ML20128G295
| ML20128G295 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/16/1985 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20128G298 | List: |
| References | |
| DPR-59-A-090 NUDOCS 8505300114 | |
| Download: ML20128G295 (9) | |
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C' surgk UNITED STATES
/.
- 3:
NUCLEAR REGULATORY COMMISSION O
'j WASHINGTON, D. C. 20666
%...**.l POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 90 License No. DPR-59 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Power Authority of the State ofNewYork(thelicensee)datedOctober2,1984assupplemented October 22, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 Comission'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 Comission'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 Ifeense amendmentandparagraph2.C(2)ofFacilityOperatingLicenseNo.
DPR-59 is hereby amended to read as follows:
@ M k Wo b 33 p
., (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 90, 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 Domenic B. Yassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
May 16, 1985 i
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e ATTACHMENT TO LICENSE AMENDMENT NO. 90 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise the Appendix "A" Technical Specifications as follows:
Remove Insert 33 33 41a 41a 43a 43a 57 57 64 64 65 65
v souice cf such radiation to the extent 3.1 BASES (cont'd)
JAFNPP necessary to prevent aww=cive turbine contamination. Discharge of sweanive amounts of radioactivity to the site subchannel. APM's B, D and F are arranged environs is prevented by the air ejector cuntlarly in the other protection trip offgas monitors which cause an isolation systen. Each protection trip systen has of the main condenser offgas line. During one more APM than is necessary to meet the the Hydrogen Addition 1bst, the normal back-miniansa nasuber required per channel. this ground Main Steam Line Dadiation Imvel is allows the bypassing of one APM per protec-expected to increase by a factor of appresti-tion trip systesu for maintenance, testing notely 5 at the =wi== hythogen addition or calibration. Additional I m channels rate as indicated in note 16, Table 3.1-1.
have also been provided to allow for Y 1he scrasa setpoint will be reset to three passing of one such channel. 1he ha n for times the projected background radiadon the scram setting for the IM, APM, high level prior to performance of the test.
reactor pressure, reactor low water level, The setpoint will be restored to nonnal main steam isolation valve (MSIV) closure, following ccupletion of the hydrogen and generator load rejection, turbine stoP acMition test..
valve closure are discussed in Sections 2.1 and 2.2.
A Reactor Mode Switch is provided which actuates or byn===aa the various scrasa Instrmentation (pressure switches) for the functions appropriate to the particular plant
. drywell are provided to detect a loss of operating status. Reference paragraph 7.2.3.7 coolant accident and initiate the core FSAR.
stan&y cooling equipnent. A high drywell pressure scram is provided at the saune the manual scran function is active in all I
setting as the Core and Contairunent Cooling modes, thus providing for a manual means of I
Systms (BOCS) initiation to minimize the rapidly inserting control rods during all energy which naast be +A.
- -hted during a modes of reactor operation.
loss-of-coolant accident and to prevent return to criticality. This instruentation 1he APM (high flux in startup or refuel) i is a backup to the reactor vessel water Systesa provides protection against excessive level instrmentation.
power levels and short reactor periods in the startup and intermediate power ranges, liigh radiation levels in the :nain steam line tunnel above that due to the normal.
1he IM System provides protection against nitrogen and oxygen radioactivity are an short reactor periods in these ranges.
indication of leaking fuel. A scram is initiated whenever such radiation level 1he Control Rod Drive Scram Systesa is exceeds three times normal background.
designed so that all of the water which 1he purpose of this scram is to reduce the Amendment rio. )( 90 33
JAFNPP TABLE 3.1-1 (cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No.
Modes in which Total of Operable Function Must Be Number of Instrument Operable Instrument Channels Trip Function Trip Level Channels Action 1
per Trip Setting Provided by (1)
System (1)
Design for Refuel Startup Run Both Trip (6)
. Systems i
2 APRM Downscale 1 2.5 indicated on X
6 Instrument A or B '
scale (9)'
Channels 2
High Reactor 3 1045 psig X(8)
X X
4 Instrument A
Pressure Channels 2
High Drywell b 2.7 psig X(7)
X (7)
X 4 Instrument A-Pressure Channels 2
Reactor Iow Water 1 12.5 in. indicated X
X X
4 Instrument A
Level level Channels'
( 1177 in, above the
[
top of active fuel) 3 Ifigh Water Level S34.5 gallons per X(2)
X X
8 Instrument A
in Scram Discharge Instrument Voltane Channels Volume 2
Main Steam Line I3x normal full X
X X
4 Instrument A
l liigh Radiation power background (16)
Channels 4
< 10% valve closure X (3) (5) X (3) (5) X (5) 8 Instrument A
. Isolation valve Channels Closure i
hnendment tio. MI. g, fr'f, 7j, Af 90 41a
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e
- JAFIE'P
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Table 3.1-1 (e nt'd)'
REACIOR PIDITCTION SYS'ITM (SCRAM) INS't1DENFATION REQUIRENENP i
/~
N0rES OF TABIE 3.1-1 (Oont'd) i
- 14. 'Ihe APR4 flow biased high neutron flux signal.is fed through a. time constant circuit of approximately 6 seconds. 'Ihe APIN fixed high neutron flux signal.does not irm,vuurate
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the. time constant, but responds directly to instantaneous neutron flux.
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- 15. His Average Power Range )bnitor scram function is. fixed point and is increased when the reactor mode switch is placed-in the Run position.
1 16.* During the pwtuecd Hydrogen Addition 'Ibst, the nornal backgiound radiation level will -
increase by approximately a factor of 5 for peak hydrogen-concentration. '11erefore, i
prior to performance of. the test, the Main Steam Line Paxliation Monitor Trip Icvel i
Setpoint will be_ raised to <_.three times the increased radiation levels. %e test will
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be conducted at power levels > 80% of normal rated power.. During controlled power
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reduction, the setpoint will be readjusted prior to going below 20% rated power. If j
due to a recirculation ptmp trip or other unanticipated power reduction event, the
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reactor drops below 20% rated power without the setpoint change, control rod withdrawal' will be prohibited until the necessary trip setpoint adjustment is raade.
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- This specification is in effect only during Operating Cycle 7.
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Amendment No. ff' 90 43a
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A 4
3.2 BASES (cont'd) 3 i
High radiation monitors in the main steam line tunnel have been provided to detect
%e trip settings of 4 300 percent of design gross fuel failure as in the control rod fl w for this high flow of 40*F above maximum l
drop accident. With the established set W anbient f r high temperature are such that of 3 times normal background, and main steam uncovering the core is prevented and. fission i
line isolation valve closure, fission product Product release is within limits.
I release is limited so that 10 CFR 100 guide-lines are not exceeded for this accident.
%e RCIC high flow and temperature instrumentation Reference Section 14.6.1.2 FSAR. During the are arranged the same as that for the HPCI. De ifydrogen Addition Test, the normal background trip setting of & 300 percent for hfgh flow and Plain Stean Line Radiation Ievel is ewW 40*F above maximum anbient for temperature are to increase by approximately a factor of 5 at based on the same criteria as the IIPCI.
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the peak hydrogen concentration as indicated in note 16, Table 3.1-1.
With the hydrogen
% e reactor water cleanup system high flow tem-addition, the fission product release would Perature instrumentation are arranged similar still be well within the 10 CFR 100 guidelines-to that for the HPCI. %e trip settings are in the event of a control rod drop accident.
such that uncovering the core is prevented and fission product release is within limits.
Pressure instrumentation is provided to close the main steam isolation valves in the run mode
% e instrumentation which initiates ECCS action when the main steam line pressure drop below is arranged in a dual bus system. As for ot'her 825 psig. % e reactor pressure vessel thermal vital instrumentation arranged in this fashion, i
transient due to an inadvertent opening of the specification preserves the effectiveness of I
the turbine bypass valves when not in the cun the system even during periods when maintenance mode is less severe than the loss of feedwater or testing is being performed. An exception to i
analyzed in section 14.5 of the FSAR, therefore, this is when logic functional testing is being j
closure of the main steam isolation valves for performed.
thermal transient protection when not in the i
run mode is not required.
% e control rod block functions are provided to prevent excessive control rod withdrawal so that Wa IIPCI high flow and temperature instru-PCPR does not de-l mentation are provided to detect a break in the HPCI steam piping. Tripping of this instrumentation results in actuation of IIPCI isolation valves. Tripping logic for the high flow is a 1 out of 2 logic.
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90 57 Amendment tn.
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JNTPP-e' 3
TABIE 3.2-1 INSTRtMNFATION 'DIAT INITIATES PIUMARY CONFADOOFF ISOIATION Minhnum Number of
'Ibtal Ntaber of Instnnent Operable. Instrument Channels
. Trip level Setting for Both Trip Systems (2)
Channels Provided by Design Action-per Trip System (1)
Instnnent 2 (6)
Reactor Im Water 2 12.5 in, Indicated 4 Inst. Onannels A'
Ievel (b 77 in. above IcVel 1
the top of active fuel) 1
' Reactor High Pressure f 75 psig
,2 Inst. Otannels D
(Shutdown Cooling Isolation) 2 Reactor Im-Im 2 -38 in. indicated 4 Inst. Channels A
Water Ievel level (b 126.5'in. above the top of active fuel) 2 (6)
' liigh Drywell Pressure i 2.7 psig 4 Inst. Channels A
2 liigh Radiation Main i 3 x Normal Rated 4' Inst. Channels B
Steam Line Tunnel Full Power Background (9)-
2 Im Pressure Main h 825 psig (7) 4 Inst. Channels B
Steam Idne 2
High Flow Main Steam 6140% of Rated Steam 4 Inst. Channels B
j Line Flow i*
2 Main Steam Line Leak 6 40 P above max '.
4 Inst. Channels B
Detection liigh anbient q
Tenperature 3
Reactor Cleanup Sys-6 40 F above max 6 Inst. Channels C
ten Equipnent Area anbient High Tenperature a
2 Im condenser Vacutan D. 8" lig. Vac (8) 4 Inst. Channels B
Closes FEIV's hvendment tb. )M,,T1,)M(, J# 90 64
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Table 3.2-1 (Cmt'd)
INSTRLMINEATION THAT INITIATES PRIMARY CONTADEENT ISOIATION NcrrES m R TABLE 3.2-1 1.
Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each function.
2.
Fr:m and after the time it is found that the first column cannot be met for one of the trip systems, that trip system shall be tripped or the aypwgiate action listed below shall be taken.
A.
Initiate an orderly shutdown and have the reactor in cold shutdsn cmdition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have rrain steam lines isolated within eight houps.
C.
Isolate Reactor Water Cleanup Systen.
D.
Isolate shutdown cooling.
3.
Deleted 4.
Bleted 5.
'Iwo required for each steam line.
6.
These signals also start SBGTS and initiate secondary containment isolation.
7.
Only required in run mode (interlocked with Mode Switch).
8.
Bypassed when reactor pressure is less than 1005 psig and turbine stop valves are closed.
'Ihe trip level setpint will be maintained at <-3 times normal rated full l
-9.
power background. See note 16 to Table 3.1-1 for re-setting trip level setpoint just prior to the Hydrogen Addition Test, and re-setting of the -
l'ain Steam Line Padiation Ibnitor for power levels below 20%.
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Amentnent No. yl pr, frf 90 65 i
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