ML20043C515
| ML20043C515 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 05/30/1990 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20043C510 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 JPTS-89-015, JPTS-89-15, NUDOCS 9006050272 | |
| Download: ML20043C515 (38) | |
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i ATTACHMENT l PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING l
ACC15ENTR5RTTORING INSTRUMENTATION l
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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l
l 900605gg7k-33 POR A
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JAFNPP LIST OF TABLES Table Title P,agg 3.1 1 Reactor Protection System (Scram) instrumentation Raquirement 41 4.1 1 Reactor Protection System (Scram) Instrument Functlanal Tests 44 4.1 2 Reactor Protection System (Scram) Instrument Calibration 46 3.21 Instrumentation that initiates Primary Containment isola 11on 64 3.2 2 instrumentation that initiates or Controls the Core and Containment 66 Cooling Systems l
3.2 3 -
Instrumentation that initiates Control Rod Blocks 72 3.2-4 (DELETED) 74 3.2 5 Instrumentation that Monitors Leakage Detection inside the Drywell 75 l
3.2-6 (DELETED) 76 3.2-7 Instrumentation that initiates Recirculation Pump Trip 77 3.2 8 Accident Monitoring instrumentation 77a 1
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4.2 1 Minimum Test and Calibration Frequency for FCIS 78 l
l 4.2 2 Minimum Test and Calibration Frequency for Core and Containment 79 Cooling System i
4.2-3
_ Minimum Test and Calibration Frequency for Control Rod Blocks 8t Actuation 4.2-4 (DELETED) 82 4.2 5 Minimum Test and Calibratian Frequency for Drywell Leak Detection 83 4.2 6.
(DELETED)
L 4.2 7 Minimum Test and Calibration Frequency for Recirculation Pump Trip 85 Amendment No. 20,93,/JO v
4 -
JAFNPP UST OF TABLES (Cont'd)
Table Title pg l
4.2-8 Minimum Test and Calibration Frequency for Accident Monitoring 86 Instrumentation 4.61 Comparison of the James A. FitzPatrick Nuclear Power Plant inservice 157 Inspection Program to ASME Inservice inspection Code Requirements 4.6-2 Minimum Test and Calibration Frequency Ior Drywell Continuous 162a Atmosphere Radioactivity Monitoring System 3.7 1 Primary Containment isolation Valves 198 4.7 1 (DELETED) 210 l
4.7 2 Exception to Type C Tests 211 3.12 1 Water Spray / Sprinkler Protected Areas 244j 3.12 2 Carbon Dioxide Protected Areas 244k 3.12 3 Manual Fire Hose Stations 2441 4,12 1 Water Spray / Sprinkler System Tests 244q 4.12 2 Carbon Dioxide System Tests 244r 4.12 3 Manual Fire Hose Station Tests 244s 6.21 Minimum Shift Manning Requirements 260a 6.10-1 Component Cyclic or Transient Umits 261 1
Amendment No. So, d,#2, Z,130,184,180 I
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...r JAFNPP 32 (cont'd) 42 (cont'd) -
E.
Drywell Leak Detection E.
Drywell Leak DetectKm The limiting conditions of operation for the instrumentation that instrumentation shall be calibrated and checked as irdcated in monitors drywell leak detection are given in Table 32-5.
Table 42-5 F.
(Deleted) l F.
(Deleted)
G.
Recirculation PumpTrip G.
Recirculation PumpTrip The limiting conditions for operation for the instrumentation that instrumentation shall be functionally tested and calibrated as trip (s) the recirculation pumps as a means of limiting the indicated in Table 42-7.
consequences of a failure to scram during an anticipated System logic shall be functionally tested as irdcated in Table transient are given in Table 32-7.
42-7.
H.
Accident Monitoring instrumentation H.
Accident Monitoring instrumentabon The limiting conditions for operation of the instrumentation that
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instrumentation must be operable whenever the reactor is entical 42-8~
and reactor coolant temperature is greater than or equal to 2127.
1.
4kv Emergency Bus UndervoltageTrip The limiting conditions for operation for the instrumentation that prevents damage to electrical equipment or circuits as a result of either a degraded or loss-of-voltage condition on the emergency electrical buses are given in Table 32-2.
Amendment No.1)l16,1;ilo 54
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JAFNPP-32 BASES (cont'd) the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service.
Ficw integrators are used to record the integrated flow of liould '
from the drywell sumps. The leak rate is calculated by dividing the integrated volume pumped out of the sumps by the time between sump pump operabons. The resultant leak rate.veJue, wtuch is expressed in gallons per rninute, is ce rged to the acceptance criterion specified in Specificahon 3.6.D.
For each parameter monitored, as listed in Table 32-8, by I
comparing the reading of each chiird to the reading on redundant or reisted instrument ctwa 4 a near continuous surveillance of instrument performance is available.
Amendment No. 96,M };16 59
.o JAFNPP
/'
32 BASES (cont'd)
The recirculation pump trip has been added at the suggestion The Emergency Bus Undervoltage Trip System transfers the 4 of ACRS as a means of limiting the consequences of the kv emergency electrical buses.to the Emergency Diesel unlikely occurrence of a failure to scram during an anticipated Generators in the event an undervoltage condition is detected.
transient. The response of the plant to this postulated event The system has two levels of protectiort (1) degraded voltage falls within the envelope of study events given in General protection, and (2) loss-of-voltage protection. ' Degraded Electric Company Topical Report, NEDO-10349, dated voltage protection prevents a sustained low voitage condition March,1971.
from damaging safety-related equipment. The degraded Accident monitoring instrumentation provides additional v Itage protection has two time delays. A short time delay information which is helpful to the operator in assessing plant coincident with a loss-of-coolant accident (LOCA) and a longer time delay to - allow normal plant evolutions without conditions following an accident by (1) providing information needed to permit the operators to take preplanned manual mcessarHy starthg h Emgency M hatas. De actions to accuuplish safe plant shutdown; (2) determining loss-of-voltage protection prevents a more severe voltage drop whether systems are performing.their intended functions; fran causing a long term interruption of power. Time delays.
(3) providing information to the operators that will enable them are included in the system to prevent s,nadvertent transfers due to determine the potential for a breach of the barrier to to spurious voltage decreases. Therefore, both the duration radioactivity release and if a barrier has been breached; and severity of the voltage drop are sensed by the Emergency (4) fumishing data for deciding'on the need to take unplanned Bus Undervoltage Trip System.
action if an automatic or manually initiated safety system is not functioning properly or the plant is not responding properly to the safety systems in operation; and (5) allowing for early indication of the need to initiate action necessary to protect the public and for an estimate of the magnitude of any problem.
This instrumentation conforms with the acceptance criteria of NUREG-0737, NUREG-0578, and NRC Generic Letter 83-36 and includes Regulatory Guide 1.97, Revision 2 Type A variables.
Amendment No.18tl, liiO,96, M!Id 60
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42 BASES
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l The instrurnentation listed in Tables 42-1 through 42-8 will be To test the trip relays requires that the channel be bypassed, functionally tested and calibrated at regularly scheduled the test made, and the systern returned to its initial state. It is intervals. The same design reliability goal as the Reactor assumed this task requires an estimated 30 rnin. to ccTylete in -
Protection System is generally applied. Sensors, trip devices a thorough and workmanlike manner and that the relays have a 4
and power supplies are tested, calibrated and checked at the failure rate of 10 failures per hr. Using this data and the above same frequency as comparable devices in the ' Reactor operation, the optimum test interval is:
Protection System.
Those instruments whict, when tripped, result in a rod block i=
M =1x10 hr.
3 4
I have their contacts arranged in a 1 out of n logic, and all are 10 capable of being bypassed. For such a tripping arrangement
= 40 days 1
with bypass capability provx$ed, there is an optimum test For additional margin a test interval of once/ month will be interval that shoc!d be maintained in order to maximize the used initially.
reliability of a given channel (7). This takes account of the fact l
that ' testing degrades reliability and the optimum interval The sensors and Monic apparatus W not been included t
l between tests is approximately given by:
We as these are ana%
wm readoun in the control room and the sensors and
,_~uur c apparatus can be checked by csTyarison with other like instruments. The i=
checks which are made on a daily basis are adequate to assure l
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operability of the sensors and electronic apparatus, and the test interva! given above provides for optimum testing of the
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the optimuminterval between tests.
relay circuits.
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the time the trip contacts are disabled from i
performing their function while _the test is in progress.
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the expected failure rate of the relays.
I Amendment No. D6,1#i 61 l
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JAFNPP TABLE 3.2-8 ACCIDENT MONITORING INSTRUMENTATM)N No.of Cterveis E% Turn No.of Prcmded by Operable Ctsi,Gs instnsTsit Design Required Action 2
1 8
1.
Stack High Range Effluent Monitor 4
(17RM-53A)
(17RM-538) 2.
Turbine Building Vent High Range Emuent Monitor 2
1 B
(17RM-4348) 3.
Radwaste Building Vent High Range Effluent Monitor 2
1 B
(17RM-4638) h 4.
Containment High Range Radiation Monitor 2
1 A
(27RM-1048) 5.
Drywell Pressure (narrow range) 2 1
A (27PI-115A1 or 27PR-115A1)
(27PI-11581 or 27PR-11581) 6.
Drywell Pressure (wide range) 2 1
A (27P' 115A2 or 27PR-115A2)
(27PI-11582 or 27PR-11582) 2 1
A 7.
DrywellTemperature (16-1TR-107)
- At less than or equal to 450 R/hr, closes vent and purge valves Amendment No.186 77a EE
JAFNPP TABLE 32-8 (cont'd)
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ACCIDENT MONITORING INSTRUMENTATION No. cf Chim a 6s Mirwnum No.of Provded by Operable Channels Ins +rument Design Rarmhed Achon 8.
TorusWater Level 2
1 A
(23U-202B or 23LR-202B/2038) 9.
- Torus Water Temperature 2
1 A
(16-1TI-131A or 16-1TR-131A)
(16-1TI-1318 or 16-1TR-1318) 10.
Torus Pressure 2
1 A
i (27PR-101A)
(27PR-10181) 11.
Drywell Hydrogen / Oxygen Concentration 2
1 F
(27PCR-101B) 12.
Reactor Vessel Pressure 2
1 A
ReactorWater Level (fuelzone) 2 1
A (02-3U-091)
(02-3LR-098) 14.
Reactor Water Level (wide range) 2 1
A (02-3U-85A)
Amendment No.Jed 77b
JAFNPP a
TABLE 32-8 (cont'd)
ACCIDENT MONITORING INSTRUMENTATION No.of Channels Mirumum No. of Provided by Operable Channels Instrument Design Required Action 15.
Core Spray Flow 1 perloop 1 perloop A
loop A (14F1-50A) loop B (14FI-508) 16.
Core Spray discharge pressure 1 perloop 1 perloop A
loop A (14PI-48A) loop B (14PI-488) 17.
LPCI(RHR) Flow 2 perloop 1 perloop A
loop A (10FI-133A)
(10FR-143 - red pen) loop B (10FI-1338)
(10FR-143 - black pen) 18.
RHR Sendce Water Row 1 perloop 1 per loop A
loop A (10FI-132A) loop B (10F1-1328) i 19.
Safety / Relief Va!ve Position Indicator 2
1 D, E (See Note C) 20.
Torus Water Level (Narrow Range) 2 1
B (27U-201 A or 27R-101 - red pen or EPIC A-1258)
(EPIC A-1260)
(See Note G) 21.
Drywell-Torus Differential Fressure 2
1 8
(16-1DPR-200 or EPIC A-3554)
(EPIC A-3551)
(See Note G)
Amendment No.
77c
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TABE 324 (Cont'd)
ACCIDENT MONITORING INSTRUMENTATM)N NOTES FOR TABE 32-8 l
A.
With the number of operable channels less than the required minimum, either restore the inoperable channels to operable status within 30 days l
or be in a cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
With the number of OPERABE den es less than required by tt e rrwnemum channels OPERABE requirements, initiate an attemate nwtics of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and: (1) eether restore the inoperable de a-6(s) to OPERABW status within 7 days of the event, or (2) prepare and submit a Special Report to the CGT.T45sion within 14 days following the event outlining the cause of the inoperability, the action taken, and the plans and schedule for restoring tbc system to OPERABE status.
C.
Each Safety Relief Valve is equipped with two acoustica! detectors, one of which is in sennce. Each SRV also has a backup thermocouple detector. In the event that a thermocouple is inoperable SRV performance shall be monitored daily with the associated is-vice acoustical detector.
D.
From and after the date that both of the acoustical detectors are ir@abie, continued operation is permisseble unt;l the next outage in which a primary containment entry is made provided that the thermocouple is operable. Both acoustical detectors shall be made operable prior to restart.
E.
In the event that both pnmary (acoushcal detectors) and secu day (mermocouple) indicatbns of this parameter for any one valve are disabled and nerther irxhcahon can be restored in forty-eight (48) hours, an orderly shutdown shall be initiated and the reactor shall be in a Hot Shutdown condition in twelve (12) hours and in a Cold Shutdown within 1;m next twenty-four (24) hours.
j F.
Refer to Specification 3.7.A.9.
G.
This parameter and associated instrumentation are not part of post-accident monitoring.
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Amendment No.
77d
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TABLE 42-1
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MINIMUM TEST AND CALIBRATION FREQUENCY FOR PCIS Instrument Channel (8)
Ins *rument Functional Test Calibreuon Freq=scy Instn,Ts4 Check (4)
- 1) Reactor High Pressure (1)
Once/3 months Ncne (Shutdown Cooling Permissive)
- 2) Reactor Low-Low-Low Water Level (1)(5)
(15)
Orce/ day
- 3) Main Steam High Temp.
(1)(5)
(15)
Once/ day
- 4) Main Steam High Flow (1)(5)
(15)
Once/ day
- 5) Main Steam Low Pressure (t)(5)
(15)
Once/ day
- 6) Reactor Water Cleanup High Temp.
(1)
Or.ce/3 months None
- 7) Corsmer LowVacuum (1)(5)
(15) -
Once/ day Loge System Functional Test (7) (9)
Frequercy 1)
Main Steam Une Isolation valves Once/6 months Main Steam Une Drain Valves Reactor Water Sample Valves 2)
RHR -Isolation Valve Control Once/6 months Shutdown Cooling Valves Head Spray 3)
Reactor Water CleanupIsolation Once/6 months 4)
Drywellisolation Valves Once/6 months TIP Withdrawal Atmospheric ControlValves 5)
Standby Gas Treatment System Once/6 nxmths Reactor Building isolation l
NOTE-See notes following Table 42-5.
Amendment No./,M 1116 78
M JAFNPP TABLE 42-2 MINIMUM TEST AND CAUBRATION FREQUENCY FOR CORE AND CONTAINMENT COOUNG SYSTEMS Instrument Channel Instrument Functional Test Calibration Frequency instrument Check (4) 1)
Reactor Water Level (1)(5)
(15)
Once/ day j
2a)
DryweII Pressure (non-ATTS)
(1)
Once/3 months None l
2b)
Drywell Pressure (ATTS)
(1)(5)
(15)
Once/ day 3a)
Reactor Pressure (non-ATTS)
(1)
Once/3 months None j
3b)
Reactor Pressure (ATTS)
(1)(5)
(15)
Once/ day l
4)
Auto Sequencing Timers None Once/ operating cycle None 5)
ADS-LPCI or CS Pump Disch.
(1)
Once/3 months None 1
6)
Trip System Bus Power Monitors (1)
None None 8)
Core Spray Sparger d/p (1)
Once/3 months Once/ day 9)
Steam Line High Flow (HPCI & RCIC)
(1)(5)
(15)
Once/ day 10)
Steam Une/ Area High Temp. (HPCI & RCIC)
(1)(5)
(15)
Once/ day 12)
HPCI & RCIC Steam Line Low Pressure (1)(5)
(15)
Once/ day 13)
HPCI & RCIC Suction Source Levels (1)
Once/3 months None 14) 4kV Ensgency Bus Under-Voltage Once/@ating cycle Once/ operating cycle None (Loss-of-Voltage, Degraded Voltage LOCA and non-LOCA) Relays and Timers.
l 15)
HPCI& RCIC Exhaust Diaphragm (1)
Once/3 months None Pressure High 17)
LPC / Cross Connect Valve Posrtien Once/ operating cycle None None NOTE-See notes following Table 42-5.
Amendment No. M,#58'p5,1)l6,1/,1Jid 79
9 JAFNPP i
TABLE 42-2 (Cont'd)
MINIMUM TEST AND CAUBRATION FREQUENCY FOR CORE AND CONTAINMENT COOUNG SYSTEMS i
Logic System Functional Test Frequency l
I 1)
Core Spray Subsystem (7) (9)
Once/6 months 2)
Low Pressure Coolant injection Subsystem (7) (9)
Once/6 months i
3)
Containment Cooling Subsystem (9)
Once/6 months 4)
HPCI Subsystem
- 9) (9)
Once/6 months 5)
HPCI Subsystem Auto isolation F) (9)
Once/6 months 6)
ADS Subsystem p) (9)
Once/6 rixmths 7)
RCIC Subsystem Auto isolation (7) (9)
Once/6 tredis 8)
ADS Relief Valve Bellow Pressure Switch
- 9) (9).
Once/ operating cycle l
NOTE-See notes following Table 42-5.
Amendment No. dis 80
JAFNPP t
TABLE 42-3 MINIMUM TEST AND CALIBRATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATION SYSTEMS
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Instrument Functional instrument Instrument Channel Test (5)
Calibration Check (4) l 1)
APRM-Downscale (1)
Once/3 rrudis -
Once/ day 2)
APRM-Upscale (1)
Once/3 months Once/ day 3) lRM-Upscale (2)
(3)
(6)
Once/ day 4)
IRM - Downscale (2)
(3)
(o7 Once/ day 5)
RBM - Upscale (1)
Once/3 months Once/ day 6)
RBM-Downscale (1)
Once/3 rrwr6 Once/ day 7)
SRM - Upscale (2)
(3)
(6)
Once/ Jay 8)
SRM - Detector Not in Startup Position (2)
(3)
(6)
None l
9)
IRM - Detector Not in Startup Position (2)
(3)
(6)
None 10)
Scram Discharge Instrument Volume-Once/ month Once/3 rrwds Once/ day High Water Level (Group B instmments)
(1) i Logic System Function Test (7) (9)
Frequency d
1)
System Logic Check Once/6 months
]
NOTE-See notes following Table 42-5.
Amendment No. 492, S4 81
JAFNPP TABLE 42-5 MINIMUM TEST AND CAUBRATION FREQUENCY FOR DRYWELL LEAK DETECTION Instrument Functiona:
Calibration Instrument Check Instrument Channel
. Test Frequency (4) 1)
Equipment Drain Sump Flow Integrator (1)
Once/3 months Once/ day 2)
Floor Drain Sump Flow Integrator (1)
Once/3 months Once/ day l
NOTE:
See notes following Table 42-5.
b F
Amendment No. Off,Ji9 83
d JAFNPP l
NOTES FOR TABLES 42-1 THROUGH 42-5 7.
aW am* aMion M M pe Ti TM once e 1.
Initially once every month until acceptance failure rate data are paw cy& h @, aR @ system WW.
available; thereafter, a request may be made to the NRC to tests W be perfW dng h testjads.
change the test frequency. The compilation of instrument failure rate data may include data obtamed from other boilirg 8.
Reactor low water level, high drywell pressure and high water reactors for which the same design instruments operate radiation main steam line tunnel are not inc'uded on Table 42-1 in a environment similar to that of JAFNPP.
since they are tested on Table 4.1-2.
2.
Functional tests are not required when these instruments are 9.
The logic system functional tests shall include a calibration of not required to be operable or are tnpped. Functional tests time delay relays and timers necessary for proper functioneg shall be performed within seven (7) daye orior to each startup.
of the trip systems.
3.
Calibrations are not required when these instruments are not 10.
At least one (1) Main Stack Difution Fan is required to be in required to be operable or are tripped. Calibration tests shall operation in order to isokirietically sample the Main Stack.
be performed withen seven (7) days prior to each startup or 11 Uses same instrumentation as Main Steam Line High prior to a pre-planned shutdown.
Radiation. SeeTable4.1-2.
4.
Instrument checks are not required when these instruments are 12.
not required to be operable or are tripped.
Calibration and instrument check sunreillance for SRM and IRM 13.
l instrumentation.s exempt from the furn,..,o. test
!nstruments are as specif;ed in Tables 4.1-1,4.1-2,42-3.
5.
This.
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definition. The functional test will consist of inject,ng a i
14.
FtmWial test is performed once each operating cycle.
simulated electrical signal into the measurement cimrird.
6.
These instrument channels will be calibrrzt using simulated 15.
Sensor calibration once per operating cycle. Master /sfave trip electrical sgnals once every three months.
unit calibration once per 6 months.
Amendment No. ;i(ptf,5(,p
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JAFNPP TABLE 4.2-8 MINIMUM TEST AND CALtBRATION FREQUENCY FOR ACCIDENT MONITORING INSTRUMENTATION InstnATe,t instrtnTeent FuncbonalTest Calibration Frequency Check instrtnTient 1.
Stack High Range Emuent Monitor Once/ Operating Cycle Once/GrwaGrg Cycle Once/ day 2.
Turbine Building Vent High Range Emuent Monitor Once/ Operating Cycle Once/Operatb.g Cycle Once/ day 3.
Radwaste Building Vent High Range Emuent Monitor Once/ Ope'a'ing Cycle Once/ Operating Cycle Once/ day 4.
Containment High Range Radiation Monitor Once/ Operating Cycle Once/Op AgCycle Once/ day 5.
Drywell Pressure (narrow range)
N/A Once/ Operating Cycle Once/ day 6.
Drywell Pressure (wide range)
N/A Once/CriaGrg Cycle Once/ day 7.
DrywellTemperature N/A Once/Op=abig Cycle Onca/ day 8.
Torus Water Level N/A Once/ Operating Cycle Once/ day 9.
Torus Water Temperature N/A Once/ Operating Cycle Once/ day 10.
Tons Pressure N/A Once/ Operating Cycle Once/ day 11 Drywell Hydrogen / Oxygen Concentration Analyzer N/A Once/3 rronths Once/ day 12.
ReactorVessel Pressure N/A Once/ Operating Cycle Once/ day 13.
ReactorWater Level (fuelzone)
N/A Once/OpaaGrg Cycle Once/ day 14.
Reactor Water Level (wide range)
N/A Once/Or-ating Cycle Once/ day Amend 6 c71 No. %
r JAFNPP
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TABLE 4.2-8 (cont'd)
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MINIMUM TEST AND CAllBRATION FREQUENCY FOR ACCIDENT MON;TORING INSTRUMENTATION Instrument instrument Instrument FunctionalTest Calibration Frequency Check 15.
Core Spray Flow N/A Once/ Operating Cycle Once/ day 16.
Core Spray Descharge Pressure N/A Onc3/ Operating Cycio Once/ day 17.
LPCI (RHR) Row N/A Chce/ Operating Cycle Once/ day 18.
RHR Service Water Flow N/A Once/ Operating Cycle Once/ day 19.
- Safety / Relief Valve Position Indicator Once/ Operating Cycle N/A Once/ month (Primary and Secondary) 20.
Tems Water Level (narrow range)
N/A Once/ Operating Cycle Once/ day 21.
Drywell-Torus Differential Pressure N/A Once/ Operating Cycle Once/ day l
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33.A.2 (cont'd) 4.3.A.2 (cont'd) b.
The control rod directional control vaives for e.
The scram dscharge volume drain and vent valves inoperable control rods shall be disarmed shall be full-travd cycled at least once per quarter to electncally.
venfy that the valves close in less than 30 seconds c.
Control rods with scram tirnes greater than those W to me pp vh Me W pah.
permitted by Specificabon 3.3.C.3 ara enoperable, f.
An instrument check of control rod position but if they can be inserted with coatrol rod drive indicabon shall be fmdvin=d once/ day.
pressurts they need not be disarmed electncally.
d.
Control rods with inoperable accumulators or those whose position cannot be posdively detsnnmd shall be considered inoperable.
e.
Inoperable control rods shai; be positioned such that Specification 3.3.A.1 is met. In addition, during reactor power operation, no more than one control rod in any 5 X 5 array may be inoperable (at least 4 operable control rods must separate any 2 inoperable ones). If this specifrJion cannot be met the reactor shall not be started, or if at power, the reactor shall be brought to a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 Amendment No. $,9, }E2,JAl6 90
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3.7 UMITING CONDITIONS FOR OPERATION 4.7 SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability-Apprecability-Applies to the operating status of the pnmary and secondary Applies to the pomary and secondary cont & rient intagnty.
containment systems.
Objective:
Objectrve:
To assure the integnty of the pnmary and seux.dery contanment To venfy the integnty of the pnmary, and secridary contanment systems.
systems.
Specification:
Specification-A.
Primary Cor'tWarsit 1.
The volume and temperature of the water in the torus shall 1.
The forus water level and temperature shall be monitored at all tunes, except as specified in Specification 3.5.F.2 be as specified in Table 42-8. The accessible interior maintained within the followmg limits:
surfaces of the drywell and above the water line of the torus shall be inspected at each re8ueleng outage for l
a.
Maximum vent subn serga (,e level of 53 inches.
evidence of deteriorabon. Whenever there is indicarbn of b.
Minimum vent sutariergence level of 51.5 inches.
p or to N suppression pool, the pool temperature shall be I
The torus water level may be outside the above continually monitored and also observed and logged every limits for a maximum of four (4) hours dunng 5 minutes until the heat addition is terminated. Whenever required operability testing of HPCI, RCIC, RHR, CS, there is indication of relief valve operation v.ith the l
and the Drywell-Torus Vacuum System.
temperature of the suppression pool reaching 1607 or c.
Maximum water temperature more and the pnmary coolant system pressure greater than 200 psig, an extemal visual exammation of the torus l
(1)
During normal power opcfation max' um M be WNwe rhng powe opeadon.
m water temoerature shall be 95'F.
Amendment No.g l
165~
.4.e JAFNPP i
l 3.7 (cont'd) 4.7 (cont'd) l l
6.
Oxygen Concentration 6.
Oxygen Concentration i
j l
The pnmary containment ohncipiere shall be a.
The primary contairrnant oxygen ocerox?tation shall a.
reduced 'to less than four percent oxygen with be mortored as speciMed in Table 42-6.
l nitrogen gas during reactor power operation wth l
reactor coolant pressure above 100 psig, except as l
specified in 3.7.A.6.b.
l b.
Within the 24 hr. penod stbsequent to placing the reactor in the rur. mode fo!!owirg a shutdown, the l
contamment atrnosphere oxygen corr.cntration shall be reduced to less than 4 percent by weight and maintained in this condition. De-inerting may commence 24 hr. prior to a shutdown.
7.
Dryweft-Torus Differential Pressure
~l 7.
Dryweli-Torus Differenbal Pressure l
a.
The pressure differential between tre dryweil ard a.
Differential pres *:ure between the drywell and torus torus shall be monitored as specified in Tat $e 426.
l shall be maintained at ecaal to or greater than 1.7 psid except as specified in (1) and (2) below:
1 Amendment No. /
t80
r JAFNPP 3.7 (corfd) 4.7 (confd) 9.
Primary cont-auvsit atmosphere shall be continuously 9.
Primary Contamment Atmospiwe N senng Instruments reonitored for hydrogen and oxygen when contidursit integdty is required. The excepton to this is when the a.
Instrumentrtim shall be b'Mwally tested and Post-Accident Samprog System is to be operated. In this calibrated as speediedin Table 424.
l instance, the contairrnent Mmospiiaie monitoring systems may be isolated for a penod not to exceed 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> in a 2+ hour penod.
The monitoring system shall be considered operable if at least one monitor is operable.
a)
From and after the time the pnmary contamment ahTicspise moriitoring i.h are found or B.
Standby GasTreatment System made to be inoperable for any reason, conhnued reactor operation is punvisssib'e for the succeeding 1.
Standby Gas Treatment System sunellance shall be thirty (30) days unless one mstrument nxmitoring performed asindicated below:
each parameter is sooner made operable, provided an appropriate grab sample is obtained and a.
At least once per operating cycle, it shall be analyzed at least once each tweref-four (24) hour demeiwirated that-period.
b)
If specificabon 3.7.A.9.a cannot be met, the reactor 0) me & p moss N 91W @
shall be placed in the cold condibon withm twenty-efficency and charcoal filters as less than 5.7 four (24) hours.
- m. of water at 6,000 scfm, and (2)
Each 39kW heater shall dissipate greater than B.
Standby GasTreatment System 29kW of electric power as calculated by the followog expressen:
1.
Except as specified in 3.7.82 below both circuits of the P =6EI Standby Gas Treatment System shall be operable at all g,~.
times when secondary contamment integnty is required.
P= Dessipated Electncal Power; E= Measured line-to-line voltage in volts (RMS);
!= Average measured phase current in amperes (RMS).
Amerximent No. )6,28, si6, GEi,99,81, S8 181
~
l JAFNPP 3.7 BASES { cont'd) l Using the rFnirrr.m or maximum downcomer submergence Usmg a 407 rise (Section 52 FSAR) in the suppressKm l
levels given a the specification, contamment pressure during ciaviber water temperature and a maximum erwtial temperature l
the design bases acodont is approximately 45 psig which is of 957, a temperature of 145P is actweved, wtuch is well below below the design of 56 psig. The r:wrumurn downcomer the 1707 temperature wtuch is used for ccTpC sutxnergence of 51.5 in. results in a,drr:inimum suppressson condensation.
chamber water volume of 105,600 ft. The majonty of me j
g,
g l
Bodega tests (9) were run with a vsbmerged length of 4 fL and with complete conv,a.abon. Thus, with respect to downceTia-temperature of 957 and assuming the normal ccTip ernen. of I
s submergence, this specificahon is adequate. Additional contamment cooing pumps (two LPCI pumps and two RHR service water p'mnps) contamment pressure is r ot required to
(
fw the aanment nt Pr am e
the a@w y of the spa.;T,ed range of submergence to ensure cwe M, W, W Hm punps.
that dynamic forces associated with pool swell do not result in Limiting suppressson pool temperature to 1307 dunng RCIC, overstress of the suppressoon cf.sstes-or associated HPCI, or relief valve operabon, when decay heet and stored structures. Leve4 insm> mentation is provided for operator use energy are removed from the pnmary system Dy discharging 1.
to maintain downcomer submergence withm the spa %,5 reactor steam directly to the suppression chamber assures range.
adequate margm for a potential blowdown any time dunng The maximum temperature at the end of blowdown tested
,H
,w v
operabon.
dunng the Humboldt Bay (10) and Bodega Bay tests was Expenmental data indicales that excesseve steem conda=ing 1707, and this is conservatively taken to be the limit for loads can be avoeded if the peak temperature of the curipiete cu.dimation of the limit for cuYW ^e condensabon suppressxm pool is mantamed below 1607 dunng any penod of the reactor coolant, although cciidarWuen would occur for of relief valve operation with sonic conditions at the discharge t
temperatures above 1707.
exit. Specircsiions have been placed on the envelope of Should it be necessry to drain the suppressxm ci.aiTiber, this reactw pa@ who so M N reactw can be en a % y_to W the ryme of should only be done when there is no requirement for
" F' pwn.=n suppr e i,ia m W ngs.
Frnergency Core Cooling Systems operability as explained in basis 3.5.F.
Amendment No. 95,3tf 188
l l
i i
j JAFNPP l
3.7 BASES (cont'o; are scheduled during startup penods, when the pnrnary system bypass leakage is detwin nod to be that which would limit the j
is at or near rated operating temperature and pressure. The 24 maximum contamment pressure rise to the design value of 56 hour6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> penod to pronoe inertog is judged to be sufficien; to psig and is a funcbon of the break size, reactor pressure decay, pasivin. the leak inspacAion and estabissh the required oxygen and time till contanment sorays are actuated. The allowable j
c.eian^uation.
test leakage,71 scfm, is approximately 10 twnes less than the l
The primary contanment is normally slightly pressunzed during maxiruxn allowable bypass Wy. A test leakage d 71 penods of reactor operabon. Nitrogen used for inviting could cmWw to a prm trm d 2 m. water /&
leak out of the containment but air could not leak in to increase e a 10 e W @ h W/tm @ to 1 $
oxygen concentration.
A dryv, ell-torus mmemum differenbal pressure of 1.7 psid has Drywellto TorusVacuum Breakers estm as W @e to me M wWe 4
torus and torus support system safety margins are maintaired l
The capacity of the five drywell to torus vacuum relief valves followmg postulated design bases acc.kh& This differential are sized to limit the pressure differenbal between the torus and lowers the water level in the torus to drywell vert dowricinais l
drywell during post-accident drywell cooling operations to well thereby reducmg dynamic forces as a result of a LOCA.
te.ar the dessgn limit of 2 psi. They are sized on the bases of Instrumentation is pronded for operator use to memtan 09 Bodega Bay pressure suppression system test. The ASME drywell-torus differential pressure.
Dailer and Pressure Vessel Code, Sechon ill, Subsection B, for B.
Standby Gas Treatment System and tnis vessel Gows a 2 psi differential. With one vacuum relief valve secured in the closed position and four operable valves, C.
Secondary Contamment contasiinent integnty is not impaired.
The sec.eny contamment is designed to minimize any l
The drywell to torus bypass leakage is limited to 71 scfm to ground level release of radioachve matenale, which might result provide assurance that steam released to the drywell will flow from a senous acc.kW.t. The reactor building provides g
to and be c,eidosioed in the torus. The maximum allowable Amendment No. 36 4
190
N e
9 JAFNPP O
Table 4.7-1 THIS PAGEIS INTENTIONALLY BLANK i
I 1
Amerninent No.
210
- =
= --.r-
i ATTACHMENT 11
^
SAFETY EVALUATION FOR PROPOSED TECHNT5KEEPECTFICATI5iT5KANGES REGARDING ACCIDENT MONITORING INSTRUMENTATION JPTS-8H15 l
l I
l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l'
l l
{
A 1 SAFETY EVALUATION Page 1 of 9 1.
DESCRIPTION OF THE PROPCSED CHANGES The proposed change to the James A. FitzPatrick Technical Specifications revises pages v, vi, 54, 50, 80, 61, 76, 77a, 77 b, 78, 79, 80, 81, 83, 84, 85, 86, 86a, 90,165,180,181,188, 190, and 210. Pages 76a through 76d are removed from the specificationt, and two new page 77c and 77d are created. The changes are detailed below.
Page v, Ust of Tables Revise the Ust of Tables to show that Tables 3.2 6 and 4.2-6 are deleted end to show that Table 4.2 7 is on page 85.
Correct the
- Instrumentation that initiates Control Rod Blocks' entry to refer to Table 3.2 3 instead of Table 3.2-2 to extract a typographical error.
Page vi, Ust of Tables (Cont'd)
Revise to show that Teole 4.2 8 is on page 86 and to show that Table 4.*' 1 is deleted.
Page 54, Specifications 3.2.F and 4.2.F These two specification are deleted.
Specification 3.2.H Add a new sentroce to this specificaton which reads, "This int.trumentation must be operable whenever the reactor is critical and reactor coolant temperature is greater than or equal to 212"F.*
I Page 59, Basos 3.2 In the last paragraph on this page, replace
- Table 3.2-6' with
- Table 3.2-8."
Delete the last sentence in its entirety. This sentence currently reads, "Any deviation in readings will initiate any early recalibration thereby maintaining the quality of the instrument readings.*
Page 60, Bases 3.2 Replace the last sentence in the second paragraph with, "This instrumentation conforms with the acceptance criteria of NUREG 0737, NUREG-0578, and NRC Generic Letter 83 36 and includes Regulatory Guide 1.97, Revision 2 Type A variables."
@Qe 61, Bases 3.2 In the first paragraph on this page, replace " Table 4.2-1 through 4.2-6" with
- Tables 4.21 tSrough 4.2-8."
o I
-e
- - - ~
w
~-
n
^ 1 SAFETY EVALUATION l
Page 2 of 9 Pages 76 through 76d, Table 3.2 6 Delete this table in its entirety, insert "This page is intentionally blank" on page 76 evi remove pages 76a through 76d from the Technical Specifications.
Page 77a, Table 3.2-8 Replace the existing table with the revised table included in Attachment 1. This table now requires three pages and is continued on pages 77b and 77c.
Page 77b, Notes for Table 3.2-8 Renumber thh page '77d." Relocate Notes 10,11, and 12 from Table 3.2-6 to this page and relable them C, D, and E respectively.
Within Note A, delete *: (1) Initiate an alternate method of monitoring the appropriate parameter (s), or (2).*
Within Note B, replace *the alternate method" with *an alternate method."
Within Note C, replace " detectors of which one is in service and a' with " detectors, one of which is in service. Each SRV also has a." Also insert "in service" after " associated."
Within Note D, replace 'none" with *both,' *is' with *are,' and
- with
- inoperable." Doiete "but the thermocouple is operable," and insert *provided that the thermocouple is operable" at the end of the first sentence.
Within Note E, insert *(acoustical detector)" after " primary" and insert *(thermocouple)"
after " secondary.'
Add a new Note F to read, ' Refer to Specification 3.7.A.9.*
Add a new Note G to read, "This parameter and associated instrumentation are not part of post accident monitorirg."
Pages 78,79,80,81, and 83, Tables 4.21 through 4.2 5 Replace the note on the bottom of these five tables with, "See notes following Table 4.2 5."
Page 81, Table 4.2 3 Remove the reference to note 12 from the " Instrument Check" heading.
P_ age 84, Table 4.2-6 Delete this table in its entirety.
Page 85, Notes for Tables 4.21 through 4.2 6 Rename this list of notes " Notes for Tables 4.21 through 4.2 5' and relocate to page 84.
In Note 5, replace ' excepted' with ' exempt
- to correct the grammatical error.
Revise to show that Note 12 is deleted.
^' 1 SAFETY EVALUATON Page 3 of 9 Page 86, Table 4.2 7 Replace, 'Once/ refueling cycle
- with *0nce/ operating cycle" for the logic system functional test frequency to achelve consistency in nomenclature. Relocate this table to page 85.
Pac,e 868, Table 4.2-8 Replace the existing table with the revised table included in Attachment I. Relocate this table to page 86. This table now requires two pages and is continued on page 86a.
Page 90, Speelfication 4.3.A.2.f Insert a new Specification 4.3.A.2.f to read, 'An instrument check of control rod position indication shall be performed once/ day,"
Page 165, Specifications 3.7.A.1 and 4.7.A.1 In five places, replace ' pressure suppression chamber" and " suppression chamber" with
- torus."
Replace the first sentence of Specification 4.7.A.1 with, "The torus water level and temperature shall be monitcred as specified in Table 4.2 8.*
Page 180, Specification 3.7.A.6.a Delete "After completion of the startup test program and demonstration of plant electrical output."
Specification 4.7.A.6.a Replace ' measured and recorded at least twice weekly" with " monitored as specified in Table 4.2-8.*
Specification 4.7.A.7 Replace 'Drywell Suppression Chamber
- with "Drywell Torus.'
Soecification 4.7.A.7.a Replace " suppression chamber shall be recorded at least once per shift" with " torus shall be monitored as specified in Table 4.2-8,"
Page 181, Specification 4.7.
Replace " Table 4.71' with
- Table 4.2-8.*
Page 100, Bases for 3.7.A At the end of the first pagraph, insert a new sentence to read, ' Level instrumentation is provided for operator use to maintain downcomer submergence within the specified range."
s>
, 1 SAFETY EVALUATION Page 4 of 9 Page 190, Bases 1or 3.7.A At the end of the fifth pagraph, insert a new sentence to read,
- instrumentation is provided for operator use to rnaintain drywell suppression chamber differential pressure."
in eight places, replace " pressure suppression chamber
- with
- torus."
Page 210, Table 4.71 Delete this table in its entirety. Insert "This paQe is intentionally blank."
ll.
PURPOSE OF THE PROPOSED CHANGES NRC Regulatory Guide 1.97, Revision 2 defines Type A variables as plant uniques variables
... that provide primary information needed to permit the control room operating personnel to take the specified manually controlled actions for which no automatic controi is provided and that are required for safety systems to accomplish their safety functions for design basis accident events."
The Authority determined which plant specific parametets should be classified as Type A variables and upgraded plant instrumentation systems such that these parameters are monitored with Category 1 instrumentation (References 3 and 7).
The Technical Specifications are being revised to include these instruments in Table 3.2-8,
' Accident Monitoring Instrumentation." The associated surveillance requirements for these Instruments are added to Table 4.2 8,
- Minimum Test and Calibration Frequency for Accident Monitoring Instrumentation." Table 3.2-8 is also being restructured to conform more closely to the corresponding Table 3.3.7.51 in the Standard Technical Specifications (STS) with regard to format and content. Tables 3.2-6 (" Surveillance Instrumentation") and 4.2 6 (* Minimum Test and Calibration Frequency for Surveillance instrumentation *) are deleted as part of the proposed changes, since the function of the old instrumentation is effectively superseded by the new, more qualified instruments.
The existing Table 3.2-6, ' Surveillance Instrumentation,' provides operability requirements for instrumentation monitoring various plant parameters, many of which are also located in revised Table 3.2-8. This table was originally intended to assure that instrumentation required for safety related purposes was available to control room operators. The specific instruments were chosen based upon engineering judgement and historical precedence.
Since that time, cignificant advances have been made with regard to instrumentation requirements during accident conditions. This has culminated with the instrumentation requirements of NUREG-0737 and Regulatory Guide 1.97, which form the bases for the revised Table 3.2-8. Since Table 3.2 8 fulfills the intent and design purpose of the existing Table 3.2-6, Table 3.2 6 is considered unnecessary and, the :
e, is being deleted from the Specifications. The associated surveillance requireman s for these instruments, contained in Table 4.2-6, " Minimum Test and Calibration Frequency for Surveillance Instrumentation," are also being deleted. Since Tables 4.21 through 4.2-6 share a 1
SAFETY EVALUATM Page 5 of 9 common list of notes, minor administrative changes are made to these tables and the list of notes to remove the references to Table 4.24.
The following paragraphs describe and discuss the differences between the old and new tables. Information or instruments that are not included on the new tables will be identified and justified.
Applicable Operational Conditions STS Table 3.3.7.51 ' Accident Monitoring instrumentation
- Includes an
- Applicable Operational Conditions
- column. Since the FitzPatrick Technical Specifications do not define or use operational conditions (or mode) as does the STS, a similar restriction is added to Specification 3.2.H. Specification 3.2.H has been revised to require that this instrumentation be operable whenever the reactor is critical with reactor coolant temperature greater than or equal to 2t2*F.
Instrument Ranges The STS do not include instrument range on its Post Accident Instrumentation table.
Measurement range of instrumentation is considered design information and is more appropriately located in the updated FSAR (Reference 1). Commitments made by the Authority with regard to implementation of Regulatory Guide 1.97 (References 3 and 4) and Generic Letter 83-36 assure that the Table 3.2-8 plant parameters will continue to be measured and Indicated in the Control Room to the full range as reviewed and approved by the NRC (Reference 7).
AdditionalInstruments Three Table 3.2 6 plant parameters will continue to be monitored in the revised Table 3.2 8.
These parameters, narrow range suppression chamber water level, drywell suppression chamber differential pressure and safety / relief valve (SRV) position indication are discussed below.
Narrow range suppression chamber water level instrumentation is used to meet the water level requirement of Specification 3.7.A.1.a and b and the surveillance requirement of Specification 4.7.A.1. Inability to monitor water level would prohibit the ability to meet these specifications, and their action statements would be invoked. Since this instrumentation does not monitor a Regulatory Guide 1.97 Category A variable, a reference is made to the new note G. This instrument, although not a " post accident monitoring" instrument, is being retained because it is part of the
- Mark l Containment Short Term improvement" instrumentation. Changes are also made to Specifications 3.7.A.1 and 4.7.A.1 on page 185 and the corresponding Bases on page 188 to note the existence of this instrumentation and its surveillance requirements on Tables 3.2 8 and 4.2-8.
To improve the consistency of the nomenclature used in the Technical Specifications, ' suppression chamber
- and ' pressure suppression chamber" are replaced with " torus
- where they occur. This change is consistent with plant terminology.
Drywell suppression chamber differential pressure instrumentation is used to meet the differential pressure (dp) requirement of Specification 3.7.A.7 and the surveillance requirement of Specification 4.7.A.7. Inoperable dp instrumentation would prohibit the ability to meet these specifications, and their action statements would be invoked. Since this instrumentation does not monitor a Regulatory Guide 1.97 Category A variable, a reference is made to the new note G. This instrument, although not a ' post accidont monitoring
- Instrument, is being retained in the revised Tables 3.2-8 and 4.2 8 because it is part of the
- Mark i Containment Short Term improvement
- Instrumentation. Changes are 1
Attachment ll 5AFETY EVALUATION Page 6 of 9 I
also made to Specification 4.7.A.7 on page 180 and the Bases for Specification 3.7.A on page 190 to note the existence of this instrumentation and its surveillance requirements on Tabies 3.24 and 4.2-8. The instrument check frequency is being revised from once/ shift to onoe/ day. This change makes this surveillanos consistent with the other instrument checks on this table. The daily surveillance frequency combined with the existing control room annunciation of this paramotor provides adequate assurance that the dp instrumentation is being actively monitored and the loss of dp will be immediately recognized by the control room operators.
The safety / relief valve (SRV) position indication instrumentation was added to lables 3.2-6 and 4.24 as Amendment 57 to the Technical Specifications (Reference 5) to incorporate certain of the TMI 2 " Lessons Learned Category 'A requirements of NUREG-0578.
Therefore, this item and its associated notes will be retained and relocated to the revised Table 3.24. The SRV position Indication surveillance requirements are also being retained and are relocated from Table 424 to the revised Table 42-8.
Instrumentation Deleted From Tables 3.2-6 and 42-6 Three Table 3.2 6 plant parameters will not be monitored in the revised Table 3.2-8. These parameters, narrow range reactor water level, control rod position indication, and neutron monitoring have alternate requirements which will assure that the instrumentation remains operable.
Narrow range reactor water level instrumentation is used during plant operations for proper operation of the feedwater control system. Reactor water level Indication in the control room is also provided by the overlapping range of the R.G.1.97 wide range reactor water level instrumentation as required by Table 3.2-8.
Control rod position Indication is required by Specification 3.3.A2.d as a condition for control rod operability. The instrument check is being retained in a new Specification 4.3.A.2.f. The frequency of this surveillance is being changed to once/ day to be consistent with Standard Technical Specification 4.1.3.7.a.
Neutron monitoring instrumentation operability requirements are specified in Table 3.11, Table 3.2-3, and Specification 3.3.B.4.
Additional Changes A change to the Bases 3.2 on page 59 removes the statement that any deviation in instrument readings will initiate an early recalibration. Deviations within reasonable instrument accuracy or " drift allowance
- range between instruments measuring the same plant parameter will not trigger an early recalibration. Gross deviations will, of course, be investigated and instruments will be recalibrated as necessary.
One of the existing Table 32-8 plant parameters will be removed from the revised table.
This parameter, drywell water level, is not a Regulatory Guide 1.97 Category A vt.rlable and, therefore, should not be included in Table 3.2-8.
Table 4.71,
- Minimum Test and Calibration Frequency for Containment Monitoring Systems" is being deleted since the Instrumentation contained in this table (drywell hydrogen and oxygen concentration) is included in the revised Table 4.2 8. Specifications 4.7.A.6.a and 4.7.A.9.a are being revised accordingly to refer to Table 42-8. The Table 32 8 action statement for hydrogen / oxygen concentration is revised to indicate Note F.
This new note refers the reader to Specification 3.7.A.9.
Specification 3.7.A.9 contains
" 1 I
SAFE 1Y EVALUATON Page 7 of 9 the existing action statement for the hydrogen / oxygen analyzers which is more restrictive that the Table 3.24 Note A action.
The set of notes which follow Table 4.24 are applicable to Tables 4.21 through 4.2-6.
Administrative changes are being made which rename this list of notes, revise the five tables which refered to these notes, and relocate the notes to page 84. Note 12 is deleted from this list of notes. This note refers the reader to the
- Definitions" section to define a term. This note does not provide any useful Information and is unncessary. The reference to this note is likewise removed from Table 4.2 3.
Table 4.2 7 is relocated to page 95 and Table 4.24 is relocated to pages 86 and 86a.
Changes are made to the notes to Table 3.24. Note A is revised to remove the ' alternate method of monitoring the appropriate parameter (s)" provision as an attemative to the shutdown requirement. The purpose of the instrumentation on this table is to provide post accident plant information. Alternate methods for monitoring these plant parameters may not be appropriate under post. accident conditions and, therefore, do not meet the intent of this table. Removing this provision is also consistent with the Standard Technical Specifications which do not have an altomative method provision for monitoring these plant parameters. Minor editorial changes are also made to Notes B, C, D, and E of Table 3.2 8. These administrative changes involve only editorial matters, intended to clarify these notes and provide greater consistency with the language of other technical specifications. The content of these notes remains unchanged. A new Note G is added to clarify the status of the narrow range torus water level and drywelltorus differential pressure instrumentation. These instruments are not Regulatory Guide 1.97 Category A instruments and are not qualified for post accident conditions.
The change to Specification 3.7 A.6.a is made to reflect the operational status of the FitzPatrick plant. This specification referred to the startup test program and demontration of plant electrical output, both of which were completed 15 years ago. The contelnment oxygen concentration requirement of this specification remains unchanged.
Ill.
IMPACT OF THE PROPOSED CHANGES The Regulatory Guide 1.97, Revision 2 type A variables were determined by the Authority to be required by control room operators under accident conditions. Instrumentation was installed to monitor and display these plant parameters in the main Control Room. The instruments are designed as Category 1 and are qualified for design basis seismic and post accident environmental conditions. The proposed changes to the Technical Specifications assure that this instrumentation will be operable under postulated accident conditions.
The minimum number of operable channels and action statements in revised Table 3.2 8 were selected based upon the redundancy in the design of the instrumentation system.
This is consistent with the operability requirements for other post-accident instrumentation at the FitzPatrick plant (Reference 6). The proposed surveillance requirements were selected based upon the characteristics of the Instruments involved, the normal operational status of the system, and consistency with other surveillance requirements already on this table The operability and surveillance requirements for the Safety / Relief Valve position indication instruments remain unchanged from the existing specifications.
l
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o
," 1 SAFETY EVALUATION Page 8 of 9 Reformatting Table 3.2-8 makes this table more consistent with the corresponding Table 3.3.7.5-1 in the Standard Technical Specifications (STS). The changes necessary to achieve this consistency with the STS do not result in a level of safety less than that provided by the STS. Design information removed as part of this change (measurement range) is more appropriately located in the updated FSAR and is not required by the STS format. The required instrument ranges are defined in Reference 3.
~
lastruments no longer included in these tables have other Technical Specification requirements which assure their operability or do not fulfill a safety-related function.
Administrative changes are made to correct existing typographical, editorial, or grammatical errors. These changes result in improvements to the specifications and have no adverse safety impact.
The overall impact of the proposed changes is an increase in the level of safety provided by the FltzPatrick plant. The > oposed changes assure that instrurrantation that monitors plant variables and systems curing and following an accident will be available to the 7
control room operators.
(
l l
IV.
EVALUAT!ON OF SIGNirlCANT HAZARDS CONSIDERATIO.N 1
I Operation of the FitzPatrick plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92, since it would not:
I 1.
Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed changes reflect the installation of new control room instrumentation. This new instrumentation has no automatic control functions l
and cannot initiate any type of plant transient or accident. This
^
instrumentation provides a highly reliable and environmentally qualified source of vital plant information in the control room. The changes ensure that instrumentation that would be necessary to mitigate accident conditions will
{
be available in the control room. Requirements for instrument operability and surveillance can not increase in the probability or consequences of a l
previously evaluated accident.
{
2.
create the possibility of a new or different kind of accident from any
{
accident previously evaluated.
l As stated aDove, the proposed changes reflect the insta!lation of new control room instrumentation which have no automatic control functions. This modification to the plant cannot initiate or contribute to any rsew or different kind of accident from any accident previously evaluated.
3.
Involve a significant reduction in a margin of safety.
The proposed change provides operability and surveillance requirements for i
instrumentation qualified for post accident seismic and environmental f
conditions. This provides a greater assurance that control room operators will I
have access to vital plant information under accident conditions. This repreevits an increase in the level of protection provided by the FitzPatrick l
l plant.
The operability and surveillance requirements for the new instrumentation are consistent with those previously approved for post-l accident instrumentation at the FitzPetrick plant. Instrumentation deleted as l
I
)
1
e Attacha "nt k SAFETY EVAJuATION
_,3.'.-
Page 9 of 9 part of these changes have other Technical Specification requirements which assure their operability or do not fulfill a safety related function. The proposed changes do not involve a significant reduction in a margin of safety.
[
V.
IMPLEMENTATION OF THE PROPOSED CHANGE
[
Implementation of the proposed changes will not impact the ALARA or Fire Protection Programs at the RtzPatrick plant, nor will the changes impact the environment.
i l
p VI.
CONCLUSION -
L The changes, as proposed, do not constitute an unrov',ewed safety question as defined in L
10 CFR 50.59. That is, they:
a.-
will not change the probability nor the consequences of an accident or malfunction of l
equipment important to safety as previously evaluated in the Safety Analysis Report; b.
will not increase the possibility of an accident or malfunction of a type different from any previously evaluated in the Safety Analysis Report; c,
will not reduce the margin of safety as defined in the basis for-any technica!
specification; and d.-
involve no significant hazards consideration, as defined in 10CFR 50.92.
l'
.Vii.
REFERENCES
'1.
James A. RtzPatrick Nuclear Power Plant Updated' Final Safety Ardysis t
l1
. Report Section 7.19 and Table 7.191.
s 1
2.
James A. RtzPatrick Nuclear Power Plant Safety Evaluation Repott (SER),
dated November 20,1972, and Supplements.
3.
NYPA letter, C.A. McNeill, Jr. to D.B. Vassallo, dated November 30, 1984, (JPN-84 077), " Supplement No.1 to NUREG-0737 (Generic Letter 82-33),
Regulatory Guide 1.97, Revision 2, implementation Report."
.^
4.
NYPA letter, J.C. Brons to USNRC Document Control Desk, dated November.
p 11,1987, (JPN-87 055), providing additional information of Regulatory Guide 1.97 Implementation.
5.
NRC letter, T.A. Ippolito to G.T. Berry (NYPA), dated July 7,1981, issuing Amendment 57 to the FitzPattick Technical Specific.ations, regarding NUREG-0578 TMi-2 Lessons Leamed Category 'A' Items.
6.
- NRC letter, D.E. LaBarge to J.C. Brons (NYPA), dated May 31,1989, issuing Amendment 130 to the FitzPatrick Technical. Specifications, regarding NUREG 0737 items.
7.
NRC lettei, H. Abelson to DE. Brons (NYPA), dated March 14,1988, "Conformance to Regulatory Guide (R.G.) 1.97, Revision 2,"
including Techn! cal Evaluation Report.
. g 8.
NRC Regulatory Guide 1.97, " Instrumentation for Ught. Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 2. dated December,1980.'
.h.
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