ML17056B267
| ML17056B267 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 02/01/1991 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17056B266 | List: |
| References | |
| NUDOCS 9102110210 | |
| Download: ML17056B267 (24) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 122 License No.
DPR-63 The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated June 20,
- 1988, as supplemented October 19, 1989, 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)'hat such activities will be corducted in compliance with the Commission's regulations; E.
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 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.
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:
E ppes 0~>0 91020g p
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(2)
Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 122, are hereby incorporated in the license.
The licensee shall operate
@he.~cility in accordance with the Technical Specifications.
This license amendment is effective as of the date of its issuance to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
122 Changes to the Technical Specifications
~Robert A. Capra, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Date of Issuance:
February 1, 1991
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
TO FACILITY OPERATING LICENS'E NO.
DPR-63 DOCKET NO. 50-220 Revise Appendix A as follows:
Remove Pa es ll 111 188 190 232b 232c 232d 232e 236 237 241i i Insert Pa es 11
,.111 188 190 232b 232c deleted deleted 236 237 24lii
SECTION DESCRIPTION PAGE 3.2.0 Reactor Coolant System Limitin Condition for 0 eration Surveillance Re uirements 74 3.2.1 Reactor Vessel Heatup and Cooldown Rates 75 3.2.2 Hinimum Reactor Vessel Temperature for Pressurization 4.2.2 Hinimum Reactor Vessel Temperature for Pressurization 77 3.2.3 Coolant Chemistry 3.2.4 Coolant Activity 3.2.5 Leakage Rate
! 3.2.6 Inservice Inspection and Testing 3.2.7 Isolation Valves 4.2.3 Coolant Chemistry 4.2.4 Coolant Activity 4.2.5 Leakage Rate 4.2,6 Inservice Inspection and Testing 4.2.7 Isolation Valves 83
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87 89 92 116 3.2.8 Safety Valves 3.2.9 Solenoid-Actuated Pressure Relief Valves 3.3.0 Primary Containment Limitin Condition for 0 eration 4.2.8 Safety Valves 4.2.9 Solenoid-Actuated Pressure Relief Valves Survei 1 lance Re uirements 121 123 125 3.3.1 Oxygen Concentration 3.3.2 Pressure and Suppression Chamber Water Temperature and Level 3.3.3 Leakage Rate 4.3.
1 Oxygen Concentration 4.3.2 Pressure and Suppression Chamber Water. Temperature and Level 4.3.3 Leakage Rate 126 129 135 Amendment No. l22
O.
III
SECTION 3.3.4 Isolation Valves 3.3.5 Access Control 3.3.6 Vacuum Relief 3.3.7 Containment Spray 3.4.0 Secondary Containment Limitin Condition for 0 eration 3.4.1 Leakage Rate 3.4.2 Isolation Valves 3.4,3 Access Control 3.4.4 Emergency Ventilation 3.4.5 Control Room Ventilation 3.5.0 Shutdown and Refueling Limitin Condition for 0 eration 3.5.1 Source Range Monitoring 3.5.2 Refueling Platform Interlock 3.6.0 General Reactor Plant Limitin Condition for 0 eration 3.6.1 Station Process Effluents U
3.6.2 Protective Instrumentation 3.6.3 Emergency Power Sources 3.6.4 Shock Suppressors (Snubbers)
DESCRIPTION 4.3.4 Isolation Valves 4.3.5 Access Control 4.3.6 Vacuum Relief 4.3.7 Containment Spray Limitin Condition for 0 eration 4.4.1 Leakage Rate 4.4.2 Isolation Valves 4.4.3 Access Control 4.4.4 Emergency Ventilation 4.4 '5 Control Room Ventilation Limitin Condition for 0 eration 4.5.1 Source Range Monitoring 4.5. 2 Refuel i ng Pl at form Inter 1 ock Limitin Condi tion for 0 eration 4.6.1 Station Process Effluents 4.6.2 Protective Instrumentation 4.6.3 Emergency Power Sources f
4.6.4 Shock Suppressors (Snubbers)
PAGE 144 150 152 158 165 166 169 171 173 178 179 180 0
185 186 188 238 241a Amendment 11o. p,
/7 122
LIHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREHENT 3.6.2 PROTECTIVE INSTRUHENTATION tlllt 4.6.2 PROTECTIVE INSTRUHENTATION Illit App1 i es to the operabi 1 i ty of the pl ant instrumentation that performs a safety function.
Applies to the surveillance of the instrumentation that performs a safety function.
~Ob'ective:
To assure the operability of the instrumentation required for safe operation.
t~ifl tl O~b'ective:
To verify the operability of protective instrumentation.
n i
li ttli a.
The set points, minimum number of trip
- systems, and minimum number of instrument channels that must be operable for each position of the reactor mode switch shall be as given in Tables 3.6.2a to 3.6.21.
a.
Sensors and instrument channels shall be checked, tested and calibrated 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, unless there is no fuel in the reactor vessel.
Amendment No.
122 188
I
LIHITING CONDITION FOR OPERATION SURVEILLANCE RE UIREHENT (8) Off-Gas and Vacuum Pump Isolation The respective system shall be isolated 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 Specification 3.1.8.c shall be applied.
(12) Control Room Ventilation The control room ventilation system shall be considered inoperable and Specification
-3.4.5 shall be applied.
b.
During operation with a Haximum Total Peaking Factor (HTPF) greater than the design value, either:
Amendment No.
122 190
<k I
Table 3.6.21 CONTROL ROOM AIR TREATMENT SYSTEM INITIATION Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable S
t Minimum No. of Operable Instrument Channels per Operable Tri S stem Set Point Reactor Mode Switch
- Position in Which Function Must Be 0 erable O
r D
(V QJ V),
CL (1) Nigh Radiation Ventilation Intake 1000 CPH X
X X
Amendment No.
122 232b
)
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Table 4.6.2l CONTROL ROOM AIR TREATMENT SYSTEM INITIATION Surveillance Re uirement Parameter (1) High Radiation Venti 1 ation Intake Sensor Check Once/shift Instrument Channel Test Once per quarter Instrument Channel Calibration Once each operating cycle not to exceed 24 months Amendment No. P
,e e
232c t
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BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION The set points on the generator load rejection and turbine stop valve closure scram trips are set to anticipate and mi nimize 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 operating power level, bypassing of the scrams below the specified power level is permissible.
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 deviations include inherent instrument error, operator setting error and drift of the set point.
These errors are compensated for in the transient analyses by c'onservatism 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 reaching the nominal set point and the worst set point due to error.
The maximum allowable set point deviations are listed below:
Neutron Flux APRH, +2.7X of rated neutron flux
- IRH,
+2.5X of rated neutron 'flux Recirculation Flow, +1'4 of rated recirculation flow Reactor
- Pressure,
+1.5.8 psig Containment
- Pressure,
+0.053 psig Reactor Hater Level, +2.6 inches of water Hain Steam Line Isolation Valve Position,
+2.5'L of stem position Scram Discharge
- Volume,
+0 and -1 gallon Condenser Low Vacuum, +0.5 inches of mercury Amendment Nos.
122 236
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P f-
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION High Flow-Hain Steam Line,
+1 psid High Flow-Emergency Cooling Line,
+1 psid High Area Temperature-Main Steam Line, +10F High Area Temperature-Clean-up and Shutdown,
+6F High Radiation-Hain Steam Line, +100K and -50K of set point value High Radiation-Emergency Cooling System Vent,
+1001. and -50K of set point High Radiation-Reactor Building Vent, +100K and -50'L of set point High Radiation-Refueling
- Platform,
+100K and -50'L of set point High Radiation-Offgas Line, +50K of set point, (Appendix D)*
The test intervals for the trip systems result to calculated failure probabilities (10-4 which corresponds to the proposed IEEE Criteria for System Failure Probability.
(IEEE SG-3, Information Docket ¹1 Protection System Reliability, April 24, 1968).
The test intervals for the trip systems result in calculated failure probabilities ranging from 6.7 x
10 7 to 1.76 x
10-10 (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 t line isolation valves can only be done during periods of Station shutdown.
These functions 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 Amendment Nos.
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t 237
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f P
BASES 3.6.11 AND 4.6.11 ACCIDENT MONITORING INSTRUMENTATION Accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.
This capability is consistent with the recommendations of NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations,"
and/or NUREG-0737, "Clarification of THI Action Plan Requirements,"
November 1980 and NUREG 0661, "Safety Evaluation Report Hark I Containment Long Term Program.".
The maximum allowable setpoint deviation for the Suppression Chamber Hater Level instrumentation is
+ l.S inches.
(
'Amendment Nos.
7 122 241 i i
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