ML17264A524
| ML17264A524 | |
| Person / Time | |
|---|---|
| Site: | Ginna  (DPR-18-A-065, DPR-18-A-65) |
| Issue date: | 06/03/1996 |
| From: | Jeffrey Mitchell NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17264A525 | List: |
| References | |
| NUDOCS 9606060055 | |
| Download: ML17264A524 (17) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555<001 ROC S
ER GAS AND ELECTRIC CORPORATION DOC NO. 50-4 R
GI NUC RPO PL ME DttIENT TO FACILITY OPERATING LIC NS Amendment No.
65 License No.
DPR-18 The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Rochester Gas and Electric Corporation (the licensee) dated Hay 8,
- 1996, as supplemented Hay 10,
- 1996, Hay 29,
- 1996, and June 3,
- 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 2.
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements.
have been satisfied.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this l.icense amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-18 is hereby amended to read as follows:
9606060055 960603 PDR ADQCK 05000244 P
(~)
ec S
ci ications N
The Technical Specifications contained in Appendix A, as revised through Amendment No.
65
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 3,
1996 Jocelyn A. Mitchell, Acting Director Project Directorate I-l Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
CH T TO CENSE AM NDM NT NO. 65 C
TY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove 1lliii 3.2-13 3 ~ 3 7
3.3-18 3.3-19 3.3-24 303-32 3 ~ 3 33 3.3-40 B 3.3-100 Insert iiiiii
- 3. 2-13 3 ~ 3 7
3.3-18 3.3-19 3.3-24 3.3-32 3 ~ 3 33 3.3-40 B 3.3-100
TABLE OF CONTENTS 1.0 1.1 1.2 1.3 1.4 2.0 2.1 2.2 3.0 3.0 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 USE AND APPLICATION Definitions Logical Connectors Completion Times Frequency SAFETY LIMITS (SLs)
Ls
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S SL Violations I
LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.
SURVEILLANCE RENDU'..".EHENT (SR) APPLICABILITY REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDH)
Core 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 1.1-1 1.1-1 1.2-1 1.3-1 1.4-1 2.0-1 2.0-1 2.0-1 3 n 3.0-4 3.1-1 3.1-1 3.1-2 3.1-4 3.1-7 3.1-11 F 1-13 3.1-15 3.1-18 32.
3.2.1 3.2.2 3.2.3 3.2.4 POWER DISTRIBUTION LIMITS Heat Flux Hot Channel Factor (Fo(Z))
Nuclear Enthalpy Rise Hot Channel Factor AXIAL FLUX DIFFERENCE (AFD)
QUADRANT POWER TILT RATIO (gPTR)
(F"m) 3.2-1 3.2-1 3.'2-4 3.2-6 3.2-11 3.3 3.3.1 3.3.2 3.3 '
3.3.4
[ 3.3.5 3.3.6 INSTRUMENTATION Reactor Trip System (RTS) Instrumentation Engineered Safety Feature Actuation System (ESFAS Instrumentation Post Accident Honitoring (PAH) Instrumentation Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation Containment Ventilation Isolation Instrumentation Control Room Emergency Air Treatment System (CREATS) Instrumentation Actuation
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3.3-1 3.3-1 3.3-20 3.3-28 3.3-34 3.3-36 3.3 tl (continued)
R.E.
Ginna Nuclear Power Plant Amendment No.
l3S.
66
TABLE OF CONTENTS 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.4.11 3.4.12 3.4.13 3.4.14 3.4.15 3.4.16 REACTOR COOLANT SYSTEH (RCS)
RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits RCS Hinimum Temperature for Criticality RCS Pressure and Temperature (P/T) Limits RCS Loops -MODE 1
> 8.5%
- RTP, 2,
and 3
RCS Loops -HODE 4 RCS Loops MODE 5, Loops Filled RCS Loops -MODE 5, Loops Not Filled Pres=urizer Pressurizer Safety Valves Pressurizer Power Operated Relief Valves (PORVs)
Low Temperature Overpressure Protection (LTOP) System RCS Operational LEAKAGE RCS Pressure Isolation Valve (PIV) Leakage RCS Leakage Detection Instrumentation RCS Specific Activity 3.4-1 3.4-1 3.4-3 3.4-4 3.4-6 3.4-7 3.4-10 3.4-13 3.4-16 3.4-18 3.4-20 3.4-22 3.4-26 3.4-32 3.4-34 3.4-38 3.4-42 3.5 3.5.1 3.5.2 3.5.3 3.5.4 EMERGENCY CORE COOLING SYSTEMS (ECCS)
Accumulators ECCS MODES 1, 2, and 3
ECCS -MODE 4 Refueling Water Storage Tank (RWST)
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3.5-1 3.5-1 3.5-3 3.5-6 3.5-8 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 3.6.7 3.7 3.F 1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7
( 3.7.8 CONTAINMENT SYSTEMS Containment Containment Air Locks Containment Isolation Boundaries Containment Pressure Containment Air Temperature Containment Spray (CS),
Containment Recircul
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Fan Cooler (CRFC),
- NaOH, and Containment Charcoal Systems Hydrogen Recombiners PLANT SYSTEHS Hain Steam Safety Valves (HSSVs)
Hain Steam Isolation Valves (MSIVs) and Non-Return Check Valves Hain Feedwater Regulating Valves (HFRVs),
Associated Bypass Valves, and Feedwater Pump Discharge Valves (HFPDVs)
Atmospheric Relief Valves (ARVs)
Auxiliary Feedwater (AFW) System Condensate Storage Tanks (CSTs)
Component Cooling Water (CCW) System Service Water (SW) System ati Pos 3.6-1 3.6-1 3.6-3 3.6-8 3.6-16 3.6 17 on t-Accident 3.6-18 3.6-24 3.7-1 3.7-1 3.7-3 3.7-5 3.7-8 3.7-10 3.7-14 3.7-15
- 3. 7-18 (continued)
R.E.
Ginna Nuclear Power Plant Amendment No. Pg 65
TABLE OF CONTENTS 3.7 3.7.9 3.7.10 3.7.11 3.7.12 3.7.13 3.7.14 PLANT SYSTEMS (continued)
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 3.7-20 3.7-24 3.7-26 3.7-27 3.7-29 3.7-32, 3.8 3.8.1 3.8.2 3.8.3 3.8.4 3.8.5 3.8.6 3.8.7 3.8.8 3.8.9 3.8.10 3.9 3.9.1 3.9.2 3.9.3 3.9.4 3.9.5 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 -HODES 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 Water Level ~ 23 Ft Residual Heat Removal (RHR) and Coolant Circulation Water Level
< 23 Ft Refueling Cavity Water Level
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and 4
3.8-1 3.8-1 3.8-8 3.8-11 3.8-13 3.8-16 3.8-18 3.8-20 3.8-22 3.8-24 3.8-26 3.9-1 3.9-1 3.9-2 3.9-4 3.9-6 3.9-8 3.9-10 4.0 4.1 4.2 4.3 5.0 5.1 5.2 5.3 5.4 5.5 5.6 5.7 DESIGN FEATURES Site Location Reactor Core Fuel Storage ADMINISTRATIVE CONTROLS Responsibility Organization Pl ant Staff equal ificat Procedures Programs and Manuals Reporting Requirements
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4.0-1 4.0-1, 4.0-1 4.0-2 5.0-1 5.0-1 5.0-2 5.0-4 5.0-5 5.0-6 5.0-18 5.0-23 R.E.
Ginna Nuclear Power Plant Amendment No. Pf, 65
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 and SR 3.2.2.1.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 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. Pg, 65
RTS Instrumentation 3.3.1
'IQ+
1 ACTIONS continued CONDITION REQUIRED ACTION COMPLETION TIME Q.
Required Action and Associated Completion Time of Condition P
not met.
Q. 1 Reduce THERMAL POWER to
< 50% RTP.
AND 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Q.2.1 Q.2.2 Verify Steam Dump System is OPERABLE.
OR Reduce THERMAL POWER to < 8% RTP.
7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> 7 hours R.
As required by Required Action A. 1 and referenced by Table 3.3.1-1..
R.1
NOTE---------
One train may be bypassed for 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 provided the other train is OPERABLE.
Restore train to OPERABLE status.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> S.
As required by Required Action A. 1 and referenced by Table 3.3.1-1.
S.1 OR S.2 Verify interlock is in required state for existing plant conditions.
Declare associated RTS Function channel(s) inoperable.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 hour (continued)
R.E.
Ginna Nuclear Power Plant 3 03 7
Amendment No. Pf, 65
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 aT Function Trip Setpoint is defined by:
1 +r,s Overterpereture'd TddTe K,+K (P-P') -K (T -T )
'f(dl) 1 +r~s Where:
aT is measured RCS dT, 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 hT reactor trip depressurization setpoint penalty coefficient, 0.000900/psi.
K, is the Overtemperature aT reactor trip heatup setpoint penalty coefficient, 0.0209/
F.
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r, is the measured lead/lag time constant, 25 seconds.
r2 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'he top and bottom halves of the core, respectively, and q, + q, is the total THERMAL POWER in percent.RTP.
when q, -
qb is s +13%
RTP when q, -
qb is
> +13%
RTP R.E.
Ginna Nuclear Power Plant 3.3-18 Amendment No. gg,
0
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 aT Function Trip Setpoint is defined by:
resT Overpower h T Mh To K,-K~ (T-T') -K, o
-f(4 r)
Where:
'I sT 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 K~ is the Overpower aT reactor 0.0/
F for T < T and; 0.0011/'F for T ~ T.
K, is the Overpower aT reactor which is:
0.0262/'F for increasing T
0.00/'F for decreasing T.
trip setpoint, 1.077.
trip heatup setpoint penalty coefficient which is:
trip thermal time delay setpoint. penal.ty and; r~ is the measured lead/lag time constant, 10 seconds.
f(nI) 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(aI) = 0 f(a))
= 1.3
( (q, - q,) -
13) when q, -
qb is ~ +13% RTP when q, - q is
> +13%
RTP R.E.
Ginna Nuclear Power Plant 3.3-19 Amendment No. Pg, t~i
ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS
-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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 1
Verify the Pressurizer Pressure Low and Steam Line Pressure Low Functions are not bypassed when pressurizer pressure
> 2000 Pslg
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24 months SR 3.3.2.7 Perform ACTUATION LOGIC TEST.
24 months R.E.
Ginna Nuclear Power Plant 3.3-24 Amendment No.
Pg, 65
~
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 2)
Post Accident Monitoring Instrumentation FUNCTION RE(UIRED CHANNELS CONDITION 15.
16.
17.
18.
19.
20.
21.
I 22.
Core Exit Temperature quadrant 1
Core Exit Temperature quadrant 2
Core Exit Temperature quadrant 3
Core Exit Temperature quadrant 4
Auxiliary Feedwater (AFW) Flow to Steam Generator (SG)
SG A Water Level (Narrow Range)
SG B Water Level (Narrow Range) 1.
Pressurizer Pressure 2.
Pressurizer Level 3.
Hot Leg Temperature 4.
RCS Cold Leg Temperature 5.
RCS Pressure (Wide Range) 6.
RCS Subcooling Monitor 7.
Containment Sump B Water Level 9.
Containment Pressure (Wide Range) 10.
Containment Area Radiation (High Range) 11.
Hydrogen Monitors 12.
Condensate Storage Tank Level 13.
Refueling Water Storage Tank Level 14.
Residual Heat Removal Flow 1 per loop 1 per loop 2
2 2
2 2
2 2
2 2
2 2(a) 2(a) 2(a) 2(a) 2 H
G (continued)
(a)
A channel consists of two core exit thermocouples (CETs).
R.E. Ginna Nuclear Power Plant 3.3-32 Amendment No. $/65
PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 2 of 2)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION I
23.
24.
(
25.
26.
SG A Water Level (Wide Range)
SG B Water Level (Wide Range)
G G
G G
, R.E.
Ginna Nuclear Power Plant 3 ~ 3 33 Amendment No.
Pg, 65
Containment Ventilation Isolation Instrumentation 3.3.5 Table 3.3.5-1 (page 1 of 1)
Contain>>nt Ventilation Isolation Instrunentation FUMCTIOH REOUIRED CHAMNELS SURVEILLANCE REOUIREHEMTS TRIP SETPOIHT 1 ~
Automatic Actuation Logic and Actuation Relays 2 trains SR 3.3.5.3 NA 2.
Contalraent Radiation a.
Gaseous SR 3.3.5.1 SR 3.3.5.2 SR 3.3.5.4 (a) b.
Particulate SR 3.3.5.1 SR 3.3.5.2 SR 3.3.5.4 (a) 3.
Contaireent Isolation Refer to LCO 3.3.2,
<<ESFAS Instrunentation,<<
Function 3, for all initiation fax:tions and requirements.
4.
Contaireent Spray -Manual Initiation Refer to LCO 3.3.2,
<<ESFAS Instrunentation,<<
Function 2.a, for all initiation fmctions and requirements.
Notes:
(a)
Per Radiological Effluent Controls Program.
R.E.
Ginna Nuclear Power Plant 3.3-40 Amendment No. Pg, 65
ESFAS Instrumentation 8 3.3.2 BASES ACTIONS (continued)
H. I If the'equired 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. I Condition N applies if a AFW Manual 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 REgUIREHENTS 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 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.
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 0
Amendment No.
65