ML17053A428
| ML17053A428 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/29/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17053A430 | List: |
| References | |
| NUDOCS 7902260152 | |
| Download: ML17053A428 (38) | |
Text
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+p*yW UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-"220 NINE MILE POINT NUCLEAR STATION UNIT NO.
1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
26 License No.
DPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:.
A.
The applications for amendment by Niagara Mohawk Power Corporation (the licensee) dated May 19 and May 25, 1977, comply with the standards and requirements o'$ 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 applications, 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; 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.
790 2260 15'
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t 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No.
DPR-63 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 26, 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 the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 29, 1979.
Thomas
. Ippolito, Chief Operating Reactors Branch ¹3 Division of Operating Reactors
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ATTACHMENT TO LICENSE AMENDMENT NO.
26 FACILITY OPERATING LICENSE NO.
DPR-63 DOCKET NO. 50-220 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
~Pa es iil 129 130 131 132 133
~188 190 236 237 Add pages:
164a 164b 164c 164d 232b 232c
SECTION 3.3.4 Isolation Valves 3.3.5 Access Control 3.3.6 Vacuum Relief 3.3.7 Containment Spray DESCRIPTION 4.3.4 Isolation Valves 4.3.5 Access Control 4.3.6 Vacuum 'Relief 4.3.7 Containment Spray 14.4'50 152 158
~
3.3.8 Drywell - Suppression Chamber Differenti'al Pressure 3.4.0 Secondary Containment Limitin Condition for 0 eration 4.3.8 Drywell Suppression Chamber Differential Pressure Surveillance Re uirements 164a 165 3.4.1 Leakage Rate 3.4.2 Isolation Valves 3.4. 3 3.4.4 3.4.5.
Access Control 4
Emergency Ventilation Control Roottt Ventilation 4.4.1 Leakage Rate 4.4.2 Isolation Valves 4.4.3 Access Control 4.4.4 Emergency Ventilation 4.4.5 Control
.",oom Wentilation 166 1.69 171 173 17v 3.5.0 Shutdown and Refueling Limitin Conditi n for eration 3.5.1 Source Range Monitoring 3.5.2 Refueling Platform Interlock 3.6.0 General Reactor Plant Surveillance Re uirements 4.5.1 Source Range Honitoring 4.5.2 Refueling Platform Interlock 179 1aO 183 185 Li.miting Condition for Operation 3.6.1 Station Process Effluents 3.6. 2 Protective Instrumentation 3.6.3 Emergency Power Sources Surveillance Requirements 4.6.1 Station Process Effluents 4.6.2
'Protecti ve Instrumentation 4.6.3 Emergency Power Sources iii 188 238
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE UIREMENT 3.3.'2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL
~tll ttll Applies to the interrelated parameters of pressure suppression system pressure and suppression chamber water temperature and level.
~0b 'ective:
To assure that the peak suppression chamber pressure does not exceed design values in the event of a loss-of-coolant accident.
~it'll The downcomers in the suppression chamber shall have a minimum submergence of three feet and a maximum submergence of four and one half feet whenever the reactor coolant system temperature is above 215F.
.b.
During normal power operation, the combination of primary containment pressure and suppression chamber water temperature shall be within the
- shaded area of.,
4.3.2 PRESSURE SUPPRESSION SYSTEM PRESSURE AND SUPPRESSION CHAMBER WATER TEMPERATURE AND LEVEL A~ti Atilt Applies to the periodic testing of the pressure suppression system pressure and suppression chamber water temperature and level.
~0b 'ecti ve:
To assure that the pressure suppression system pressure and suppression chamber water temperature and level are within required limits..
t~iti tl a.
At least once per day the suppression chamber water level and temperature and pressure suppression system pressure shall be checked.
b.
A visual inspection of the suppression chamber interior, including water line regions, shall be made at each major refueling outage.
c.
Whenever heat from relief valve operation is being added to the suppression pool the pool temperature shall be continually monitored and also observed and logged every 5 minutes until the heat addition is terminated.
129
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUiRENENT (1) Figure 3.3.2a when downcomer submergence is
> 4 feet, or (2) Figure 3.3.2b when downcomer submergence is
> 3 feet. If these temperatures are exceeded, pool cooling shall be initiated immediately.
c.
If Specifications a and b above are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be shutdown using normal shutdown procedures.
d.
During testing of relief valves which add heat to the torus pool, the water temperature shall not exceed 10F above the 'normal power operation limit specific in b above.
In connection with such testing, the pool temperature must be reduced within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to below the normal power operation limit specified in b above.
e.
The reactor shall be scrammed from any operating condition when the suppression pool temperature reaches llOF.
Operation shall not be resumed until the pool temperature is reduced to below the normal power operation limit specified in b above.
d.
Whenever operation of a relief valve is indicated and the suppression pool temperature reaches 160F or above while the reactor primary coolant system pressure is greater than 200 psig, an external visual examination of the suppression chamber shall be made before resuming normal power operation.
.f.
During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal cooldown rates if the pool temperature reaches 120F.
130
LEFT INTENTIONALLY BLANK
5.0 Figure 3.3.2 a
ALLOWABLE PRESSURE SUPPRESSION SYSTEM 4
FOOT DOWNCOMER SUBMERGENCE CD I
M CL CD t/l UJ CL CL C/l 3.0 2.0 68 72 76 80 84 88 92 SUPPRESSION CHAMBER WATER OPERATIONAL TEMPERATURE F
96 100 132
C9 C/l CL I
CCj CX C/l UJ CC O
5.0 Figure 3,3.2 b
ALLOWABLE PRESSURE SUPPRESSION SYSTEM WN F
S R
R ENCE CD CL, CD Ccl I
C/l C/l CD C/l C/l Cu CC Ct Ct C/l 4.0 3.0 2.0 68 72 76'0 84 88 92 SUPPRESSION CHAMBER MATER OPERATIONAL TEMPERATURE F
96 100 133
0
LIMITING CONDITIONS FOR OPERATION SURYEILLANCE RE VIREMENTS 3.3.8 Drywell-Suppression Chamber Differential Pressure 4.3.8 Drywell-Suppression Chamber Differential Pressure
~ll li billet Applies to the operational status of drywell-suppression chamber differential pressure system.
~0b ective:
To assure that the pressure suppression system will remain functional during a design basis loss-of-coolant accident.
~A1i bill Applies to the periodic testing requirements for the drywell-suppression chamber differential pressure system.
~0b'ective:
To verify the operability of the drywell-suppression chamber differential system.
~Si B Differential pressure between the drywell and suppression chamber shall be maintained at or above levels according to Figure 3.3.8 except as specified in (1) and (2) below:
{1)
The differential pressure shall be
'stablished within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of achieving operating pressure and temperature.
The differential pressure may be reduced to less than that specified by Figure 3.3.8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.
(2)
This differential may be decreased to less than that of Figure 3.3.8 for a maximum of four (4} hours during required operability testing of the drywell-pressure suppression chamber vacuum breakers.
~5i fi a.
The pressure differential between the drywell and suppression chamber shall be recorded at least once each shift.
164a
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS b.
If the differential pressure of Specification 3.3.8.a cannot be maintained and the differential pressure cannot be restored within the sub-sequent six (6) hour period, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition within the following 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />.
l64b
FIGURE 3.3.8 Drywell-Suppression Chamber Differential Pressure vs'owncomer Submergence 2.0 CY UJ CQ
<<t trJ
~ QC C/l K Cll CD 4J C/l A UJU O
O0 I CL N 4J I
LC M La UJ O QC CD 1.75 1.5-1.25 1.0 NORMAL OPERATING RANGE I
3.0 3.5
~4.0 4.5 5
DOWNCOMER SUBMERGENCE (FEET) 164c
BASES FOR 3.3.8 AND 4.3.8 DRYWELL-SUPPRESSION CHAMBER DIFFERENTIAL PRESSURE In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system and attached piping.
The maintenance of a drywell-suppression chamber differential pressure in accordance with Figure 3.3.8 and suppression chamber water level corresponding to a downcomer submergence range of 3.0 to 4.5 feet will assure the integrity of: the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
164d
LIMITING CONDITION FOR OPERATION SURVEILLANCE RE(UIREMENT 3.6. 2 PROTECTIVE INSTRUMENTATION AJJ1i billet,'pp'lies to the operability of the plant instrumentation that. performs a safety function.
4.6. 2 PROTECTIVE INSTRUMENTATION A
1 icabi 1 it:
Applies to the surveillance of the in-strumentation that performs a safety function.
Ob ective:
To assure the operability of the instru-mentation required for safe operation.
Ob ective:
To verify the operability of protec-tive instrumentation.
S ecification:
S ecification:
a.
The set points, minimum number of trip systems, and minimum number of instrument channels that must be op-erable for each position of the re-actor mode switch shall be as given in Tables 3.6.2a to 3.6.21.
Sensors and instrument channels shall be checked, tested and cali-brated at least as frequently as listed in Tables 4.6.2a to 4.6.21.
If the requirements of a table are
'not met, the actions listed below for the respective type of instrumentation shall be taken.
(1)
Instrumentation that initiates.
scram-control rods shall be inserted.
188
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT (8)
Off-Gas and Vacuum Pump Isolation-The respective system shall be iso-lated or the instrument channel shall be considered inoperable and Specification 3.6. 1 shall be applied.
(9)
Diesel Generator Initiation - The diesel generator shall be considered inoperable and Specification 3.6.3 shall be applied.
(10) Emergency Ventilation Initiation-The emergency ventilation system shall be considered inoperable and Specification 3.4.4 shall be applied.
(ll) High Pressure Coolant Injection Initiation - The high pressure coolant injection system shall be considered inoperable and Spec-ification 3.1.8.c shall be applied.
(12) Primary Containment Monitoring-The primary containment monitoring instrumentation shall be considered inoperable and Specification 3.3.8 shall be applied.
b.
During operation with a Maximum Total Peaking Factor (MTPF) greater than the-design value, either:
(1)
The APRM scram and rod block set-tings shall be reduced to the values given by the equations in Specifica-tion 2.1.2.a; or (2)
The power distribution shall be changed such that the MTPF no longer exceeds the design value.
190
Table 3.6.21 PRIMARY CONTAINMENT MONITORING Limitin Condition For 0 eration Parameter Minimum No.
of Tripped or Operable Tri S stems Minimum No. of Operable Instrument Channels Per Operable Tri S stem Set Point Reactor Mode Switch Position in Hhich Function Must Be 0 erable s
CL Cl U
W S
CJ (ted Cfl C/l (1)
Drywel1-Suppression
'hamber Differential Pressure Figure 3.3.8 X
X (2)
Suppression Chamber Hater Level Specification 3.3.2 X
X 232b
lp-
Table 4.6.21 PRIMARY CONTAINMENT MONITORING Surveillance Re uirement Parameter Sensor Check Instrument Channel Test Instrument Channel Calibration (1)
Drywell-Suppression Chamber Differential Pressure once/day N/A (2)
Suppression Chamber Mater Level once/day N/A Once Every Six Months 232c
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The set points on the generator load re9ection and turbine stop valve. closure scram trips are set to anti-cipate and minimize the consequences of turbine trip with failure of the turbine bypass system as described in the bases for Specification 2.1.2.
Since the severity of the transients is dependent on the reactor op-erating power level, bypassing of the scrams below the specified power level is permissible.
The primary containment monitoring system is provided to alert the operator of conditions which cou'1d reduce safety margins during a postulated Loss of Coolant Accident.
Appropriate operator corrective action is described in Specification 3.3.8, should Limiting Conditions for Operation be exceeded.
This monitoring instrumentation does not automatically initiate engineered safeguards systems.
Although the operator will set the setpoints at the values indicated in Tables 3.6.2.a-l, the actual values of the various set points can differ appreciably from the value the operator is attempting to set.
The de-viations include inherent instrument error, operator setting error and drift of the set point.
These errors are compensated for in the transient analyses by conservatism in the controlling parameter assumptions as discussed in the bases for Specification 2.1.2.
The deviations associated with the set points for the safety systems used to mitigate accidents have negligible effect on the initiation of these systems.
These safety systems have initiation times which are orders of magnitude greater than the difference in time between reach-ing the nominal set point and the worst set point due to error.
The maximum allowable set point deviations are listed below:
Neutron Flux APRH, +2.7/ of rated neutron flux IRtl, +2.5X of rated neutron flux Recirculation Flow, + 1/. of rated recirculation flow Reactor Pressure,
+15.8 psig Containment
- Pressure,
+0.053 psig Reactor Water Level, +2.6 inches of water Hain Steam Line Isolation Valve Position, +2.5X of stem position Scram Discharge Volume, + 0 and -
1 gallon Condenser Low Vacuum, 40.5 inches of mercury 236
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION High Flow-Main Steam Line, +1 psid High Flow-Emergency Cooling Line, +1 psid High Area Temperature-Hain Steam Line, +10F High Area Temperature-Clean-up and Shutdown,
~6F High Radiation-Hain Steam Line, +100Ã and -50Ã of set point value High Radiation-Emergency Cooling System Vent, +100K and -504 of set point High Radiation-Reactor Building Vent-, +100K and -50Ã of set point High Radiation-Refueling Platform, +100K and -50Ã of set point High Radiation-Offgas Line, +50K of set point, (Appendix D)*
Drywell-Suppression Chamber Differential Pressure,
+O.l psid Suppression Chamber Water Level, +1.8 inches
-4 The test intervals for the trip systems result to calculated failure probabilities
<10 which corresponds to the proposed IEEE Criteria For System Failure Probability.
(IEEE SG-3, Information Docket 81 - Protec-tion System Reliability, April 24, 1968).
The test interyals for the trip systems result in calculated failure probabilities ranging from 6.7 x 10 to 1.76 x 10 1" (Fifth Supplement,
- p. 115).*
The more frequent sensor checks result in even less probability that the particular system will.fail.
Because of local high radiation, testing instrumentation in the area of the main steam 'line isolation valves can only be done during periods of Station shutdown.
These func-.
tions include high area temperature isolation, high radiation isolation and isolation valve position scram.
Testing of the scram associated with the shutdown position of the mode switch can be done only during periods of Station shutdown since it always involves a scram.
- FSAR 237