ML20024E951
| ML20024E951 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  (DPR-059) |
| Issue date: | 08/26/1983 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20024E952 | List: |
| References | |
| DPR-59-A-075 NUDOCS 8309070477 | |
| Download: ML20024E951 (14) | |
Text
?th UNITED STATES
- e '%
NUCLEAR REGULATORY COMMISSION fi wAsmucTos.o c.20sss
.9
$l,;$tg',5 di 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. 75 License No. DPR-59 i
The Nuclear Regulatory Commission (the Commission) has found that:
1.
The application for amendment by the Power Authority of A.
the State of New York (the licensee) dated July 7,1983 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; The facility will operate in conformity' with the appif cation, B.
the provisions of the Act, and the rules and regulations of the Commission; There'is reasonable assurance (i) that the activities authorized C.
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;
.The issuance of this amendment will not be inimical to the D.
common defense and security or to the health and safety of l
the public; and l
The issuance of this amendment is in accordance with 10 CFR I
E.
Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
l Accordingly, the amendment authorizes the removal of the interim 2.
conditions to Facility Operating License No. DPR-59 that were implemented by the Commission's Orders of October 2,1980, and L
i January 9,1981; and fulfills the conditions specified_in the l
Commission's Order of June 24, 1983.
The license is further amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) l-of Facility Operating License No. DPR-59 is hereby amended to l
read as follows:
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.8309070477 830826 DR ADOCK 05000333 i
(2) Technical Specifications The Technical Specifications contained-in Appendices A~
and B,. as revised through Amendment No. 75, 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.
fgI FOR THE NUCLEAR REGULATORY COMMISSION
^
wL Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Speci fica tions -
Date of Issuance: August'26, 1983 4
4 t
I
-....=__..
E ATTACHMENT TO LICENSE AMEf'DMENT NO.
FACILITY OPERATING LICENSE NO. DPR-59
- DOCKET NO.-50-333 5
Revise the Appendix "A" Technical' Specifications as follows:
- Remove Replace i-6a-32 32
-34 34 41 a.'
41a
-44 44 '
45a 45a 46 46 t
72 72 73 73 l
89a' 89a 95 96 i
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JAFNPP 2
Z.
Top of Active Fuel The Top of Active Fuel, corresponding to the top of the enriched fuel column of each fuel bundle, is located 352.5 inches above vessel zero, which is the lowest point in the inside bottom of the reactor vessel. (S)e General Electric drawing No. 919D690BD.)
AA.
Rod Density Rod density is the number of control rod notches inserted expressed as a fraction of the total number of control rod notches. All rods fully inserted is a condition representing 100 percent rod density.
Amendment No. 75 6a i
. =.
JAFilPP 1-3.1 BASES.
The reactor protection system automatically initiates The outputs of the subchannels are combined in a 1 out of 2 logic; i.e., an input signal on a reactor scram to either one or both of the subchannels will cause
- 1. - Preserve the integrity of the fuel cladding.
a trip system trip. The outputs of the trip aystems are arranged so that a trip on both 2.
Preserve the integrity of,the Reactor Coolant systems is required to produce a reactor scram.
System.
This system meets the intent of IEEE-279 (1971) 3.
Minimize the energy which must be absorbed for fluclear Power Plant Protection Systems. The following a loss of coolant accident, and system has a reliability greater 'than that of a prevent inadvertent criticality.
2 out of 3 system and somewllat less than that of a 1 out of 2 system.
This specification provides the limiting conditions for operation necessary to preserve the ability With the exception of the average power range of the system to perform its intended function monitor ( APRM) -channel the. intermediate range even during periods Ehen instrument channels may monitor (IRM) channels, the ' scram discharge volume, be out of service because of maintenance. When the main steam isolation valve closure and the necessary, one channel may be made inot3erable for turbine stop valve closure, each subchannel has brief intervals to conduct required functional one instrument channel. When the minimum tests and calibrations.
condition for operation on the numh'er of operable instrument channels per untripped protection 1-The Reactor Protec' tion System is of the dual channel trip systen. la met or if it'cannot be met and the type (Reference subsection 7.2 FSAR). The System affected protection trip system is placed in a l
ts made up of two independent trip systems, each tripped condition, the effectiveness of the i
huving two subchannels of tripping devices.Each protection system is preserved.
subchannel has'an input from at least one instrument i
channel which monitors a critical parameter.
Three APlut instrument channels are provided for each protection trip system. APRM's A and E j
operate contacts in one subchannel and APRH's j
~
I C and E operate contacts in the other J
l AmendmentNo.,F(,j7I, 75
~
32
JAFilPP o
3.1 BASES (cont'd) the IRM and APM are required in the refuel and
- Thus, is discharged from the reactor by a scram can startup/ hot standby modes.
In the power range Each be accommodated in the discharge piping.
the APRM System provides required protection scram discharge instrument volume accommodates (reference paragraph 7. 5.7 FGAR).
Thus the IRM is the low The APRM's in excess of 34 gallons of water and System is not required in the run mode.
point in the piping.130 credit was taken for this cover only the power range.
The IRM's and APRM's volume in the design of the discharge piping as provide adequate coverage in the startup and cqqcerns the. amount of water which must be accommodated intermediate range.
during a scram.
'The high reactor pressure, high drywell pressire, During normal operation the discharge volume reactor low water. level and scram discharge volume J
,is empty; however, should'it fill with water, high level scrams are required for startup and run the water discharggd to the piping from modes of plant operation. They are, therefore, the reactor could not bh accommodated, which would required to be operational fo'r these modes of result in slow scram times or partial control rod reactor operation.
level insertion. To preclude this occurrence, Idetectioninstrumentshavebeenprovidedindach The requirement to have thu scram functions instrument volume which alarm and scram the reactor indicated'in Table 3.1-1 operable 'in the refuel modo assures that shifting to the, refuel mode reaches 34.5 gallons, when the volume.of water As indicated above, there is sufficient volume in during reactor power operation does not diminish the piping to accommodate the scram without the protection provided by the Beactor Protection impairment of the' scram times or scount of insertion This function shuts the reactor System.
of the control rods.
l down while suf ficient volume remains to acconunodate Turbine stop valve clos,ure occurs at 10 percent the discharged water and precludes the situation Below 217 psig turbine first of valvo closure.
in which a scram would be required bat not be able (30 percent of rated), the scram stage pressure is to perform itc fune'lon adequately.
signal due to turbine stop valve closure bypassed because the flux and pressure scrams are d
A Source Range Monitor (SRM) System is also provided adequate to protect the reactor.
to supply additional neutron Icvel information scram functions during startup but has no (reference paragraph 7.5.4 PSAR).
{
34
/
j Amendment No. 75
~
'ilAFNPP l-
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3 l
TABl$1 3.1-1 (cont'd) f
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'y 2
3
.t REACTGR PHOTECTION SYSTI31 (SCRAll) It1STRillM11TATION REQtlIREMEtIT
'\\
)
Minimum No.
Mxles in $11Eich '
Total
, Functiod Hust De thimber of f
of Operable
(
Instrument.
+ Operable Instrument 1
Channels Trip ninction Trip Level Channels Action
)
8-Provided (1) per Trip Setting System (1) y
,g-N by Design 4
Refuel St.artup Hun
- for Both 1
i
\\,
'(6)
Trip' Systems i
y v
i 2
APRM Downscale 12.5 indicatect en :
'X 6 Instrument A*or B
(
l S
Charinclus i
scale (9) i s
s 2
Illgh hea<itor 51045 psig
~r
- X 4 Instrument A
X(0) / \\ X a!
'f Channels Pressure ~
/
,'t' s
2 liigh Drywell f2.7psig
'X (7 )
X(7)
X 4 Instrument A
Channels Pressure
>s
? : '
3 1
I
(
i 2
Reactor tlow Water.
?,I 2. 5 41n. indicated X
X X
4 InpErument A
, Idvc.L-Channels Level g
(AIT7 in aboise the y
i
,fcp of active fuel) s
3 Iligh Water Level 53[.5 gallons per X(2)
X X
o Instrument A
in Scram Discharge Instrument Volume Channels Volume
~
s N
i Si
' <3x normal full X
Xx
.X 4 Instrument A
2 Main > Steam lind
~ power background Channels Iligh Radiation N
$10Cvalve
X (3) (5) X(3) (5) X(5)
U Instrument A
Isolation Valve closure Channels Closure Amendment lio.
),
, 75 41a N
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Table 4.lY4'
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. t, REAC'IOR PROTECTION SYSTEM (SCRAM) 'IflS'lltUttEllT FmiCTIONAL TESTS i
EINIMUM FUNCTIONAL TEST FREQUENCIES FOR SAFl?rY IllSTi1UllEttf AND COtrlROL CIRCUITS
/
l Instrument Channel b Group Functiopal 3 st
' Minimum Frequency (3) is Shu'down A
Place shle_ Switch in Shutdown Each refueling outage.
Mode Switch it t
i' Manual Scram
/
A
'Irlp Channel and Alarm
,, -Every 3 rentl.s.
x RPS Channel Tedt S tch A'
- Trip Channel and Alarm
['
Every refueling outage or
^
after phannel maintenance.
/~ _
IRH a
b liigh Flux C
Trip Channel and Alarm (4)
Onco per week ditring re-fueling or startup and before cach startup.
,4 J
s InoFerative E C Trip Chatinel and Alarm (47 Obce'por Eeck during re-s fueling or startup and bcfore cach startup.
APRM Iligh Flun B.
Trip output Relays (4)
, Once/ week.-
f Inoperative B,
Trip output Relays (4)
Once/ week Downscale B
Trip output Relays (4)
Once/ week Flow Dias B
Calibrato Flow Dias Signal (4)
' Once/ month (1)
Itigh Flux in Startup or Refuel C
Trip Output Relaya(4)
Once per week during refueling or startup and before each startup.
Iligh Reactor Pressure B
Trip Channel and Alarm (4)
Once/ month. (1) (Instrument check once per da:
liigh Drywell Pressure A
Trip Channel and Alarm Once/ month (1)
Reactor Iow Water Level (5)
A Trip Channel and Alarm Once/ month (1) liigh Water Imvel in Scram A
Trip Channel once/ month (7)
Discharge Instrument Volume IDischargeInstrumentv'olume liigh Water Ievel in Scram n
Trip Channel avid Alarm once/ month Main Steam Line Iligh Radiation il Trip Channel and Alarm (4)
Once/ week.
f 44 Amendment flo. f, pd, 75
- s.
JAFNPP Table 4.1-1 (cont'd)
REACTOR PROTECTIOrl SYSTEM (SCitAli)ltlSTidit1EllT VUllCTIOllAI. Tl;STS
- MINIMUM FUtlCTIONAb TEST FREQUEllCIES FOlt SAFETY IllSTRiit1EllT AllD COf fritOL CIRCUITS 3:' ~
,/
IIOTES FOR TABLE 4.1-1 (cont'd)
The water level in the reactor vessel will be perturbed and the correslonding Icvet indicator changes will be This perturbation test will be performed every month after completion of tlie functional test 5.
monitored.
program.
6.
Deleted.
The funct[onal test'shall be performed utilizing a water column or similar device to prc,sride assurance that:
7.
le detecte'd.
damage to a float or other portions of the float assembly sill s
i Amendment-130
- 24. 6d. 75 1'i a
.IAFilPP Tablo 4.1-2 9
REACTOR _PROTECTIOtl SYSTEtt (SCitAM) IllSTittlttEllT CALIDilATIOtl MIt11MilM CAI.IBIIATION FitEQUEtICIES FOR REACTOlt PitOTECTIOli IllSTRiftlEllT CitAtitlELS
-9
- =.
Minimum Frequency Once/wdek Instrument Channel Group (1)
Calibration (4)
C Comparison to APRM on Maximum frequency once/ week IBM Iligh Flux ControlIed Shutdowns APRM liigh Flux Output Signal ~
D lleat Dalance Daily Internal Power and Evcity refueling outage n
Flow Bias Signal Flow Test with Stan-dard Pressure Source n
TIP System Traverne Every 1000 effective full LPRtt Signal power hours B
Standard Pressure Once/ operating cycle Illgh Reactor Pressure Source A
Standard Pressure Source Every 3 months liigh Drywell Pressure Reactor Low Water Level A
Pressure Standard Every 3 months Illgh Water Level in Scram Dis-A flater Column, llote (6)
Once/ operating cycle, llote (6) charge Ipstrument Volume D
Standard Pressure Source Every 3 months l Illgh Water level in Scrata Discharge Instrument Volume flain Steam Line Isolation Valve A
tiote (5) llote (5)
Closure Main Steam Line High Radiation n
Standard Current Source (3)
Every 3 months Turbine Plant Stage Pressure A
Standard Pressure Source Every 6 months i
Permissive Turbine Control Valve Past Closure A
Standard Pressure Source Once/ operating cycle Oil Pressure Trip 1
/
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=
JAFNPP TABLE'3.2-3 INSTRUllENTATIOtt TIIAT IHiiIATES cot 3 TROL ROD DLOCKS f
Minimum no.
Total Humber of of Operable Instrument-Instrument Trip Level Setting Instrument Channels Action Provided by Design Channels Per for 80th Channels Trip System 2
APlitM Upscale (Flow Biased) s f (0.66Wl42%)xFRP 6 Inst. Channels (1)
HFLPD 2
APRM Upscale (Start-up f12%
6 Inst. Channels (1)
Mode) 2
.APRM Downscale
~12.5 indicated on scale 6 Inst." Channels (1)
Char $nels (1)
J 1 (6) liod. Block Monitor
~
S $0.66WIK (a) 2 Inst.
(Flow Biased) 1 (6)
_ Rod-Block Honitor 12.5 indicated 2 Inst. Channels (1)
(Downscale) on scale 3
IRM Downscale,(2) 1 2% of tull scale 8 Inst. Ch'annels (1) 3 IRM Detector not in (7) 8 Inst.. Channels (1)
Start-up Position 3
IRM Upscale
<86.4% of full scale 8 Inst. Channels (1) i 2 (4)
SRM Detector not in (3) 4 Inst. Channels (1)
Start-up position' 5
2 (4) (5)
SRM Upscale
{10 counts /sec 4 Inst. Channels (1) 1 Scram Discharge Instrument f26.0' gallons per 2 Inst. Channels (9) (10)
Volume High Water Level instrument volume 1
_ NOTES FOR TABLE 3.2-3 1.
For the Start-up and-Rdn positions of t.he Reactor 11 ode Selector Switch, there shall be two operable er.
tripped trip systems for each function.
Th'o SitM and 'IltH block need not be operable in run mode, and Amendment No. jd, fd, 75 72
s JAFNPP TABLE 3.2-3 (Cont'd)
INSTRUMENTATION TIIAT INITIATES CONTROL ROD BLOCi'S NO'IES FOR TADIE. 3.2-3 i
f the APRM and RBH rod blocks need not be operable in start-up mode.
From and after the time it is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter3 if this condition lasts longer than seven days, the system shall be tripped.
From and after the time it is found that the first column cannot be met for both trip systems, the systems shall be tripped.
2.
IRM downscale is bypassed when it is on its lowest range.
3.
This function is bypassed when the count rate is.Z.100* cps.
4.
One of the four SRM inputs may be bypassed.
5.
This'SRM Function is bypassed when the IBM range switches are on range 8 or above.
6.
The trip is bypassed when the reactor power is $
30s.
7.
This function is bypassed when the Mode Switch is.placed in Run.
8.
S = Rod Block Monitor Setting in percent of initial.
W = Inop recirculation ficw in percent of rated K = Intercept values of 39% 40%, 41% and 42% can be used with appropriate 11CPR limits from Section 3.1.B.
9.
When the reactor is subcritical and the reactor water temimrature is less than 212 F, the control rod block is required to be operable only if any control rod in a control cell containing fuel' is not fully inserted.
10.
Wnen one of the instruments associated with scram discharge instrument volume high water' rod blocks l
is not operable, the trip system shall be tripped.
g W
Amendment No.Jas, pf, J,'75 f
73
JAFNPP C) 'b.
The' control rod directional control f.
The scram discharge. volume drain and ai.D id-valves for inoperable control rods vent valves shall each be full-travel shall be disarmed electrically.
cycled at least once per quarter to verify that the valves close in less c.
Control rods with scram times greater than 30 seconds and to assure proper than those permitted by Specification valve stroke and operation.
3.3.C.3 are inoperable, but if.they can be inserted with control rod drive pressure they-need not be disarmed g.
At least once per operating. cycle, the electrically.
operability of the entire scram dis-charge system as an integrated whole d.
Control rods with a failec1 " Full-in" shall be demonstrated by a scram of control rods from a normal control rod or " Full-cut", position switch may configuration of less than or equal-to be bypassed in the' Rod Sequence Control System and considered 50% rod density by verifying that.the drain and vent valves:
operable if the actual rod position is known.
These rods must be moved in.
1.
Close upon receipt of a signal for e.
When it is initially determined that ntrol rods to scram; and a control rod is incapable of normal 2.
.Open when the scram signal is reset.
Insertion, an attempt to fully insert the control rod shall be made.
If the control rod cannot-be fully This requirement may be satisfied as part inserted; of any scram originating from the rod.
density conditions specified above, pro-shutdown margin test shall be made vided that Specification 4.3.A.2.f is to demonstrate under this condition independently satisfied during the that the core can be made subcritical quarter in which the scram occurs.
for any reactivity condition during the remainder of the operating cycle with the analytically determined, highest worth control rod capable of withdrawal, fully withdrawn, and all other control rods capable of in-sertion fully inserted.
If Specification 3.3.A.1 and 4.3.A.] are met, reactor startup may proceed.
gf Amendment NO. /5, pd, 75 89a
JAFNPP 4.3 (cont'd) 3.3 (con t' d) 2.
The average of the scram insertion 2.
At 8-week intervals, 15 percent of times for the three. fastest the-operable control rod drives shall be scram timed above 950 psig. When-operable control rod.s of all groups of four control rods.in a two-by-two ever such scram time measurements are
-array shall be no greater than:
made, an evaluation shall be made to provide reasonable assurance that-Control Rod Average Scram proper control rod drive performance Notch Position Insertion Time is.being maintained.
Observed (Sec.)
3.
All control rods shyll be determined 46 0.361 operable once each operating cycle 38 0.977 be demonstrating the scram discharge 24 2.112 volume drain and, vent valves operable 04 3.764 when the scram test initiated by placing the' mode switch in the SilOTDOWN i
position is. performed as required by Table 4.1-1 and by verifying that the drain and. vent valves:
a.
'Close in less than 30 seconds af ter l receipt of a signal for. control rods to scram, and b.
Open when tie scram signal is reset or the scram discharge instrument volume trip is. bypassed.
D e
9 e
d 96 f,
d, 75-Amendment No.