loops MODES 1 s 8.5% RTP, 2, and 3 3.4-7 3.4.6 RCS Loops MODE 4 3.4-10 3.4.7 RCS Loops MODE 5, Loops Filled 3.4-13 3.4.8 RCS Loops MODE 5, Loops Not Filled 3.4-16 3.4.9 Pressurizer 3.4-18 3.4.10 Pressurizer Safety Valves 3.4-20 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4-22 3.4.12 Low Temperature Overpressure Protection (LTOP) Sys tern 3.4-26

.3.4.13 RCS Operational LEAKAGE 3.4-32 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage . 3.4-34 3.4.15 ,RCS Leakage Detection Instrumentation 3.4-38 3.4.16 RCS Specific Activity 3.4-42 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5-1 3.5.1 Accumulators . 3.5-1 3.5.2 ECCS MODES 1, 2, and 3 3.5-3 3.5.3 ECCS- HODE 4 3.5-6 3.5.4 Refueling Water Storage Tank (RWST) 3.5-8 3.6 CONTAINMENT SYSTEMS 3.6-1 3.6 ' Containment 3.6-1 3.6.2 Containment Air Locks 3.6-3 3.6.3 Containment Isolation Boundaries . 3.6-8 3.6.4 'Containment 3.6-~ '<

3.6.5 Air Temperature Pressure'ontainment 3.6-~ l~

3.6.6 Containment Spray (CS), Containment Recirculation Fan Cooler (CRFC), NaOH, and Containment Post-Accident' Charcoal Systems . '. . . . . . . . . . . . . . . 3.6-~<<

3.6.7 Hydrogen Recombiners . . . . . . . . . . . . . . . 3.6-~ e0 3.7 PLANT 'SYSTEMS 3 '-1 3.7.1 Hain Steam Safety Valves (HSSVs) 3.7-1 3.7.2 Hain Steam Isolation Valves (HSIVs) and Non-Return Check Valves 3~7 3

)~

~ ~ ~

3.7.3 Hain Fee er um ischarge Valve H PDVs 3.7.4 At

~ain Feedwater Regu a ing .a ves (MFRVs, o p Associated B ass Valves ic e ref alves (A Vs) 3.7-5 3.7-8 3.7-10 3.7.5 Auxiliary Feedwater (AFW) System .

3.7.6 Condensate Storage Tanks (CSTs) 3.7-14 3.7.7 Component Cooling Water (CCW) System . 3.7-15 3.7.8 Service Water (SW) System 3. 7-4 (continued)

R.E. Ginna Nuclear Power Plant Amendment No. 61

TABLE OF CONTENTS 3.7 PLANT SYSTEMS (continued) 3.7.9 Control Room Emergency Air Treatment System (CREATS) 3. 7-kP-~

3.7.10 Auxiliary Building Ventilation System (ABVS) 3.7-~~"

3.7.11 Spent Fuel Pool (SFP) Water Level 3.7-R&- >>

3.7.12 Spent Fuel Pool (SFP) Boron Concentration 3. 7-Z&-~>

3.7.13 Spent Fuel Pool (SFP) Storage 3. 7-&- <~

3.7.14 I

Secondary Specific Activity 3.7~Du 3.8 ELECTRICAL POWER SYSTEMS . 3.8-1 3.8. AC Sources -MODES 1, 2, 3, and 4 . 3.8-1 1'.8.2 AC Sources -MODES 5 and. 6 3.8-8 3.8.3 Diesel Fuel Oil 3.8-11 3.8.4 DC Sources -MODES 1, 2, 3, and 4 . 3.8-13 3.8.5 DC Sources -MODES 5 and 6 3.8-16 3.8.6 Battery Cell Parameters ~ ~ ~ ~ 3.8-18 3.8.7 AC Instrument Bus Sources -MODES 1, 2, 3 , and 4 3.8-20 3.8.8 AC Instrument Bus Sources -MODES 5 and 6 3.8-22 3.8.9 Distribution Systems MODES 1, 2, 3, and 4 3 '-24 3.8.10 Distribution Systems -MODES 5 and 6 3.8-26 3.9 REFUELING OPERATIONS . 3.9-'1 3.9.1 Boron Concentration 3.9-1 3.9.2 Nuclear Instrumentation 3.9-2 3.9.3 Containment Penetrations . 3.9-4 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation Water Level ~ 23 Ft 3.9-6 3.9.5 Residual Heat Removal.(RHR) and Coolant Circulation Water Level < 23 Ft 3.9-8 3.9.6 Refueling Cavity Water Level 3.9-10 4.0 DESIGN FEATURES 4.0-1, 4.1 Site Location 4.0-1 4.2 Reactor Core . 4.0-1 4.3 Fuel Storage . 4.0-2 5.0 ADMINISTRATIVE CONTROLS ~ ~ 5.0-1 5.1 Responsibility . ~ ~ 5.0-1 5.2 Organization . 5.0-2 5.3 Plant Staff gualificati ons 5.0-4 5.4 Procedures . 5.0-5 5.5= Programs and Manuals . 5.0-6 5.6 Reporting Requirements 5.0-18 5.7 High Radiation Area 5.0-23 R.E. Ginna Nuclear Power Plant Amendment No. 61

gPTR 3.2.4 ACTIONS CONDITION 'E(UIRED ACTION COMPLETION TIME A. (continued) A.6 --------NOTES--------

1. Only required to be performed if the cause of the gPTR alarm is not associated with inoperable gPTR instrumentation.
2. Required

'Action A.6 must be completed when Required Action A.5 is completed and Note 1, above, does not apply.

3. Only one of the Completion Times, whichever becomes applicable first, must be met.

'Perform SR 3.2.1.1 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and SR 3.2.2. 1. after reaching RTP OR Within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasing T ERMAL POWER above the l>mits of Required Actions A.l and A.2 (continued)

R.E. Ginna Nuclear Power Plant 3.2-13 Amendment No. 61

RTS Instrumentation 3.3.1 ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME met.'.l Required Action and Associated Completion Time not of Condition P AND Reduce THERMAL to < 50% RTP.

POWER 6 hours Q.2.1 Verify Steam Dump 7 hours System is OPERABLE.

OR Q.2.2 Reduce THERMAL POWER 7 hours to < 8% RTP.

R. As required by R.1 --------NOTE---------

Required Action A.l train and referenced by may be ypassed for Table 3.3.1-1. up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing rgb the other train.

OPIA QgP Restore train to 6 hours OPERABLE status.

S. As required by S.1 Verify interlock is hour Required Action A.l in required state for and referenced by existing plant Table 3.3.1-1.. conditions' OR S.2 Declare associated 1 hour RTS Function channel(s)'noperable.

(continued)

R.E. Ginna Nuclear Power Plant 3.3-7 Amendment No. 61

RTS Instrumentation 3.3.1 Table 3.3. 1-1 (page 5 of 6)

Reactor Trip System Instrumentation Note 1: Overtem erature aT The Overtemperature aT Function Trip Setpoint is defined by:

-P') -K, (T - T') 1 + r,s Overtceperature h 7 Sh Te K,+K, (P

+ 1'P f(h I)

Where: ~

aT is measured RCS zT, F.

GTO is the indicated zT at RTP, F.

s is the Laplace transform operator, sec '.

T is the measured RCS average temperature, 'F.

T is the nominal T., at RTP, 'F.

P is the measured pressurizer pressure, psig.

P is the nominal RCS operating pressure, psig.

K, is the Overtemperature aT reactor trip setpoint, 1.20.

K2 is the Overtemperature zT reactor trip depressurization setpoint penalty coefficient, 0.000900.

K~ is the Overtemperature aT reactor trip heatup setpoint penalty coefficient, 0.0209.

~, is the measured lead/lag time constant, 25 seconds.

r2 is the measured lead/lag time constant, 5 seconds.

f(sI)'s .a function of the indicated difference between the top and bottom

~

detectors of the Power Range Neutron Flux channels where q, and q, are the percent power in the top and bottom halves of the core, respectively, and q, + P, is the total THERMAL POWER in 'percent RTP.

Q(

3 ~ ll f(z I) = 0 when q, - q, is~+13% RTP

~(~~) = @RE (q, - ~.) - 13I when q, - q, is > +13% RTP

~O. 0 <Z R.E. Ginna Nuclear Power Plant 3.3-18 Amendment No. 61 I

L

, RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)

Reactor Trip System Instrumentation Note 2: Over ower aT The Overpower aT Function Trip Setpoint is defined by:

r~sT Overpower h T Sb T K,-K (T-Tl) -K -f(h I) f~S +

Where:

aT is measured RCS sT, F.

zTo is the indicated aT at RTP, 'F.

s .is the Laplace transform operator, sec' T is the measured RCS average temperature, 'F ~

T is the nominal T,, at RTP, F.

K4 is the Overpower aT reactor trip setpoint, 1.077.

Ks is the Overpower aT reactor trip heatup setpoint penalty coefficient which is:

0.0 for T < T'nd; 0.0011 for T z T2.

K6 is the Overpower zT reactor trip thermal time delay setpoint penalty which is:

0.0262 for increasing T and; 0.00 for decreasing T.

r3 is the measured'lead/lag time constant, 10 seconds.

f(aI) is a function of the indicated difference between the top and bottom detectors of the Power Range Neutron Flux channels where,q, and q, are the percent power in the top and bottom halves of the core, respectively, and q, + q, is the total THERMAL POWER in percent RTP.

f(61) = 0 when q, - q, i +13% RTP T(a() = gP( (q, - q,) - 13I when q, - q, is > +13% RTP c) c)<3 R.E. Ginna Nuclear Power Plant 3.3-19 Amendment No. 61

ESFAS Instrumentation 3.3.2 SURVEILLANCE REgUIRENENTS

NOTE8-

~ Refer to Table 3.3.2-1 to determine which Function.~

~ ~ SRs apply for each ESFAS SURVEILLANCE FRE(UENCY SR 3.3.2. 1 Perform CHANNEL CHECK. 12 hours SR 3.3.2.2 Perform COT. 92 days SR 3.3.2.3 NOTE-Verification of relay setpoints not required.

Perform TADOT. 92 days SR 3.3.2.4 -NOTE Verification of relay. setpoints not required.

Perform TADOT. 24 months SR 3.3.2.5 . Perform CHANNEL CALIBRATION. 24 months SR 3.3.2.6 Verify the Pressurizer Pressure Low and 24 months Steam Line Pressure Low Functions are not bypassed when pressurizer pressure > 2000 Pslg.

SR 3.3.2.7 Perform ACTUATION LOGIC TEST: 24 months

\

R.E. Ginna Nuclear Power Plant 3.3-24 Amendment No. 61

PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2)

Post Accident Monitoring Instrumentation 1

RE(UIRED

,FUNCTION CHANNELS CONDITION

1. Pressurizer Pressure
2. Pressurizer Level
3. Reactor Coolant System (RCS) Hot Leg 1 per loop Temperature
4. RCS Cold Leg Temperature 1 per loop
5. RCS Pressure (Wide Range) 2
6. RCS Subcooling Monitor
7. Reactor Vessel Mater Level
8. Containment Sump B Water Level
9. Containment Pressure (Wide Range)
10. Containment Area Radiation (High Range) 2 Hydrogen Monitors 2
12. Condensate Storage Tank Level 2
13. Refueling Mater Storage Tank Level .2
14. Residual Heat Removal Flow 2 G
15. Core Exit Temperature quadrant 1 2(a)
16. Core Exit Temperature quadrant 2 2(a) G
17. Core Exit Temperature quadrant 3 2(a)
18. Core Exit Temperature- quadrant 4 2(a)
19. Auxiliary Feedwater (AFM) Flow to Steam 2 Generator (SG)

~

A'FW

20. F ow to SG B
21. SG Wa er Level (Narrow Range)

G'continued)

22. SG Water Level (Narrow Range)

(a} A channel consists of two core exit thermocouples (CETs).

R.E. Gihna Nuclear Power Plant 3.3-32 Amendment No. 61

PAM Instrumentation 3;3.3 Table 3.3.3-1 (page 2 of 2)

Post Accident Monitoring Instrumentation

~

RE(VIREO

~

FUNCTION CHANNELS CONDITION 23.

24.

25.

26.

SG SG SG Wa assure SG+P)%ssure

~

er Level (Wide Range)

~

er Level (Wide Range) G

.G G

R.E. Ginna Nuclear Power Plant 3.3-33 Amendment No. 61

Containment Ventilation Isolation Instrumentation 3.3.5 Table 3.3.5-1 (page 1 of 1)

Contaiment Ventilation Isolation Instrunentat ion SURVEILLANCE FUNCTIOH REQUIRED CHANNELS TRIP SETPOIHT REQUIREHENTS

1. Automatic Actuation Logic and 2 trains SR 3.3.5.3 HA Actuation Relays
2. Contairnw.nt Radiation
a. .Gaseous SR 3.3.5.1 (a)

SR 3.3-5.2 SR 3.3.5.4

b. Particulate SR 3.3.5.1 (a)

SR 3.3.5.2 SR 3.3.5.4

3. Contairmcnt Isolation Refer to LCO 3.3.2, <<ESFAS Instruaentation,<< Function 3, for all initiation functions and requirements.
4. Conta nt Spray -Hanual Refer to LCO 3.3.2, <<ESFAS Instrunentation,<< Function 2.a, for all initiation functions and requirements.

~~nb ~'m~

Hotes:

(a) Per Radiological Effluent Controls Program.

R.E. Ginna Nuclear Power Plant 3.3-40 Amendment No. 61

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 HODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least HODE 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 AFW Hanual Initiation channel is inoperable. If a manual initiation switch is inoperable, the associated AFW or SAFW pump must be declared inoperable and the applicable Conditions of LCO 3.7.5, "Auxiliary Feedwater (AFW) System" must be entered immediately. Each AFW manual initiation switch controls one AFW 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 The SRs for each ESFAS Function are identified by the SRs RE(UIREHENTS column of Table 3.3.2-1. Each channel of process protection supplies both trains of the ESFAS. When testing Channel I, Train A and Train B 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.

+~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 0

Attachment III Proposed Technical Specifications Included Pages:

TOC i TOC ii TOC iii 3.2-13 3.3-7 3.3-18 3.3-19 3.3-24 3.3-32 3.3-33 3.3-40 B3.3-100

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1-1 1.1 Definitions 1.1-1 1.2 Logical Connectors 1.2-1 1.3 Completion Times . 1.3-1 1.4 . Frequency 1.4-1 2.0 SAFETY LIMITS (SLs) 2.0-1 2.1 S Ls ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2.0-1 2.2 SL Vi ol ati ons 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . 3.0-.1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.0-4 3;1 REACTIVITY CONTROL SYSTEMS . 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) 3.1-1 3.1.2 Core Reactivity 3.1-2 3.1.3 Moderator Temperature Coefficient (MTC) 3.1-.4 3.1.4 Rod Group Alignment Limits . 3.1-7 3.1.5 Shutdown Bank Insertion Limit 3.1-11 3.1.6 Control Bank Insertion Limits 3.1-13 3.1.7 Rod Position Indication 3.1-15 3.1.8 PHYSICS TESTS Exceptions -MODE 2 . 3.1-18 3.2 POWER DISTRIBUTION LIMITS 3.2-1 3.2.1 Heat Flux Hot Channel Factor (Fa(Z)) 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F"~~) 3.2-4 3.2.3 AXIAL FLUX DIFFERENCE (AFD) 3.2-6 3.2.4 QUADRANT POWER TILT RATIO (QPTR) 3.2-11 3.3 INSTRUMENTATION 3.3-1 3.3.1 Reactor Trip System (RTS) Instrumentation ~ ~ 3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS )

Instrumentation 3.3-20 3.3.3 Post Accident Monitoring (PAM) Instrumentation . 3.3-28 3.3.4 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3-34

[

3.3.5 Containment Ventilation Isolation Instrumentation 3.3-36 3.3.6 Control Room Emergency Air Treatment System (CREATS) Instrumentation Actuation 3.3-41 (continued)

R.E. Ginna Nuclear Power Plant Amendment No. Q

0 )

TABLE OF, CONTENTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure'rom Nucleate. Boiling (DNB) Limits 3.4-1 3.4.2 RCS Minimum Temperature for Criticality 3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4-4 3.4.4 'CS Loops -MODE 1 > 8.5% RTP . . . . . . . . . . . 3.'4-6 3.4.5 RCS Loops -MODES 1 s 8.5% RTP, 2, and 3 3.4-7 3.4.6 RCS Loops -MODE 4 3.4-10 3.4.7 RCS Loops -MODE 5, Loops Filled 3.4-13 3.4.8 RCS*Loops -MODE 5, Loops Not Filled 3.4-16 3.4.9 Pressurizer 3.4-18 3.4.10 Pressurizer Safety Valves 3.4-20 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4-22 3.4.12 Low Temperature Overpressure Protection (LTOP)'Syst em 3.4-26 3.4.13 RCS Operational LEAKAGE 3.4-32 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage . . . . 3.4-34 3.4.15 RCS Leakage Detection Instrumentation 3.4-38 3.4.16 RCS Specific Activity 3.4-42 l

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5-1 3.5.1 Accumulators . . . ; . . . . . . . . . . . . . . .- 3.5-1 3.5.2 ECCS -MODES 1, 2, and 3 3.5-3 3.5.3 ECCS -MODE 4 3.5-6 3.5.4 Refueling Water Storage Tank (RWST) 3.5-8 3.6 CONTAINMENT SYSTEMS 3.6-1 3.6.1 Containment 3.6-1 3.6.2 Containment Air Locks 3.6-3 3.6.3 Containment Isolation Boundaries . . . . . . 3.6-8 3.6.4 Containment Pressure 3.6-16 3.6.5 Containment Air Temperature 3.6 17 3.6.6 Containment Spray (CS), Containment Recircul ati on

Fan Cooler (CRFC), NaOH, and Containment Pos t-Accident Charcoal Systems 3.6-18 3.6.7 Hydrogen Recombiners 3.6-24 3.7 PLANT SYSTEMS 3.7-1 3.7.1 Hain Steam Safety Valves (MSSVs) 3.7-1 3.7.2 Hain Steam Isolation Valves (MSIVs) and Non-Return Check Valves 3~7 3 3.7.3 Main Feedwater Regulating Valves (HFRVs),

Associated Bypass Valves, and Feedwater Pump Discharge Valves (HFPDVs) 3.7-5 3.7.4 Atmospheric Relief Valves (ARVs) . . . . . 3.7-8 3.7.5 Feedwater (AFW) System . .'uxiliary 3.7-10 3.7.6 Condensate Storage Tanks (CSTs) 3.7-14 3.7.7 Component Cooling Water (CCW) System . 3.7-15

] 3.7.8 Service Water (SW) System 3.7-18 (continued)

R.E. Ginna Nuclear Power Plant Amendment No. g

J tt

c TABLE OF CONTENTS 3.7 PLANT SYSTEMS (continued) 3.7.9 Control Room Emergency Air, Treatment System (CREATS) ~ ~ 3.7-20 3.7.10 Auxiliary Building Ventilation System (ABVS) ~ ~ 3.7-24 3.7.11 Spent Fuel Pool (SFP) Water Level ~ ~ 3.7-26 3.7.12 Spent Fuel Pool (SFP) Boron Concentration ~ ~ 3.7-27 3.7.13 Spent'Fuel Pool (SFP) ~ ~ 3.7-29 3.7.14 Specific Activity Storage'econdary

~ ~ 3 ~ 7 32 3.8 ELECTRICAL POWER SYSTEMS . ~ ~ 3.8-1 3.8.1 AC Sources MODES 1, 2, 3, and 4 . ~ ~ 3.8-1 3.8.2 AC Sources MODES 5 and 6 ~ ~ 3.8-8 3.8.3 Diesel Fuel Oil ~ ~ 3.8-11 3.8.4 DC Sources MODES 1, 2, 3, and 4 . ~ ~ 3.8-13 3.8.5 DC Sources MODES 5 and 6 ~ ~ 3.8-16 3.8.6 Battery Cell Parameters ~ ~ ~ ~ ~ ~ 3.8-18 3.8.7 AC Instrument Bus Sources -MODES 1, 2, 3 , and 4 ~ ~ 3.8-.20 3.8.8 AC Instrument Bus Sources -MODES 5 and 6 ~ ~ ~ ~ ~ 3.8-22 3.8.9 Distribution Systems MODES 1, 2, 3, and ~ ~ 3.8-24 3.8.10 Distribution Systems -MODES 5 and 6 ~ ~ 3.8-26 3.9 REFUELING OPERATIONS . 3.9-1 3.9.1 Boron Concentration 3.9-1 3.9.2 Nuclear Instrumentation 3.9-2 3.9.3 Containment Penetrations . ~ ~ 3.9-4 3.9.4 Residual Heat Removal (RHR) and Cool ant Circulation -Water Level z 23 Ft ~ ~ 3.9-6 3.9.5 Residual Heat Removal (RHR) and Cool ant Circulation -Water Level < 23 Ft ~ ~ 3.9-8 3.9.6 Refueling Cavity Water Level ~ ~ 3.9-10 4.0 DESIGN FEATURES 4.0-1 4.1 Site Location 4.0-1 4.2 Reactor Core . 4.0-1 4.3 Fuel Storage . 4.0-2 5.0 ADMINISTRATIVE CONTROLS 5.0-1 5.1 Responsibility . 5.0-1 5.2 Organization . ~ ~ ~ 5.0-2 5.3 Plant Staff gualificati ons . 5.0-4 5.4 Procedures . . . . . . 5.0-5 5.5 Programs and Manuals . 5.0-6 5.6 Reporting Requirements ~ ~ 5.0-18 5.7 High Radiation Area ~ ~ 5.0-23 R.E. Ginna Nuclear Power Plant Amendment No. g

QPTR'3.2.4 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.6 --------NOTES--------

1. Only required to be performed if the cause of the QPTR alarm is not associated with inoperable QPTR instrumentation.
2. Required Action A.6 must be completed when Required Action A.5 is completed and Note 1, above, does not apply.
3. Only one of the Completion Times, whichever becomes applicable first, must be met.

Perform SR 3.2.1.1 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and SR 3.2.2.1. after reaching RTP OR Wi thin 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasing THERMAL POWER above the limits of Required Actions A. 1 and A.2 (continued)

R.E. Ginna Nuclear Power Plant 3.2-13 Amendment No. g

0 a dt '

ll I l

t RTS Instrumentation 3.3.1 ACTIONS continued CONDITION REQUIRED ACTION COMPLET'ION TIME Q. Required Action and Q. 1 Reduce THERMAL POWER 6 hours Associated Completion to < 50% RTP.

Time of Condition P not met. AND Q.2.1 Verify Steam Dump 7 hours System is OPERABLE.

OR Q.2.2 Reduce THERMAL POWER 7 hours to < 8% RTP.

R. As required by R.l --------NOTE---------

Required Action A. 1 One train may be and referenced by bypassed for up to 4 Table 3.3.1-1.- hours for surveillance testing provided the other train is OPERABLE.

6 hours Restore train to OPERABLE status.

S. As required by S.1 Verify interlock is 1 hour Required Action A. 1 in required state for and referenced by existing plant Table 3.3.1-1. conditions.

OR S.2 Declare associated 1 hour RTS Function channel(s) inoperable.

(continued)

R.E. Ginna Nuclear Power Plant 3 ~3 7 Amendment No. Pg

t RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 6)

Reactor Trip System Instrumentation Note 1: Overtem erature aT The Overtemperature nT Function Trip Setpoint is defined by:

+r,s - f(h I)

Overtempereture h T 8h T0 K+K2 4

(P-P') -K (T -7') 1

+r2 s 1

Where:

aT is measured RCS aT, 'F.

aTO is the indicated aT at RTP, 'F.

s is the Laplace transform operator, sec' T is the measured RCS average temperature, 'F.

T is the nominal T,, at RTP, 'F.

P is the measured pressurizer pressure, psig.

P is the nominal RCS operating pressure, psig.

K, is the Overtemperature aT reactor trip setpoint, 1.20.

K, is the Overtemperature aT reactor trip depressurization setpoint penalty coefficient, 0.000900.

K, is the Overtemperature aT reactor trip heatup setpoint penalty coefficient, 0.0209.

r, is the measured lead/lag time constant, 25 seconds.

~, is the measured lead/lag time constant, 5 seconds.

f(aI) is a function of the indicated difference between the top and bottom detectors of the Power Range Neutron Flux channels where q, and q, are the percent power in the top and bottom halves of the core, respectively, and q, + q, is the total THERMAL POWER in percent RTP.

f(aI) = 0 when q, - q, is s +13% RTP f(aI) = 0.013 { (q, - q,) - 13I when q, - q, is > +13% RTP R.E. Ginna Nuclear Power Plant 3.3-18 Amendment No. g

t RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 6)

Reactor Trip System Instrumentation Note 2: Over ower aT The Overpower sT. Function Trip Setpoint is defined by:

rsST OverpoMsr h T MC) T~ K~-K,(T-T') -K, 0 -f(C) I) f,s+1 Where:

aT is measured RCS sT, 'F.

aTO is the indicated aT at RTP, 'F.

s is the Laplace transform operator, sec' T is the measured RCS average temperature, 'F.

T is the nominal T., at RTP, F.

K4 is the Overpower aT reactor trip setpoint, 1.077.

K, is the Overpower aT reactor trip heatup setpoint penalty coefficient which is:

0.0 for T ( T and; 0.0011 for T z T2.

K, is the Overpower'T reactor trip thermal time delay setpoint penalty which is:

0.0262 for increasing T and; 0.00 for decreasing T.

1 3 i s the measured l ead/1 ag time constant, 10 seconds .

f(sI) is a function of the indicated difference between the top and bottom detectors of the Power Range Neutron Flux channels where q, and qb are the percent power in the top and bottom halves of the core, respectively, and q, + qb is the total THERMAL POWER in percent RTP.

f(61) = 0 when q, - q, is s +13% RTP f(aI) 0.013 ( (q, - q,) - 13I when q, - q, is ) +13% RTP R.E. Ginna Nuclear Power Plant 3 c3-19 Amendment No. g

0 i ESFAS Instrumentation 3.3.2 SURVEILLANCE RE(UIREHENTS NOTE Refer to Table 3;3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FRE(UENCY SR 3.3.2. 1 Perform CHANNEL CHECK. 12 hours SR 3.3.2.2 Perform COT. 92 days SR 3.3.2.3 NOTE-Verification of relay setpoints not required.

Perform TADOT.. 92 days SR 3.3.2.4 -NOTE Verification of relay setpoints not required.

Perform TADOT. 24 months SR 3.3.2.5 Perform CHANNEL CALIBRATION. 24 months SR 3.3.2.6 Verify the Pressurizer Pressure Low and Steam Line Pressure Low Functions are not

'4 months bypassed when pressurizer pressure > 2000 Pslg.

SR 3.3.2.7 Perform ACTUATION LOGIC TEST. 24 months R.E. Ginna Nuclear Power Plant 3.3-24 Amendment No. Q

Table 3.3.3-1 (page 1 of 2)

Post Accident Honitoring Instrumentation REQUIRED FUNCTION CHANNELS CONDITION

1. Pressurizer Pressure
2. Pressurizer Level
3. Reactor Coolant System (RCS) Hot Leg 1 per loop Temperature RCS Cold Leg Temperature 1 per loop
5. RCS Pressure (Wide Range) 2
6. RCS Subcooling Monitor 2
7. Reactor Vessel Water Level 2
8. Containment Sump B Mater Level 2 9 Containment Pressure (Wide Range) 2
10. Containment Area Radiation (High Range) 2 Hydrogen Monitors 2
12. Condensate. Storage Tank Level 2
13. Refueling Water Storage Tank Level 2
14. Residual Heat Removal Flow 2
15. Core Exit Temperature Quadrant 1 2(a)
16. Core Exit Temperature Quadrant 2 2(a)
17. Core Exit Temperature Quadrant 3 2(a)
18. Core Exit Temperature Quadrant 4 2(a)
19. Auxiliary Feedwater (AFW) Flow to Steam 2 Generator (SG) A
20. AFW Flow to SG B
21. SG A Water Level (Narrow Range)
22. SG B Water Level (Narrow Range)

(continued (a) A channel consists of two core exit thermocouples (CETs).

R.E. Ginna Nuclear Power Plant 3.3-32 Amendment No. g

N I%+ ec t PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)

Post Accident Monitoring Instrumentation RE(VIREO FUNCTION, CHANNELS CONDITION

23. SG A Water Level (Wide Range)
24. SG B Water Level (Wide Range)

(

25. SG A Pressure
26. SG B Pressure R.E. Ginna Nuclear Power Plant 3 ~ 3 33 Amendment No. Q

tk C

11'

C ontainment Ventilation isolation Instrumentation 3.3.5 Table 3.3.5-1 (page 1 of 1)

Contairaent Ventilation Isolation Instrwentation SURVEILLANCE FUHCTIOH REQUIRED CNAHHELS TRIP SETPOINT REQUIRENEHTS

1. Automatic Actuation Logic and 2 trains SR 3.3.5.3 HA Actuation Relays
2. Containment Radiation
a. Gaseous SR 3.3 '.1 (a)

SR 3.3.5 '

SR 3.3.5.4

b. Particulate SR 3.3.5.1 (a)

SR 3.3.5.2 SR 3.3.5.4

3. Containment Isolation Refer to LCO 3.3.2, "ESFAS Instrwentation," Function 3, for all initiation functions and requirements.
4. Contaiment Spray -Nanual, Refer to LCO 3.3.2, "ESFAS Instrwentation,a Function 2.a, for all I Initiation initiation functions and requirements.

Notes:

(a) Per Radiological Effluent Controls Program.

R.E. Ginna Nuclear Power Plant 3.3-40 Amendment No. 'g

ESFAS Instrumentation B 3.3.2 BASES ACTIONS H.l (continued)

If the Required Actions and Completion Times of Condition L are not met, the plant must be brought to a HODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least HODE 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..1 Condition N applies if a AFW Hanual Initiation channel is inoperable. If a manual initiation switch is inoperable, the associated AFW or SAFW pump must be declared inoperable and the applicable Conditions of LCO 3.7.5, "Auxiliary Feedwater (AFW) System" must be entered immediately. Each AFW manual initiation switch controls one AFW 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 The SRs for each ESFAS Function are identified by the SRs REQUIREHENTS 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 B 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 B 3.3-100 Revision 0

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