ML20207E675
| ML20207E675 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/19/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20207E679 | List: |
| References | |
| DPR-59-A-103 NUDOCS 8701020253 | |
| Download: ML20207E675 (7) | |
Text
ancuq'o UNITED STATES
^g 8
NUCLEAR REGULATORY COMMISSION o
y WASHINGTON, D. C. 20656
\\...../
POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET N0. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.103 License No. DPR-59 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Power Authority of the State.
ofNewYork(thelicensee)datedJune 25, 1986, complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954,
, and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.
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 (1) 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.
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-59 is hereby amended to read as follows:
8701020253 861219 PDR ADOCK 05000333 P
PDR i
- (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.103, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: December 19, 1986
ATTACHMENT TO LICENSE AMENDMENT NO.103 FACILITY OPERATING LICENSE NO DPR-59 DOCKET NO. 50-333 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.
Pages 55 56 64 206 i
,.nc-
i JAFWPp 3.2 DASES In addition to reactor protection instrumentation has a direct bearing on safety, are chosen at a level which initiates a reactor seren, protective instru-away from the nocimal operating range to prevent ined-mentation has been provided Which initiated action to vertent actuation of the safety system involved and l
mitigate the consequences of accidents which are exposure to abnormal situations.
beyond the operator's ability to control, or termi-nates operator errors before they result in serious Actuation of primary containment valves is initiated consequences.
This set of specifications provides by protective instrumentation shown in Table 3.2-1 the limiting conditions of operation for the primary which senses the conditions for which isolation is system isolation function, initiation of the Core required.
Such instrumentation must be available cooling Systems. Control Rod Block and Standby Gas whenever primary containment integrity is required.
Treatment Systems.
The objectives of the specifica-tions are to assure the effectiveness of the protec-The instrumentation which initiates primary system tive instrumentation when required.
even during isolation is connected in a dual bus arrangement, periods when portions of such systems are out of service for maintenance, and to prescribe the trip The low water level instrumentation set to trip at settings required to assure adequate performance.
177 in. above the top of the active fuel closes all When necessaryg one channel may be made inoperable isolation valves except those in Group 1.
Details of.
for brief intervals to conduct required functional valve grouping and required closing times are given tests and calibrations, in Specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent j
some of the settings on the instrumentation that uncovering the core in the case of a break in the initiate or control core and containment cooling have largest line assuming a 60 see valve closing time.
I tolerances explicitly stated Where the high and low Required closing times are less than this.
l values are both critical and may have a substantial l
effect on safety. The set points of other instrumen-The low-low reactor water level instrumentation is tation, where only the high or low end of the setting set to trip when reactor water level is 126.5 in.
j above the top of active fuel
(-38 in.
on the instrument). This trip 4
1 i
I i
?
103 Amendment No.
55
1 JAFNPP Initiates the NpCI and RCIC and trips the the breaks discussed above, this lastrumentation will lrectrostattom pumps.
The low-low-low reactor watec generally initiate ECCS operation before the low-low-level instrumentation is set to trip when the water low water level lastrumentation; thus the results I
level is 18 in, above the top of active fuel.
This given above are applicable here also.
See Specif!-
trip activates the rematador of the ECCS subsystems, catloa 3.7 for isolation valve closure group.
The closes the mala steam isolatloa valves, main steam water level lastrumentation lattistes protectica for line drain valves and reactor water sample line the full spectrum of loss-of-coolant accidents.
isolattos valves, and starts the emergency diesel j
generators.
These trip level settings were chosen to Yeaturis are provided la the mala steen 11aes as a means of measurlag steam flow and also 11mittag the be high enough to prevent spurious actuation but low enough to taltista ECCS operation and primary system loss.of mass laventory feed the vessel declag a steam j
isolation so that post-accident coo 11ag can be 11ae break accident.
The primary function of the accomp11shed and the guidelines of 10CFt100 will not instrumentation is to detect a break la the main i
be exceeded.
For large breaks up to the complete steam 11ae.
For the worst case accident, mala steam circumferential break of a 24 la. recirculation line line break outside the drywell, a trip settlag of 140 av*
with the trip setting given
- above, ECCS percent of rated steam flow la conjunctica with the i
lattiation and primary system isolation are lattiated flow limiters and main steam line valve closure, la time to meet the above criteria.
Reference limits the mass leventory loss such that fuel is not paragraph 6.5.3.1 FSAR.
uncovered, fuel temperatiere peak at approximately 1,000*F and release of radioactivity to the environs The high drywell pressure instrumentation is a diverse is below 10CFR100 guidelines.
Reference Section signal for malfunctions to the water level lastrumen-14.6.5 FSAR.
tation and la addition to lattisting ECCS, it causes isolation of Groupe B and 3 1solatfor. valves. For l
1 i
e Amendment. No. J, )d, )$=103 d
56
JAFNPF l
TABLE 3.2-1 I w arrog imum--_- aATION TRAT 1marxAuma Pgmaar wegar--
m 1
Total Number of Instrument Mlalaum Number of Chamaels Provided by Destga Action Cperable Instrument Chamaels per Trio System (1)
Instrument Trio Level settlam for Both Trio systems (2) l l
2 (6) teactor Low Water 5 12.5 la. Indicated 4 Inst. Channels A
l Level Level () }77 in. above the. top o active fuel) 1 Reactor Righ Pressure f75 pelg 2 Inst. Chamaels D
(Shutdows Cooling Isola. tion) 2 Reactor Low-Low-Low
)18 in. above the top of 4 Inst. Channels A
Water Level active fuel.
2 (6)
~
Nigh Drywell Pressure (2.7 pelg 4 Inst. Channels A
2 Nigh Radiatloc Mala
( 3 x Normal Rated 4 Inst. Chaamels 3
Steen Line Tunnel Full Power Background (9) 2 Low Pressure Malm 5 825 pels (7) 4 Inst. Chamaels B
Steam Line l
2 Nigh Flow Main Steam
( 140% of Rated Steam 4 Inst. Channels B
Line Flow 2
Main Steam Line Leak f40*F above mar 4 Inst. Channels 8
Detection High ambient Temperature 3
Reactor Cleanup sys-f40*F above maz 6 Inst. Chamaels C
tem Equipment Area ambient Nigh Temperature 2
Low Condenser Vacuum 58" Ng. Vac (S) 4 Inst. Channels B
Closes MSIV's Amendment No. % f[, F Pf. f,103 I
d 64
JAFNPP NOTES FOR TABLE 3.7-1 ISOLATION SIGNAL CODES Simnal Description A*
Reactor vessel low water level - (A scram occurs at this level also. This is the highest of the three low water level signals.
38 Reactor vessel low-low-low water level - (This is the lowest of the three low wate: level signals.
CS High radiation - main steam line D*
Line break - main steam line (steam line high steam flow)
E*
Line break - main steam line (steam line high temperature)
F*
High drywell pressure G
Reactor vessel low water level or high drywell pressure (Emergency Core Cooling Systems are started)
E JS Line break in teactor Water Cleanup System - high space temperature K*
Line break in RCIC System steam line to turbine (high S
steam line space temperature, high steam flow, low steam line pressure, or high turbine exhaust pressure)
La Line break in HPCI system steam line to turbine (high steam line space temperature, high steam flow. Iow steam line pressure, or high turbine exhaust pressure)
M l
PS Low main steam line pressure at inlet to main turbine (RUN mode only)
S Low drywell pressure T
Low reactor pressure permissive to open core spray and RN1-LPCI valves
- These are the isolation functions of the Primary Containment and Reactor Yessel Isolation Control System; other functions are given for information only.
1 of 4 Anendseat No. )<f 103 206
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