ML17054A710
| ML17054A710 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/08/1984 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Niagara Mohawk Power Corp |
| Shared Package | |
| ML17054A711 | List: |
| References | |
| DPR-63-A-060 NUDOCS 8405210676 | |
| Download: ML17054A710 (14) | |
Text
t UNITED STATES t
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.
60 License No. DPR-63 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated January 5,, 1984, 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.
D.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangerino the health and safety of the public, and (ii) that such 'activities will be conducted in compliance with the Commission's regulations; 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.
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:
840M10676 840508 PDR ADOCK 05000220 P
I'
'I
(2) Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 60, 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 COPWISSION
Attachment:
Changes to the Technical Speci ficati ons i
Date of Issuance:
tray 8, 1984 Domenic B. Vassallo, Chief Operating Reactors Branch 82 Division of Licensing
ATTACHMENT TO LICENSE AMENDMENT NO.
60 FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Revise the Appendix A Technical Specifications by removing and inserting the following pages:
Existing
~Pa e
118 205 206 235 Revised
~Pa e
118 205 206 235 The revised areas are indicated by marginal lines.
LIMITING COND II'IONS FOR OPFRAT ION Table 3.2.7 REACTIH"COOLANT SYSTEM ISOLATION VALVES
~L4>e or E stew Location Relative No. of Valves to Primary Normal
~Each Line Contalnaent Posltlon Hotlve Power Maximum Action on lnlt Iating Signal Oper.
Time Initiating (A}I Valves Nave Malo Steam (Two Ones}
Main Steaw Mar~m-u
'Pwo Gneiss Main Steam-Emer enc Coolin Vents Tfwo Ones Feedwater TTwo Clnes}
Emerge~ac Coolln Steam Leayin Reactor
)T~ Ones Condenser Return to Reactor Tmmlnes Reactor Cleanu 1
Inside I
Outside Outside Outside Outs lde Outside Outside Outs Ide Inside Outs Ide Open Open A'1.P.O.
A. I.P.O.
10 10 Closed A.I.P.O.
Open A, I.P.O.
38 Open A.I.P.O.
38 Selt Act. Ck.
Closed A.I.P.O.
60 Open A, I.P.O.
5 Open R.H.P.O. a 60 Self Act. Ck.
Close Close Close Close Close Close Close Reactor water level Iow-low, or main steam line high radiation, or main steam line high flow, or Iow condenser vacuum, or high temperature in the pipe tunnel iiigh system flow Mater Leav in Reactor 7t}ice TTne Mater Return to Reactor ti}ne Gne}
Shutdown Coolin Mater Leavin Reactor
$0neGne Mater Return to Reactor w
Inside Outs lde inside Outside Ins ide Outs lde lns ide Outside Open Open Open Closed Closed Closed A.I.P.O.
A. I.P.O.
A.I.P.O.
Self Act. Ck.
A.1.P.O.
A.I.P.O.
A. I.P.O Self Act. Ck.
18 18 18 40 40 40 Close Close Close Close Close Close Reactor water level low-low, or high area temperature, liquid poison initiation or high system pressure, or low system flow, or high system temperature Reactor water level low-low, or high area temperature 118 Amendment No.
60
Table 3.6.2c INSTRUMENTATION THAT INITIATES OR ISOLATES EMERGENCY COOLING Limitin Condition for 0 eration Parameter Minimum No.
of Tripped or Operable
~Tri S stems Minimum Ho. of Operable Instrument Channels per Operable
~Ti i Set-Point Reactor Mode Switch Position in Which Function Must Be
~11 C
~
sty o
~
a.
0 Q
V C
OC S-CU CX ill C/)
EMERGENCY COOLING fAIYMAf6R Ti~Egg-Reactor Pressure (1080 psig (b)
X X
(2)
Low-Low Reactor Water Level JS inches Tindicator Scale)
(b)
X X
EMERGENCY COOLING TAo.Qfot~
Tier eaoii of teo systems)
(3)
High Steam Flow Emergency Cooling System 2(a)
.19 psid X
X 205 Amendment Ho.
60
Table 4.6.2c IHSTRUHEHTATIOH THAT INITIATES OR ISOLATES EMERGENCY-COOLING Surveillance Re uirement Parameter Instrument Sensor Check Channel Test Instrument Channel Calibration ENERGENCY COOLING IJifflNEIGH tTJ if~g Reactor Pressure (2)
Low-Low Reactor Mater Level Hone Once/day Once per month(c)
Once per month(c)
Once per 3 months(c)
Once per 3 months(<)
EMERGENCY COOLING ISGLATIIA
+or each of two systems)
(3)
High Steam Flow Emergency Cooling System Hone Once per 3 months(c)
Once per 3 oonths(c) 206 Amendment Ho.
60
BASES FOR 3.6.2 AND 4.6.2 PROTECTIVE INSTRUMENTATION a.
The set points included in the tables are those used in the transient analysis and the accident analysis.
The high flow set point for thy main steam line is 105 psi differential.
This represents a flow of approximately 4.4xlOo ib/hr.
The high flow set point for the emergency cooling system supply line is 19 psi differential.
This represents a flow of approximately 8.7xllP at rated conditions.
Normal background for the main steam line radiation monitors is defined as the radiation level which exists.in the vicinity of main steam lines after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or more of sustained full rated power.
The dose rate at the monitor due to activity from the control rod -drop accident of Appendix E or from gross failure of one rod with complete fission product release from the rod would exceed the normal background at the monitor.
The automatic initiation signals for the emergency cooling systems have to be sustained for more than 10 seconds to cause opening of the return valves.
If the signals last for less than 10 seconds, the emergency cooling system operating will not be automatically initiated.
The high level in the scram discharge volVme is provided to assure that there is still sufficient free volume in the discharge system to receive the control rod drives discharge.
Following a
- scram, bypassing is permitted to allow draining of the discharge volume and resetting of the reactor protection system relays.
Since all control rods are completely inserted following a scram and since the bypass of this particular scram initiates a control rod block, it is permissible to bypass this scram function.
The scram trip associated with the shutdown position of the mode switch can be reset after 10 seconds.
The'condenser low vacuum, low-low vacuum and the main steam line isolation valve position signals are bypassed in the startup and refuel positions of the reactor mode switch when the reactor pressure is less than 600 psig.
These are bypassed to allow warmup of the main steam lines and a
heat sink during startup.
235 Amendment No.
60