BECO-93-156, Proposed Tech Specs Requesting Changes Supporting 24 Month Fuel Cycle (Submittal 3)
| ML20058N026 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/10/1993 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20058N016 | List: |
| References | |
| BECO-93-156, NUDOCS 9312210199 | |
| Download: ML20058N026 (139) | |
Text
{{#Wiki_filter:. .. -_ Attachment B to BECo letter 93 156 Amended Technical Specification Pages l Page l 29 40 45 ! 46a ; 47 ; 48 : 49 50 50a 53a ! 55a 60 ,
~
61 62 63 i 64
- 68 69 77 !
125b 137a 137b 137c 158 158B 158C l 169 : 172 : 173 190 ,- 193 193a 194a . 196 197 201 : l l l l l
- 9312210199 931210 E PDR ADOCK 05000293 P PDR J
l ; l NOTES FOR TABLE 3.1.1 (Cont'd) l
- 2. Permissible to bypass, with control rod block, for reactor protection system j reset in refuel and shutdown positions of the reactor mode switch. ;
- 3. Permissible to bypass when reactor pressure is 5600 psig.
- 4. Permissible to bypass when turbine first stage pressure is 5305 psig. !
- 5. IRM's are bypassed when APRM's are onscale and the reactor mode switch is in ;
the run position.
- 6. The design permits closure of any two lines without a scram being initiated. i
- 7. When the reactor is subcritical, fuel is in the reactor vessel and the reactor wcter temperature is less than 212 F, only the following trip functions need to be operable:
A. Mode switch in shutdown B. Manual scram C. High flux IRM D. Scram discharge volume high level i E. APRM (15%) high flux scram l
- 8. Not required to be operable when primary containment integrity is not j required. !
- 9. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t). ,
- 10. Not required to be operable when the reactor pressure vessel head is not !
bolted to the vessel. '
- 11. Deleted
- 12. Deleted
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs l per level or there is less than 50% of the normal complement of LPRM's to an i APRM. I
- 14. Deleted I
- 15. The APRM high flux trip level setting shall be as specified in the CORE l OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal power.
- 16. The APRM (15%) high flux scram is bypassed when in the run mode.
- 17. The APRM flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.
- 18. Within 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power.
Amendment No. 6, 15, 27, 42, 86, 117, 118, 133, 147, 29 t _Y
l l l 3.1 BASES (Cont'd) Scram Discharae Instrument Volume The control rod drive scram system is designed so that all of the water that is discharged from the reactor by a scram can be accommodated in the discharge , piping. The two scram discharge volumes have a capacity of 48 gallons of water ' each and are at the low points of the scram discharge piping. During normal operation the scram discharge volume system is empty; however, j should it fill with water, the water discharged to the piping could not be l accommodated which would result in slow scram times or partial control rod I insertion. To preclude this occurrence, redundant and diverse level detection I devices in the scram discharge instrument volumes have been provided. The l instruments are set to alarm, initiate a control rod block, and scram the l reactor at three different progressively increasing water levels in the volume. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function properly. 4.1 BASES The reactor protection system is made up of two independent trip systems. There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with General Electric Company Topical Report NEDC-30851P-A, l
" Technical Specification Improvement Analysis for BWR Reactor Protection System," as approved by the NRC and documented in the safety evaluation report (NRC letter to T. A. Pickens from A. Thadani dated July 15,1987).
A comparison of Tables 4.1.1 and 4.1.2 indicates that two instrument channels have not been included in the latter table. These are: mode switch in shutdown and manual scram. All of the devices or sensors associated with these scram functions are simple on-off switches and, hence, calibration during operation is not applicable (i.e., the switch is either on or off). The sensitivity of LPRM detectors decreases with exposure to neutron flux at a slow and approximately constant rate. This is compensated for in the APRM system by calibrating every three days using heat balance data and by calibrating individual LPRM's every 1000 effective full power hours using TIP traverse data. l l Amendment No. 42, 133, 138, 147 40
PNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION Operable Instrument Channels Per Trip System (1) Minimum Available Instrument Trip Level Settinn Action (11 2(7) 2 Reactor Low Water Level 29" indicated level (3) A and D 1 1 Reactor High Pressure $110 psig , D , 2 2 Reactor Low-Low Water Level at or above -49 in. A indicated level (4) 2 2 Reactor High Water Level $48" indicated level (5) B 2(7) 2 High Drywell Pressure $2.5 psig A 2 2 High Radiation Main Steam $7 times normal rated B Line Tunnel (9) full power background 2 2 Low Pressure Main Steam Line 2810 psig (8) B 2(6) 2 High Flow Main Steam Line 5136% of rated steam flow B l 2 2 Main Steam Line Tunnel Exhaust Duct High Temperature $170 F B 2 2 Turbine Basement Exhaust Duct High Temperature $150 F B 1 1 Reactor Cleanup System High Flow $300% of rated flow C 2 2 Reactor Cleanup System High Temperature $150 F C Amendment No. 86, 147, 150 45
- 3. Instrument set point corresponds to 137.96 inches above top of active fuel. l
- 4. Instrument set point corresponds to 77.26 inches above top of active fuel.
- 5. Not required in Run Mode (bypassed by Mode Switch).
- 6. Two required for each steam line.
- 7. These signals also start SBGTS and initiate secondary containment isolation. l f
- 8. Only required in Run Mode (interlocked with Mode Switch). i
- 9. Within 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be ;
I determined and associated trip setpoints shall be set within 24 hours of re- l l establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power. ; I I i l l 1 l l l l i Amendment No. 147 46a l ! /
PNPS TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR GONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trip System (1) Trip Function Trio Level Setting Remarks 2 Reactor Low-Low Water at or above -49 in. 1. In conjunction with Low Level indicated level (4) Reactor Pressure, initiates Core Spray and LPCI.
- 2. In conjunction with liigh Drywell Pressure, 94.4-115.6 second time delay and LPCI or Core Spray pump interlock initiates Auto Blowdown (ADS).
- 3. Initiates llPCI; RCIC.
- 4. Initiates starting of Diesel Generators.
2 Reactor liigh Water Level s+48" indicated Trips 11PC1 and RCIC turbines. level 1 Reactor Low Level >-151" indicated Prevents inadvertent operation l (inside shroud) level of containment spray during accident condition. (Indicative
-of 2/3 core coverage) l 2 Containment liigh Pressure 1.55 5 p 5 1.82 psig Prevents inadvertent operation l of containment spray during accident condition. Instrument is set to trip at or before 1.82 increasing and reset at or before 1.55 decreasing.
Amendment No. 90 47 N _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _..__ __ _ . _ _ , _ _ _ _ _ _ . _ _ _ _ . . _ . _ . __ _ - . _ _ . , , . . .
l I PNPS TABLE 3.2.B (Cont'd) INSTRUMENTATION TIIAT INITIATES OR CONTR.0LS Ti1E CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trin System (1) Trip Function Trip Level Setting Remarks 2 liigh Drywell Pressure $2.5 psig 1. Initiates Core Spray; LPCI; ilPCI.
- 2. In conjunction with Low-Low Reactor Water Level, 94 4 - 115.6 second time .
delay and LPCI or Core Spray pump running, initiates Auto Blowdown (ADS)
- 3. Initiates starting of Diesel Generators
- 4. In conjunction with Reactor Low Pressure initiates closure of IIPCI vacuum breaker containment isolation valves.
1 Reactor Low Pressure 400 psig i 5 Permissive for Open , Core Spray and l LPCI Admission va' s. 1 Reactor Low Pressure 5110 psig In conjunction with PCIS signal permits closure of R11R (LPCI) inj ection valves. 1 Reactor Low Pressure 400 psig i S In conjunction with Low-Low Reactor l Water Level initiates Core Spray and LPCI. 2 Reactor Low Pressure 900 psig i 5 Prevents actuation of LPCI break detection circuit. 2 Reactor Low Pressure 80 psig S Isolates llPCI and in conjunction with l High Drywell Pressure initiates closure of IIPCI vacuum breaker containment isolation valves. Amendment No. 42;-113, 148 48
, - - . _ = - _ . . . _ . . - -. . _.. ._ . - . __ _ - - -.- - - _ . .- .
PNPS TABLE 3.2.B (Cont'd) INSTRUNENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINNENT COOLING SYSTENS i Minimum # of Operable Instrument Channels Per Trio System (1) Trio Function Trio Level Settina Remarks 1 Core Spray Pump Start 0.21 < t < 1 :oc. Initiates sequential starting of Timer CSCS pumps on any auto start. 1 LPCI Pump Start Timer 4.16 < c < 5.84 sec. 1 LPCI Pump Start Timer 9.5 < t < 11.5 sec. 1 Auto Blowdown Timer 2 94.4, 5115.6 sec. In conjunction with low Low Reactor Water Level, High Drywell Pressure and LPCI or Core Spray-Pump running interlock, initiates Auto Blowdown. t 'l 2 ADS Drywell Pressure 9 1 t 5 15.4 min. Permits starting CS and LPCI l Bypass Timer pumps and actuating ADS SRV's if RPV water level is low and drywell pressure is not high. 2 RHR (LPCI) Pump. Discharge 150 10 psig Defers ADS actuation pending
- Pressure interlock confimation of Low Pressure core
cooling system operation. 2 Core Spray Pump Discharge 150 10 psig (LPCI or Core Spray Pump Pressure Interlock running interlock.) 4 2 Emergency Bus Voltage 20-25% of rated 1. Permits closure of the Diesel Relay voltage resets Generator tr un unloaded at less than or emergency bus. equal to 50% l-
- 2. Permits starting of CSCS 4 kV motors.
l Amendment No. 40,-106, 120 49 i 4 m__.-. .--____..__.____.-____..___._,_______._._._.___._t_m._.m_ _ _ _ _ _ _ _ . , _ _ _ . _ _ _ _ . _ _ _ . , _ _ _ . , _ _ _ _ _ . , . ____._______...m . . . . . , , , , _ , . _ _ . , _ ,,.,__m., -
. , , .y,,,m..,ymm,,.,_,,,,m, _ , , , , r.....ry..._._ym,. _
PNPS TABLE 3.2.B (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trio System (1) Trin Function Trip level Settina Remarks 2 Startup Transformer At 0 Volts between 1. Trips Startup Transformer to Loss of Voltage 0.96 5 t 5 1.34 seconds Emergency Bus Breaker. Time Delay.
- 2. Locks out automatic closure of Startup Transformer to Emergency Bus.
- 3. Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
- 4. Prevents simultaneous starting of CSCS components.
- 5. Starts load shedding logic for Diesel Operation in conjunction with (a) Low Low Reactor Water Level and low Reactor Pressure or (b) High drywell pressure or (c) Care Standby Cooling System components in service in conjunction with Auxiliary Transformer breaker open.
Amendment 50
PNPS TABLE 3.2.B (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trio System (1) Trio Function Trip Level Settina Remarks 2 Startup Transformer 3878.7V .51% with 10.24 1. Trips Startup Transformer to Degraded Voltage 0.36 seconds time delay. Emergency Bus Breaker.
- 2. Locks out autcmatic closure of Startup Transformer to Emergency Bus.
- 3. Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
- 4. Prevents simultaneous starting of CSCS components.
- 5. Starts load shedding logic for Diesel Operation in conjunction with a) Low Low Reactor Water Level and Low Reactor Pressure or b) High drywell pressure or c) Core 5tandby Cooling System components in service in conjunction with Auxiliary Transformer breaker open.
Amendment No. 42, 61, 108, 120 50a
PNPS TABLE 3.2.B.1 INSTRUMENTATION THAT MONITORS EMERGENCY BUS VOLTAGE Minimum # of Operable Instrument Channels Per Trip system Function Settinn Remarks 1 Emergency 4160V Buses AS 3958. 5V + 0. 5% , -0.24% Alerts Operator to possible i
& A6 Degraded Voltage ,
with 10.24 1 0.36 seconds degraded voltage conditions. Annunciation (1) seconds time delay Provides permissive to initiate load shedding in conjunction with LOCA signal. i (1) In the event that the alarm system is determined inoperable, commence logging safety related bus voltage everyW
- hour until such time as the alarm is restored to operable status.
Amendment No. 42;-61;-198;-129 53a _ . _ . _ . . . _ _ . . . _ . _ _ . . _ . . _ _ _ . . . . ~ . _ _ _ _ . . . . _ . . _ . . . _ . _ _ _ .
PNPS TABLE 3.2.C-2 CONTROL R0D BLOCK INSTRUMENTATION SETPOINTS J j Trio Function Trio Setpoint
- APRM Upscale (1) (2) i i APRM Inoperative Not Applicable !
APRM Downscale 2 2.5 Indicated on Scale Rod Block Monitor (Power Dependent) (1) (3)
- Rod Block Monitor Inoperative Not Applicable l Rod Block Monitor Downscale (1) (3) ,
t IRM Downscale 2 5/125 of Full Scale IRM Detector not in Startup Position Not Applicable
- i i 1 IRM Upscale 5 108/125 of Full Scale IRM Inoperative Not Applicable SRM Detector not in Startup Position Not Applicable SRM Downscale 2 3 counts /second ;
- SRM Upscale 1 105counts /second l SRM Inoperative Not Applicable Scram Discharge Instrument Volume 518 gallons Water level - High !
l Scram Discharge Instrument Volume - Not Applicable l Scram Trip Bypassed i Recirculation Flow Converter - Upscale 1 120/125 of Full Scale Recirculatior: Flow Converter - Not Applicable Inoperative Recirculation Flow Converter - 5 8% Flow Deviation , Comparator Mismatch l (1) The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT. , (2) When the reactor mode switch is in the refuel or startup positions, the APRM rod block trip setpoint shall be less than or equal to 13% of rated thermal power, but always less than the APRM flux scram trip setting. (3) The RP,M bypass time delay (td2) shall be < 2.0 seconds. I Amendment No. 42, 110, 129, 133, 138 55a
PNPS TABLE 4.2.A MINIMUM TEST AND CALIBRATION FREQUENCY FOR PCIS Instrument Channel (5) Instrument Functional Test Calibration Freauency Instrument Check
- 1) Reactor liigh Pressure (1) Once/3 months None
- 2) Reactor Low-Low Water Level (1) (7) (7) Once/ day
- 3) Reactor High Water Level (1) (7) (7) Once/ day
- 4) Main Steam High Temp. (1) Once/3 months None
- 5) Main Steam High Flow Once/3 months (7) (7) Once/ day l
- 6) Haiti Steam low Pressure (1) (7) (7) Once/ day
- 7) Reactor Water Cleanup High Flow (1) Once/3 months Once/ day
- 8) Reactor Water Cleanup High Temp. (1) Once/3 months None Logic S_Ystem Functional Test (4) (6) FreauenCY
- 1) Main Steam Line Isolation Vvs. Once/ operating cycle Main Steam Line Drain Vvs.
Reactor Water Sample Vvs. l
- 2) RHR - Isolation Vv. Control Once/ operating cycle Shutdown Cooling Vvs.
Head Spray Discharge to Radwaste
- 3) Reactor Water Cleanup Isolation Once/ operating C-te
- 4) Drywell Isolation Vvs. Once/ operating cycle TIP Withdrawal l Atmospheric Control Vvs.
Sump Drain Valves
- 5) Standby Gas Treatment System Once/ operating cycle Reactor Building Isolation l Amendment No. 107, 130 60 4
't
- . _ _ . _ _ _ ____.m..-_m_ _ ._ _ . _ _ _ _ _ . _ - _ - _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ - - __e- __ -*,e_.- - ~-. - - - - ,.-,--___-=m _ _ - - _ _ _ _
PNPS TABLE 4.2.B MINIMUM TEST AND CALIBRATION FREQUENCY FOR CSCS Instrument Channel Instrument Functional Tes.t. Calibration Frequency Instrument Check
- 1) Reactor Water Level (1) (7) (7) Once/ day
- 2) Drywell Pressure (1) (7) (7) Once/ day
- 3) Reactor Pressure (1) (7) (7) Once/ day
- 4) Auto Sequencing Timers NA Once/ operating cycle None
- 5) ADS - LPCI or CS Pump Disch.
Pressure Interlock (1) Once/3 months None
- 6) Start-up Transf. (4160V)
- a. Loss of Voltage Relays Monthly Once/ operating cycle None ,
- b. Degraded Voltage Relays Monthly Once/ operating cycle None
- 7) Trip System Bus Power Monitors Once/ operating cycle NA Once/ day
- 8) Recirculation System d/p (1) Once/3 months Once/ day
- 9) Core Spray Sparger d/p NA Once/18 months Once/ day l
- 10) Steam Line High Flow (HPCI & RCIC) (1) Once/3 months None
- 11) Steam Line High Temp. (HPCI & RCIC) (1) Once/3 months None
- 12) Safeguards Area High Temp. (1) Once/3 months None
- 13) RCIC Steam Line Low Pressure (1) Once/3 months None
- 14) HPCI Suction Tank Levels (1) Once/3 months None
- 15) Emergency 4160V Buses A5 & A6 Monthly Once/ operating Cycle None Loss of Voltage Relays 61 Amendment No. 42,-615-99, 148 i
_ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _- _
PNPS TABLE 4.2.B MINIMUM TEST AND CALIBRATION FRE0VENCY FOR CSCS Logical System Functional Test (4) (6) Freauency Remarks
- 1) Core Spray Subsystem Once/ operating cycle
- 2) Low Press. Coolant Injection Subsystem Once/ operating cycle
- 3) Containment Spray Subsystem Once/ operating cycle
- 4) HPCI S'Jbsystem Once/ operating cycle
- 5) HPCI Subsystem Auto Isolation Once/ operating cycle
- 6) ADS Subsystem Once/ operating cycle
- 7) RCIC Subsystem Auto Isolation -
Once/ operating cycle
- 8) Diesel Generator Initiation Once/ operating cycle
- 9) Area Cooling for Safeguard System Once/ operating cycle i
i 4 Amendment No. 130 62
PNPS TABLE 4.2.C MINIMUM TEST AND CALIBRATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATION Instrument Channel Instrument Functional Calibration Instrument Check Test APRM - Downscale Once/3 Months Once/3 Months Once/ Day APRM - Upscale Once/3 Months Once/3 Months Once/ Day APRM - Inoperative , Once/3 Months , Not Applicable Once/ Day IRM - Upscale (2) (3) Startup or Control Shutdown (2) IRM - Downscale (2) (3) Startup or Control Shutdown (2) IRM - Inoperative (2) (3) Not Applicable (2) RBM - Upscale Once/3 Months once/6 Months Once/ Day RBM - Downscale Once/3 Months once/6 Months once/ Day RBM - Inoperative Once/3 Months Not Applicable Once/ Day SRM - Upscale (2) (3) Startup or Control Shutdown (2) SRM - Inoperative (2) (3) Not Applicable (2) l SRM - Detector Not in Startup Position (2) (3) Not Applicable (2) SRM - Downscale (2) (3) Startup or Control Shutdown (2) IRM - Detector Not in Startup Position (2) (3) Not Applicable (2) Scram Discharge Instrument Volume once/3 Months Refuel Not Applicable Water Level-liigh Scram Discharge Instrument once/3 Months Not Applicable Not Applicable Volume-Scram Trip Bypassed Recirculation Flow Converter Not Applicable Once/ Operating Cycle Once/ Day [ Recirculation Flow Converter-Upscale Once/3 Months Once/3 Months once/ Day Recirculation Flow Converter-Inoperative Once/3 Months Not Applicable Once/ Day Recirculation Flow Converter-Comparator once/3 Months once/3 Months once/ Day Off Limits Recirculation Flow Process Instruments Not Applicable Once/ Operating Cycle Once/ Day l Lonic System Functional Test (4) (6) System Logic Check once/ Operating cycle l Amendment No. 110,-130, 147 63
PNPS TABLE 4.2.D MINIMUM TEST AND CALIBRATION FRE0VENCY FOR RADIATION MONITORING SYSTEMS Instrument Channels Instrument Functional Calibration Instrument Check (2) Test
- 1) Refuel Area Exhaust Monitors - Upscale (1) Once/3 months Once/ day
- 2) Refuel Area Exhaust Monitors - Downscale (1) Once/3 months Once/ day Loaic System Functional Test (4) (6) FreauencY
- 1) Reactor Building Isolation Once/ operating cycle
- 2) Standby Gas Treatment System Actuation Once/ operating cycle I
t Amendment No. 89, 130 64
i i l BASES: 3.2 In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block, and standby gas treatment systems. The objectives of the Specifications are, (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods ,
- when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
l 1 Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety. The set l points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations. Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever primary containment integrity is required. l l The instrumentation which initiates primary system isolation is connected in a 1 dual bus arrangement. l ! The low water level instrumentation closes all isolation valves except l l those in Groups 1, 4 and 5. This trip setting is adequate to prevent core uncovery in the case of a break in the largest line assumina a 60 second valve closing time. Requh ed closing times are less than this. The low low reactor water level instrumentation closes the Main Steam Line Isolation Valves, Main Steam Drain Valves, Recirc Sample Valves (Group 1) activates the CSCS subsystems, starts the emergency diesel generators and trips the recirculation pumps. This trip setting level was chosen to be
- high enough to prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that no fuel damage will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system isolation are initiated in time to meet the above criteria.
l Amendment No. 105, 113 68
3.2 BASES (Cont'd) The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here also. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of Group 1 isolation valves. Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. The primary function of the instrumentation is to detect a break in the main steam line. For the worst case accident, main steam line break outside the drywell, the steam flow trip setting in conjunction with the flow limiters and main steam line valve closure, limits the mass iriventory loss such that fuel is not uncovered, fuel temperatures remain approximately 1000 F and release of radioactivity to the environs is well below 10 CFR 100 guidelines. Temperature monitoring instrumentation is provided in the main steam line tunnel and the turbine basement to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves. The setting of 170 F for the main steam line tunnel detector is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop acci-1 l Amendment No. 34, 113 69
4.2 BASES (Cont'd) The automatic pressure relief instrumentation can be considered to be a 1 out of 2 logic system and the discussion above applies also. The instrumentation which is required for the recirculation pump trip and alternate rod insertion systems incorporate analog transmitters. ; The transmitter calibration frequency is once per refueling outage, which is consistent with both the equipment capabilities and the ; requirements for similar equipment used at Pilgrim. The Trip Unit i Calibration and Instrument functional Test is specified at monthly, which is the same frequency specified for other similar protective devices. An instrument check is specified at once per day; this is considered to be an appropriate frequency, commensurate with the design applications and the fact that the recirculation pump trip and alternate rod insertion systems are backups to existing protective instrumentation. Control Rod Block and PCIS instrumentation common to RPS instrumentation have surveillance intervals and maintenance outage times selected in accordance with NEDC-30851P-A, Supplements 1 and 2 as approved by the NRC and documented in SERs (letters to D. N. Grace from C. E. Rossi dated September 22, 1988 and January 6, 1989). ; A logic system functional test interval of 24 months was selected to l minimize the frequency of safety system inoperability due to testing and , to minimize the potential for inadvertent safety system trips and their . attendant transients. l F k Amendment No. 42, 121, 130, 147 77
l \ l r I r ! LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS I ! l 3.6.C.2 Leakaae Detection Systems 4.6.C.2 Leakaoe Detection Systems ! (Cont'd) (Cont'd) l
- 2. One channel of a drywell 2. An instrument channel !
atmospheric particulate calibration at least radioactivity monitoring once per operating : system, or cycle. i
- 3. One channel of a drywell b. For each required drywell ,
atmospheric gaseous atmospheric radioactivity ] radioactivity monitoring monitoring system perform . system. l
- 1. An instrument' check at
- b. 1. At least one drywell sump least once per day, I monitoring system shall be ,
Operable; otherwise, be in 2. An instrument functional i Hot Shutdown within the test at least once per ! next 12 hours and in Cold 31 days, and l Shutdown within the following 24 hours. 3. An instrument channel calibration at least !
- 2. At least one gaseous or once per operating ;
particulate radioactivity cycle. ; monitoring channel must be Operable; otherwise, reactor operation may ! continue for up to 31 days ! provided grab samples are ! obtained and analyzed every ! 24 hours, or be in Hot ! Shutdown within the next 12 , hours and in Cold Shutdown i within the following 24 hours. i
- c. With no required leakage ,
detection systems Operable, be in Cold Shutdown within 24 hours. Amendment No. 139 125b _y_-___ _
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.6.I Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snubbers)
- 1. During all modes of operation The following surveillance except Cold Shutdown and Refuel, requirements apply to all safety i all safety-related snubbers related hydraulic and mechanical l listed in PNPS Procedures shall snubbers listed in PNPS Procedures.
be operable except as noted in I 3.6. I.2 through 3.6. I.3 below. The required visual inspection )' interval varies inversely with the An Inoperable Snubber is a observed cumulative number of properly fabricated, installed inoperable snubbers found during an and sized snubber which cannot inspection. Inspections performed I pass its functional test. before that interval has elapsed may be used as a new reference point to i Upon determination that a determine the next inspection. I snuhber is either iuproperly However, the results of such early ; fabricated, installed or sized, inspections performed before the ! the ctrrective action will be as original time interval has elapsed ' specified for an inoperable may not be used to lengthen the snuober in Section 3.6.I.2. required interval, j
- 2. From and after the time that a Number of snubbers found inoperable ;
snubber is determined to be during inspection or during inoperable, replace or repair inspection interval: the snubber during the next 72 hours, and initiate an Subsequent engineering evaluation to Inoperable Visual Inspec-determine if the components Snubbers tion Interval supported by the snubber (s) were adversely affected by the 0 24 Months 25% inoperability of the snubbers 1 18 Months 1 25% and to ensure that the supported 2 12 Months 25% l component remains capable of 3,4 6 Months i 25% meeting its intended function in 5,6,7 124 Days 25% , the specific safety system 8,9 62 Days 25% j l involved. 10 or more 31 Days i 25% : i Further corrective action for The required inspection interval this snubber, and all shall not be lengthened more than generically susceptible one step at a time. snubbers, shall be determined by an engineering evaluation. Snubbers may be categorized in two groups, " accessible" or From and after the time a J
- 3. " inaccessible" based on their snubber is determined to be accessibility for inspection during inoperable, improperly reactor operation. These two groups fabricated, improperly installed may be inspected independently or improperly sized, if the according to the above schedule.
requirements of Section(s) 3.6.I.1 and 3.6.I.2 cannot be 1. Visual Inspection Acceptance met, then the affected safety Criteria system, or affected portions of that system, shall be declared A. Visual inspections shall inoperable, and the limiting verify: condition for that system entered, as appropriate. I Amendment No. 20, 60, 93 137a
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.6.I Shock Suppressors (Snubberst 4.6.I Shock Suppressors (Snubbers)
- 4. Snubbers may be added to, or 1. That there are no visible removed from, per 10 CFR 50.59, indications of damage or safety related systems without impaired operability.
prior NRC approval. The ! addition or deletion of snubbers 2. Attachments to the foundation shall be reported to the NRC in or support structure are such accordance with 10 CFR 50.59. that the functional capability of the snubber is not suspect. B. Snubbers which appear INOPERABLE as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval provided that:
- 1. The cause of the rejection is clearly established and remedied for that particular snubber, and
- 2. The affected snubber is functionally tested, when necessary, in the as found condition and determined OPERABLE per specifications 4.6.I.2.8., 4.6.I.2.C., as applicable.
i C. For any snubber determined l l inoperable per specification l 4.6.I.2, clearly establish the cause of rejection and remedy the problem for that snubber, I and any generically susceptible snubber. l 2. Functional Tests (Hydraulic and Mechanical Snubbers) A. Schedule At least once per operating cycle, a representative sample l (12.5% of the total of each type: hydraulic, mechanical) of snubbers in use in the plant shall be functionally tested, either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of l l Amendment No. 20, 60, 93 137b
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 4.6.1 Shock Suporessors (Snubbers) Specification 4.6.I.2.B, or ! 4.6.I.2.C, as applicable, an additional 12.5% of that l type of snubber shall be functionally tested. B. General Snubber Functional Test Acceptance Criteria (Hydraulic and Mechanical) , The general snubber functional test shall verify that:
- 1. Activation (restraining action) is achieved within the specified range of velocity or acceleration in ;
both tension and t compression. l
- 2. Snubber release, or bleedrate, as applicable, where required, is within the specified range in .
compression or tension. - For snubbers specifically required not to displace ! under continuous load, the ; ability of the snubber to withstand load without , displacement shall be i , verified. l C. Mechanical Snubbers functional- ! Test Acceptance Criteria i The mechanical snubber functional test shall verify that:
- 1. The force that initiates free movement of the snubber rod in either tension or compression is
, less than the specified I maximum drag force. Drag force shall not have increased more than 50% since the last functional test.
- 3. Snubber Service Life Monitorina A. A record of the service life l
l Amendment No. 60 137c
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.7.B Standby Gas Treatment System 4.7.B Standby Gas Treatment System and Control Room Hiah and Control Room Hiah Efficiency Air Filtration Efficiency Air Filtration System System
- 1. Standby Gas Treatment System 1. Standby Gas Treatment System l
- a. Except as specified in a. (1.) At least once per 3.7.B.l.c below, both trains operating cycle, it shall of the standby gas treatment be demonstrated that system and the diesel pressure drop across the l generators required for combined high efficiency i operation of such trains shall filters and charcoal be operable at all times when adsorber banks is less than ;
secondary containment 8 inches of water at 4000 i integrity is required or the cfm. reactor shall be shutdown in 36 hours. (2.) At least once per , operating cycle, j
- b. (1.) The results of the in- demonstrate that the place cold 00P tests on inlet heaters on each train HEPA filters shall show are operable and are capable 199% D0P removal. The of an output of at least 14 !
results of halogenated kW. l hydrocarbon tests on I charcoal adsorber banks (3.) The tests and analysis of i shall show 299% Specification 3.7.B.I.b. ) l halogenated hydrocarbon shall be performed at least - Once per operating cycle removal. l l or following painting, fire ; l (2.) The results of the or chemical release in any i laboratory carbon sample ventilation zone l l analysis shall show 195% communicating with the i methyl iodide removal at system while the system is a velocity within 10% of operating that could system design, 0.5 to 1.5 contaminate the HEPA filters i mg/m3 inlet methyl iodide or charcoal adsorbers. I concentration, 270% R.H. 1 and 2190 F. The analysis (4.) At least once per results are to be operating cycle, automatic verified as acceptable initiation of each branch of within 31 days after the standby gas treatment l sample removal, or system shall be declare that train demonstrated, with inoperable and take the Specification 3.7.B.I.d actions specified satisfied. 3.7.B.I.c. I t l l Amendment No. 50, 51, 52, 112, 144 158
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.7.B (Continued) 4.7.B (Continued)
- 2. Control Room Hiah Efficiency Air 2. Control Room Hiah Efficiency Air Filtration System Filtration System
- a. Except as specified in a. At least once per operating Specification 3.7.B.2.c cycle the pressure drop across below, both trains of the each combined filter train shall Control Room High Efficiency be demonstrated to be less than Air Filtration System used 6 inches of water at 1000 cfm or for the processing of inlet the calculated equivalent.
air to the control room under accident conditions and the b. (1.) The tests and analysis of diesel generator (s) required Specification 3.7.B.2.b for operation of each train shall be performed once of the system shall be per operating cycle or operable whenever secondary following painting, fire or containment integrity is chemical release in any required and during fuel ventilation zone handling operations. communicating with the system while the system is ,
- b. (1.) The results of the in- operating. '
place cold D0P tests on HEPA filters shall show (2.) In-place cold D0P testing 199% D0P removal. The shall be performed after j results of the each complete or partial halogenated hydrocarbon replacement of the HEPA tests on charcoal filter bank or after any adsorber banks shall structural maintenance on show 299% halogenated the system housing which hydrocarbon removal when could affect the HEPA test results are filter bank bypass leakage. extrapolated to the initiation of the test. (3.) Halogenated hydrocarbon testing shall be performed (2.) The results of the after each complete or laboratory carbon sample partial replacement of the analysis shall show 195% charcoal adsorber bank or methyl iodide removal at after any structural a velocity within 10% of maintenance on the system system design, 0.05 to housing which could affect 0.15 mg/m3 inlet methyl the charcoal adsorber bank iodide concentration, bypass leakage. l 270% R.H., and 2125 F. l The analysis results are (4.) Each train shall be l to be verified as operated with the heaters l acceptable within 31 in automatic for at least l days after sample 15 minutes every month. , removal, or declare that l train inoperable and (5.) The test and analysis of take the actions Specification 3.7.B.2.b.(2) specified in 3.7.B.2.c. shall be performed after l every 720 hours of system i operation. ; During RF0 #9, one train can be without its safety-related bus and/or its emergency diesel generator without entering the LCO action statement i provided the conditions listed on page 158A are met. Amendment No. 50, 51, 52, 101, 112, 144 158B
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS . 3.7.B (Continued) 4.7.B (Continued)
- c. From and after the date that c. At least once per operating one train of the Control Room cycle demonstrate that the High Efficiency Air inlet heaters on each train Filtration System is made or are operable and capable of found to be incapable of an output of at least 14 kw.
supplying filtered air to the control room for any reason, d. Perform an instrument reactor operation or functional test on the refueling operations are humidistats controlling the i permissible only during the heaters once per operating I succeeding 7 days providing cycle. ! that within 2 hours all : active components of the other CRHEAF train shall be demonstrated operable. If the system is not made fully operable within 7 days; ; reactor shutdown shall be initiated and the reactor shall be in cold shutdown within the next 36 hours and irradiated fuel handling operations shall be terminated within 2 hours. i Fuel handling operations in progress may be completed. '
- d. Fans shall operate within 10% of 1000 cfm. ;
i l l During RF0 #9, one train can be without its safety-related bus and/or its emergency diesel generator without entering the LC0 action statement provided the conditions listed on page 158A are met. 1 1 I Amendment No. 50, 51, 57, 112, 144 158C i
BASES: 3.7.A & 4.7.A Primary Containment Group 6 - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level , isolation is provided. Group 7 - The HPCI vacuum breaker line is designed to remain operable when the HPCI system is required. The signals which initiate isolation of the HPCI vacuum breaker line are indicative of a break inside containment and reactor pressure below that at which HPCI can operate. l The maximum closure time for the automatic isolation valves of the primary , I containment and reactor vessel isolation control system have been selected in ; consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment. In satisfying this design intent an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance, prior to exceeding the design closure times. In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding ! perforations would be avoided for main steam valve closure times, including , instrument delay, as long as 10.5 seconds. ! These valves are highly reliable, have low service requirements and most are normally closed. The initiating sensors and associated trip channels are also , checked to demonstrate the capability for automatic isolation. The test ! interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 7 that a line will not isolate. More frequent testing for valve operability results in a greater assurance that the valve will be operable when needed. The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability. l The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment. A program for periodic testing and examination of the excess flow check valves is in place. Primary Containment Painting The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage assures the paint is intact. Experience at Pilgrim Station and other BWRs with this type of paint indicates that the inspection interval is adequate. 1 1 Amendment No. 113 169 t l _ - .-. , .~ _ - , . . . . . = , . - . _ . -
BASES: 3.7.B.1 and 4.7.B.1 - Standby Gas Treatment System The Standby Gas Treatment System is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. Upon containment isolation, both standby gas treatment fans are
- designed to start to bring the reactor building pressure negative so that all
, leakage should be in-leakage. After a preset time delay, the standby fan 1 automatically shuts down so the reactor building pressure is maintained approximately 1/4 inch of water negative. Should one system fail to start, )
i the redundant system is designed to start automatically. Each of the two trains has 100% capacity. High Efficiency Particulate Air (HEPA) filters are installed before and after , the charcoal adsorbers to minimize potential release of particulates to the ! l environment and to prevent clogging of the iodine adsorbers. The charcoal j adsorbers are installed to reduce the potential release of radiciodine to the ' environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and i a HEPA filter efficiency of at least 99 percent removal of cold D0P , particulates. The laboratory carbon sample test results should indicate a , methyl iodide removal efficiency of at least 95 percent for expected accident conditions. The specified efficiencies for the charcoal and particulate filters is sufficient to preclude exceeding 10 CFR 100 guidelines for the accidents analyzed. The analysis of the loss of coolant accident assumed a charcoal adsorber efficiency of 95% and TID 14844 fission product source terms, hence, installing two banks of adsorbers and filters in each train provides adequate margin. A 14 kW heater maintains relative humidity below 70% in order to ensure the efficient removal of methyl iodide on the impregnated charcoal adsorbers. Considering the relative simplicity of the heating circuit, the test frequency of once/ operating cycle is adequate to l demonstrate operability.
- i i Air flow through the filters and charcoal adsorbers for 15 minutes each month :
assures operability of the system. Since the system heaters are automatically , controlled, the air flowing through the filters and adsorbers will be 570% , relative humidity and will have the desired drying effect. , l Tests of impregnated charcoal identical to that used in the filters indicate that shelf life of five years leads to only minor decreases in methyl iodide removal efficiency. Hence, the frequency of laboratory carbon sample analysis '
, is adequate to demonstrate acceptability. Since adsorbers must be be removed to perform this analysis this frequency also minimizes the system out of service time as a result of surveillance testing. In addition, although the 3
halogenated hydrocarbon testing is basically a leak test, the adsorbers have - charcoal of known efficiency and holding capacity for elemental iodine and/or i methyl iodide, the testing also gives an indication of the relative efficiency )
- of the installed system. The 31 day requirement for the ascertaining of test ,
results ensures that the ability of the charcoal to perform its designed ; function is demonstrated and known in a timely manner.
]
The required Standby Gas Treatment System flow rate is that flow, less than or equal to 4000 CFM which is needed to maintain the Reactor Building at a 0.25 inch of water negative pressure under calm wind conditions.. This capability is adequately demonstrated during Secondary Containment Leak Rate Testing performed pursuant to Technical Specification 4.7.C.I.c. a i Amendment No. 42, 112 172 l
i l l BASES: l 3.7.B.1 and 4.7.B.1 (continued) The test frequencies are adequate to detect equipment deterioration prior to l significant defects, but the tests are not frequent enough to load the filters l or adsorbers, thus reducing their reserve capacity too quickly. The filter , testing is performed pursuant to appropriate procedures reviewed and approved l l by the Operations Review Committee pursuant to Section 6 of these Technical Specifications. The in-place testing of charcoal filters is performed by , injecting a halogenated hydrocarbon into the system upstream of the charcoal ! adsorbers. Measuretrents of the concentration upstream and downstream are l made. The ratio of the inlet and outlet concentrations gives an overall indication of the leak tightness of the system. A similar procedure substituting dioctyl phthalate for halogenated hydrocarbon is used to test the l HEPA filters. Pressure drop tests across filter and adsorber banks are performed to detect ! plugging or leak paths though the filter or adsorber media. Considering the i l relatively short times the fans will be run for test purposes, plugging is unlikely and the test interval of once per operating cycle is reasonable. l System drains and housing gasket doors are designed such that any leakage $ would be in-leakage from the Standby Gas Treatment System Room. This ensures l that there will be no bypass of process air around the filters or adsorbers. Only one of the two Standby Gas Treatment Systems (SBGTS) is needed to maintain the secondary containment at a 0.25 inch of water negative pressure l upon containment isolation. If one system is found to be inoperable, there is l no immediate threat to the containment system performance and reactor l operation or refueling activities may continue while repairs are being made. In the event one SBGTS is inoperable, the redundant system's active components I will be tested within 2 hours. This substantiates the availability of the j operable system and justifies continued reactor or refueling operations. l l If both trains of SBGTS are inoperable, the plant is brought to a condition l l where the SBGTS is not required. l Amendment No. 42, 112, 173
TABLE 4.8-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Instrument Source Channel Functional Instrument Check Check Calibration Test
- 1. Gross Beta or Gamma Radioactivity Monitors Providing Alarm and Auto-matic Isolation
- a. Liquid Radwaste Effluents Line i NA Once per Quarterly l 18 months 2
- 2. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line i NA Once per Quarterly l 18 months 1During or prior to release via this pathway.
2Previously established calibration procedures will be used for these requirements. 4 Amendment No. 89 190
TABLE 4.8-4 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Source Instrument Functional Instrument Check Check Calibration Test
- 1. Main Stack Effluent Monitoring System
- a. Noble Gas Activity Monitor Daily l Monthly Once per Quarterly l (Two Channels) 18 Months 4
- b. lodine Sampler Cartridge NA NA NA NA
- c. Particulate Sampler Filter NA NA NA NA
- d. Effluent System Flow Rate Daily l NA Once per Quarterly l Measuring Device 18 Months
- e. Sampler Flow Rate Measuring Daily l NA Once per Quarterly l Device 18 Months
- 2. Reactor Building Ventilation Effluent Monitoring System
- a. Noble Gas Activity Monitor Daily l Monthly Once per Quarterly l 18 Months 4
- b. Iodine Sampler Cartridge hi NA NA NA
- c. Particulate Sampler Filter NA NA NA NA
- d. Effluent System Flow Rate Daily l NA Once per Quarterly l Measuring Device 18 Months 1
Amendment No. 89 193
i TABLE 4.8-4 (continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS j Instrument Instrument Source Instrument Functional j Instrument Check Check Calibration Test
- e. Sampler Flow Rate Measuring Daily l NA Once per Quarterly l 1 Device 18 Months j
; 3. Steam Jet Air Ejector Radioactivity
- Monitor
- a. Noble Gas Activity Monitor Daily 3 NA Once per Quarterly
, operaging
- cycle
- 4. Augmented Offgas Treatment System Explosive Gas Monitoring System
. a. Hydrogen Monitor Daily 2 NA Quarterly 5 Monthly 1 During releases via this pathway 2
- During augmented offgas treatment system operation.
3 During operation of the steam jet air ejector. 4 Previously established calibration procedures will be used for these requirements. 5 Calibrate at 2 points with standard gas samples differing by at least 1% but not exceeding 4%. i i i Amendment No. 89 193a
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.9.A Auxiliary Electrical Eauipment Surveillance (Cont'd)
- 1. Verifying de-energization of the emergency buses and load shedding from the emergency buses. ,
- 2. Verifying the diesel starts ;
from ambient condition on the auto-start signal, energizes i the emergency buses with permanently connected loads, energizes the auto-connected emergency loads through the load sequence, and operates for 2 5 minutes while its generator is loaded with the emergency loads. During performance of this surveillance verify that HPCI and RCIC inverters do not trip. The results shall be logged.
- c. Once per operating cycle with the diesel loaded per !
4.9. A.l.b verify that on diesel generator trip, secondary (offsite) AC power is automatically connected within 11.8 to 13.2 seconds l to the emergency service buses and emergency loads are energized through the load sequencer in the same manner as described in 4.9.A.1.b.1. 1 The results shall be logged. , i i Amendment No. 42, 61, 141 194a 4
r 1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 1 3.9.B Operation with Inoperable 4.9.A Auxiliary Electrical Eauipment l Eauipment Surveillance (Cont'd) l l Whenever the reactor is in Run 3. Emergency 4160V Buses A5-A6 l Mode or Startup Mode with the Degraded Voltage Annunciation j reactor not in a Cold Condition, System. the availability of electric power shall be as specified in a. Once each operating cycle, 3.9.B.1, 3.9.B.2, 3.9.B.3, calibrate the alarm sensor. ! 3.9.B.4, and 3.9.B.5.
- 1. From and after the date that channel functional test on the incoming power is not available alarm system.
from the startup or shutdown transformer, continued reactor c. In the event the alarm system - operation is permissible under is determined inoperable under this condition for seven days. 3.b above, commence logging i During this period, both diesel safety related bus voltage ; generators and associated every 30 minutes until such ; emergency buses must be time as the alarm is restored ! demonstrated to be operable. to operable status. l
- 2. From and after the date that 4. RPS Electrical Protection incoming power is not available Assemblies from both startup and shutdown i transformers, continued a. Each pair of redundant RPS t operation is permissible, EPAs shall be determined to provided both diesel generators be operable at least once and associated emergency buses per 6 months by performance are demonstrated to be operable, of an instrument functional ;
all core and containment cooling test. : systems are operable, reactor ' power level is reduced to 25% of b. Once per 18 months, each l t design and the NRC is notified pair of redundant RPS EPAs within one (1) hour as required shall be determined to be by 10CFR50.72. operable by performance of an instrument calibration !
- 3. From and after the date that one and by verifying tripping '
of the diesel generators or of the circuit breakers associated emergency bus is made upon the simulated or found to be inoperable for conditions for automatic any reason, continued reactor actuation of the protective operation is permissible in relays within the following 1 accordance with Specification limits: : 3.5.F if Specification 3.9.A.1 and 3.9.A.2.a are satisfied. Overvoltage i 132 volts i Undervoltage 1 108 volts i 4 From and after the date that one Underfrequency 2 57Hz of the diesel generators or 1 associated emergency buses and either the shutdown or startup transformer power source are made Amendment No. 88, 127,145 196
t i LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS l
}
3.9 AUXILI ARY ELECTRICAL SYSTEM (Cont) B. Operation with Inoperable Equipment (Cont) l or found to be inoperable for any reason, continued reactor ; operation is permissible in ! accordance with Specification 3.5.F provided either of the following conditions are satisfied {
- a. The startup transformer and l both offsite 345 kV !
transmission lines are j available and capable of automatically supplying . j i auxiliary power to the ; emergency 4160 volt buses. :
- b. A transmission line and associated shutdown j transformer are available and capable of automatically supplying auxiliary power to ,
the emergency 4160 volt ( buses, j
- 5. From and after the date that one ,
of the 125 or 250 volt battery l systems is made or found to be inoperable for any reason, l continued reactor operation is I permissible during the succeeding three days within electrical safety considerations, provided repair work is initiated in the most expeditious manner to return the ' failed component to an operable state, and Specification 3.5.F is satisfied.
- 6. With the emergency. bus voltage less than 3958.5V but above 3878.7V(excluding transients) during normal operation, transfer the safety related buses to the diesel generators.
If grid voltage continues to degrade be in at least Hot Shutdown within the next 4 hours and in Cold Shutdown within the following 12 hours unless the grid conditions improve. l Amendment No. 42;-61;-88 1 -129 1 -127 197
1 a
- l BASES: (Cont'd) i l
d 4.9 ) ! deliver full flow. Periodic testing of the various components, plus a functional test once per cycle, is sufficient to maintain adequate reliability. ; Although station batteries will deteriorate with time, utility experience ;
- indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification has been demonstrated over the years j to provide an indication of a cell becoming irregular or unserviceable long t before it becomes a failure. f The Service Discharge Test provides indication of the batteries' ability to i i satisfy the design requirements (battery duty cycle) of the associated de system.
This test will be performed using simulated or actual loads at the rates and for i l the duration specified in the design load profile. A once per cycle testing ! interval was chosen to coincide with planned outages. ! The Performance Discharge Test provides adequate indication and assurance that the batteries have the specified ampere hour capacity. The results of these i tests will be logged and compared with the manufacturer's recommendations of ; acceptability. This test is performed once every five years in lieu of the i Service Discharge test that would normally occur within that time frame. l The diesel fuel oil quality must be checked to ensure proper operation of the ! , diesel generators. Water content should be minimized because water in the fuel l could contribute to excessive damage to the diesel engine. l The electrical protection assemblies (EPAs) on the RPS inservice power supplies, l ! either two motor generator sets or one motor generator and the alternative i supply, consist of protective relays that trip their incorporated circuit breakers on overvoltage, undervoltage, or underfrequency conditions. There are ; two EPAs in series per power source. It is necessary to periodically test the ; relays to ensure the sensor is operating correctly and to ensure the trip unit is , , operable. Based on experience at conventional and nuclear power plants, a six-j month frequency for the channel functional test is established. This frequency is consistent with the Standard Technical Specifications. l l The EPAs of the power sources to the RPS shall be determined to be operable by j performance of a channel calibration of the relays once per 18 months. During l calibration, a transfer to the alternative power source is required; howe /er, : prior to switching to alternative feed, de-energization of the applicable i1G ' set power source must be accomplished. This results in a half scram on the , channel being calibrated until the alternative power source is connected and the half scram is cleared. Based on operating experience, drift of the EPA protective relays is not significant. , l l Amendment No. 127, 135, 141, 145 201 ! i I
i e l Attachment C to BECo letter 93 1.re au Marked-Up Pages l i i i
l l i i NOTES FOR TABLE 3.1.1 (Cont'd) d l
-l ,
- 2. Permissible to bypass, with control rod block, fo reactor protection system reset in refuel and shutdown positions of the re ctor mode switch. 4
- 3. Permissible to bypass when reactor pressure is'y600 psig. . .w p -
- 4. Permissible to bypass when turbine first stage pressure is ' lass thaw 305 psig.
- 5. IRM's are bypassed when APRM's are onscale and the reactor mo e switch is in the ;
run position. '
- 6. The design permits closure of any two lines without a scram being initiated.
- 7. When the reactor is subtritical, fuel is in the reactor vessel and the reactor watertemperatureislessthan212ff,onlythefollowingtripfunctionsneedto be operable:
A. Mode switch in shutdown B. Manual scram C. High flux IRM D. Scram discharge volume high level l E. APRM (15%) high flux scram ,
- 8. Not required to be operable when primary containment integrity is not required.
- 9. Not required while performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MW(t).
- 10. Not required to be operable when the reactor pressure vessel head is not bolted ,
to the vessel. ;
- 11. Deleted 1
^
- 12. Deleted i (U 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per '
level or there is less than 50% of the normal complement of LPRM's to an APRM.
- 14. Deleted l
- 15. The APRM high flux trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal l
power. i l
- 16. The APRM (15%) high flux scram is bypassed when in the run mode.
- 17. The APRM flow biased high flux scram is bypassed when in the refuel or I startup/ hot standby modes. l
- 18. Within 24 hours prior to the planned start of hydrogen injection with the l reactor power at greater than 20% rated power, the normal full power radiation i background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels I resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power.
l I ' Q C NR _i si m rft . 65 ) Amen men fo. 6, 15, 27, 42, 86, 117, 118, 133, gm 29
)
3.1 BASES (Cont'd) Scram Discharge Instrument Volume : V l The control rod drive scram system is designed so that all of the water ,1 l that is discharged from the reactor by a scram can be accommodated in the ' l discharge piping. The two scram discharge volumes have a capacity of 48 gallons of water each and are at the low points of the scram discharge
) I piping. f During normal operation the scram discharge volume system is empty; ;
however, should it fill with water, the water discharged to the piping ',, I could not be accommodated which would result in slow scram times or l partial control rod insertion. To preclude this occurrence, redundant and f I diverse level detection devices in the scram discharge instrument vo mes ) i
~
have been provided.PT , . . - . - . . . - - . ~ . -os o e I g instrumen alarmf1Wer jpel les ini iate a ' control rod b ock fef e 'na M m ' ' 'm 'm e' !eveO and scrim the reactor - '1 i g e ': ate- leve! re c e M y As indicated above, there is , f b sutticient volume in the piping to accommodate the scram without l impairment of the scram times or amount of insertion of the control rods. q This function shuts the reactor down while sufficient volume remains to l l accommodate the discharged water and precludes the situation in which a ' i j 64 scram would be required but not be able to perform its function properly. ; l l OAO ed ute4. 4.1 BASES -
?
1 ; The reactor protection system is made up of two independent trip systems. m l l There are usually four channels to monitor. each parameter with two f i ! channels in each trip system. The outputs of the channels in a trip / ' system are combined in a logic so that either channel will trip that trip ' - l system. The tripping of both trip systems will produce a reactor scram. j The system meets the intent of IEEE-275 for nuclear power plant protection - R systems. Specified surveillance intervals and surveillance and / ! l maintenance outage times have been determined in accordance with General Electric Company Topical Report NEDC-30851P-A, " Technical Specification Improvement Analysis for BWR Reactor Protection System," as approved by f the NRC and documented in the safety evaluation report (NRC letter to T. ' l A. Pickens from A. Thadani dated July 15,1987). - l A comparison of Tables 4.1.1 and 4.1.2 indicates that two instrument / channels have not been included in the latter table. These are: mode switch in shutdown and manual scram. All of the devices or sensors ' y l associated with these scram functions are simple on-off switches and, ' hence, calibration during operation is not applicable (i.e., the switch is either on or off). / The sensitivity of LPRM detectors decreases with exposure to neutron flux ' at a slow and approximately constant rate. This is compensated for in the / APRM system by calibrating every three days using heat balance data and by calibrating individual LPRM's every 1000 effective full power hours using - TIP traverse data. - g g V ("iric""O. IM Amendi6ent No. 42,133,138,h 40
PNPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES FRIMARY CONTAINMENT ISOLATION Operable Instrument Channels Per Trip System (1) Minimum Available Instrument Trip Level Settinn Action (2) 2(7) 2 Reactor Low Water Level 29" indicated level (3) A and D 2 1 1 Reactor liigh Pressure $110 psig D 2 . 2 Reactor Low-Low Water level at or above -49 in. A - indicated level (4) 2 2 Reactor fligh Water Level $48" indicated level (5) B i 2(7) 2 liigh Drywell Pressure $2.5 psig A 2 2 liigh Radiation Main Steam $7 times normal rated B Line Tunnel (9) full power background 2 2 Low Pressure Main Steam Line 2810 psig (8) B 2(6) 2 High Flow Main Steam Line $ f rated steam flow B 2 2 Main Steam Line Tunnel IN Exhaust Duct High Temperature $170 F B 1 2 2 Turbine Basement Exhaust Duct High Temperature $150 F B 1 1 Reactor Cleanup System High Flow $300% of rated flow C 2 2 Reactor Cleanup System High Temperature $150 F G C, u a m. 16,\ ( i Amen 86, 147, 45
, ,.3 i
_ _ _ _ _ _ _ _ _ - - _ _ - . _ _ _ _ _ _ . - _ _ . . - - _ _ - _ - - _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ , ._. _ - - _ . _ . _ _ _ - - _ _ . -. _ _ - . _ . - - . - _ - _ . - . _ . . _ . . _ ~ . - - _ _ _ _ _ . -
i [137#$b
- 3. Instrument set point corresponds to (yO inches above top of active fuel.
~
i
- 4. Instrument set point corresponds to 77.26 inches above top of active fuel.
]
- 5. Not required in Run Mode (bypassed by Mode Switch).
- 6. Two required for each steam line.
l
- 7. These signals also start SBGTS and initiate secondary containment isolation.
- 8. Only required in Run Mode (interlocked with Mode Switch).
- 9. Within 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background ,
radiation level shall be determined and associated trip setpoints shall be ; set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at i reactor power levels below 20% rated power. { l l l l i I
*4 m
(' (3iid5IeiiOloTmsson440-A v
-M" 46a I
m
) ,
I ( ,) PNPS TABLE 3.2.B INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trip. System (1) Ti lp Function Trip Level Setting Remarks 2 Reactor Low-Low Hater at or above -49 in. 1. In conjunction with low Level indicated level (4) Reactor Pressure, initiates Core ' spi ay and LP_CI m
^< l
- 2. Inconjunctionw) gl 4 '.( ')
h Orywell Pressur - f-h ( ({/ s..
, second time delay an LPCI or Core Spray pump interlock initiates Auto Blowdown (ADS).
- 3. Initiates HPCI RCIC.
- 4. Initiates starting of Diesel Generators.
~ ~ ~ ~ ~ ' '
2 '- Reactor High Hater Level 948" indicated Trips HPCI and RCIC turbines. level 1 -Reactor Low Level 1307bbove-vessh Prevents inadvertent operation A - (Inside shroud) ,, woffapproximately during I fgf jore' height)' of containment accident conditlon. spray (Indicaw anu "zwh_ A ct m cae cong)
~
'{J 2 Containment High Pressure _1 JJ p < 2)psig Prevents inadvertent operation
,A of containment spr ay during accident condition.
I 5g m 1.s x -a u w ,r 4ad nmL16 nL a hag
^ 5 s~ J k u e $ ct b Q .
Amenom..n i Ho. .y 47 Y._______.____-_---_-_-_______ _- _ _ - - - - - - - - - - _ - - - -
( e PNPS TABLE 3.2.B (Cont'd) Minimtun es of INSTRUMENTATION TilAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Operable Instrument Channels Per Trip System (1) Trip Function Trip Level Settinn Remarks 2 liigh Drywell Pressure $2.5 psig 1. Initiates Core Spray; LPCl ;* IIPCI .
- 2. In conjunction wi Low-Low Reactor p- Water Level, 3 second time delay and f
, //
b'g A TCI or Core Spray pump running, initiates Auto Blowdown (ADS)
- 3. Initiates starting of Diesel Generators
- 4. In conjunction with Reactor Low Pressure initiates closure of liPCI vacuum breaker containment isolation valves.
S 1 Reactor Low Pressure 400 psig i Permissive for Opening Core Spray and LPCI Admission valver. 1 Reactor Low Pressure $110 psig In conjunction with PCIS signal permits closure of RllR (LPCI) injection valves. 3_ 1 Reactor Low Pressure 400 psig i In conjunction with Low-Low React or Water Level initiates Core Spray and LPCI. 5 2 Reactor Low Pressure 900 psig i Prevents actuation of LPCI break detection circuit. f ~ -~~ ~ - 2 Reactor Low Pressure (100>P450-psigg
- - / Isolates HPCI and in conjunction wit h High &q Drywell Pressure initiates closure of IIPCI fB C />St&f 6' vacumi breaker containment isolation valves.
Qtevis16n'l'G7%. ,_ AAen'dinetit No. 42;-113, IE) 48
yh b. f:%.
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.g
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PNPS TABLE 3.2.B (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trio System (1) Trio Function Trio Level Settino Remarks 1 Core Spray Pump Start 2' Initiates sequential starting Timer t<1 sec. of CSCS pumps on any auto 1 .LPCI Pump Start Timer W4 t @ TEE ~~ f, ,,j. start. 1 LPCI Pump Start Timer q.rS t<0") sec. u s. s-1 Auto Blowdown Timer In conjunction with Low Low 1@fiQ.sec. ev.y sis.4 Reactor Hater Level, High Drywell Pressure and LPCI or Core Spray Pump running interlock, initiates Auto Blowdown. 2 ADS Drywell Pressure Q1 2 mi f Permits starting CS and LPCI l Bypass Timer 9 3 g , jf, y g pumps and actuating ADS SRV's if RPV water level is low and drywell pressure is not high. 2 RHR (LPCI) Pump Discharge 150 10 psig Defers ADS actuation pending Pressure Interlock confirmation of Low Pressure core cooling system operation. 2 Core Spray Pump Discharge 150 10 psig (LPCI or Core Spray Pump i Pressure Interlock running interlock.) l 2 Emergency Bus Voltage 20-257. of rated 1. Permits closure of the Relay- voltage resets Diesel Generator to an at less than 507. unloaded emergency bus. Permits starting of CSCS } 2. ' [c .) 6. 4 kV motors.
%d AmendmentNo.AB,106,[/f 49
____m._ - _ _ _ _ _ ______._ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - . . . . . - . . + - - .
- _ . . . - - - . . . - - - - - - + - -e ,e c -----*.-se---
_ - - - - - -.+e- _ _ - - - --. . . - . _ _ _ e _ m. ._____.___e____. -*e
d~' f3
%v 'Mdyp3 f'<
L@!. ' l TABLE 3.2.8 (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of i Operable Instrument Channels Per Trio System (1) Trio Function Trio level Settino Remarks Startup Transformer At 0 Volts _between 1. Trips Startup Transformer 1 2 Loss of Voltage 61-to+P.95econd s to Emergency Bus Breaker. Time Delay
- 2. Locks out automatic closure of Startup Transformer to Emergency O 9d f f f /, 3Y Bus.
- 3. Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
- 4. Prevents simultaneous f starting of CSCS components.
- 5. Starts load shedding logic r
for Diesel Operation in conjunction with (a) Low Low ! Reactor Hater Level and Low Reactor Pressure or (b) High drywell pressure or (c) Core Standby Cooling System components in service in j-conjunction with Auxiliary Transformer , breaker open. ,
.i 4
f hedik.
i c t ( PNPS TABLE 3.2.B (Cont'd)
. INSTRUMENTATION THAT INITIATES ORJ ONTRQLS THE CORE AND CONTAINMENT CROLINS lYSTEMS Minimum # of Operable Instrument Channels Per Trip _Sys_ tem (1) Trio Function Trio level Settin R_emarks 2 Startup Transformer 0.57. with 1. Trips Startup Transformer Degraded Voltage (3868V 9.2 0.5 seconds to Emergency Bus Breaker.
( i t_ime delay
- 2. Locks out automatic 37 78 7 V i o.5/T closure of Startup w; A f e. w f 0. 76 Transformer to Emergency
.5c...d, /,; e 4./ Bus.
7
- 3. Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
- 4. Prevents simultaneous starting of CSCS components.
- 5. Starts load shedding logic for Diesel Operation in conjuction with a) Low Low Reactor Hater Level and Low Reactor Pressure or b) High drywell pressure or c) Core Standby Cooling System components in service in conjuction with Auxiliary Transformer breaker open.
i Revision 123 Amendment No. 42, 6T, 108, 120 50a
~
j f PHPS TABLE 3.2.8.1 INSTRUMENTATION THAT MONITORS EMERGENCY BUS VOLTAGE i Minimum # of l Operable Instrument ! Ch_annels Per Trip System Function S.e.t.tirig Remarkt 1 Emergency 4160V Buses A5 0.57. wit Alerts Operator to possible l
& A6 Degraded Voltage f959V 9.2 0.5 seconds degraded voltage conditions.
Annunication (1) Q1medelay _ Provides permissive to initiate load shedding in j gqg g, S V4 o.S 7o, conjunction with LOCA signal. J
-- c. 2t/fo tjd ic. m.t
' Co % Secends /she de/ay t I (1) In the event that the a,larm system is determined inoperable, commente logging safety-
~
, related bus voltage every 1/2 hour until such time as the alarm is restored to operable
- .ta tus .
M, > hhb; i@ i Amendment No. 42, 61, 108, 53a
! PNPS i l TABLE 3.2.C-2 l CONTROL R0D BLOCK INSTRUMENTATION SETPOINTS l Trio Function Trio Setooint APRM Upscale (1) (2) l APRM Inoperative Not Applicable APRM Downstale 1 2.5 Indicated on Scale Rod Block Monitor (Power Dependent) (1) (3) Rod Block Monitor Inoperative Not Applicable Rod Block Monitor Downstale (1) (3) IRM Downstale 1 5/125 of Full Scale i IRM Detector not in Startup Position Not Applicable IRM Upscale 1 108/125 of Full Scale l IRM Inoperative Not Applicable , SRM Detector not in Startup Position Not Applicable SRM Downstale 1 3 counts /second
.c SRM Upscale 1 10 5counts /second SRM Inoperative Not Applicable Scram Discharge Instrument Volume 1 18 gallons Water Level - High Scram Discharge Instrument Volume - Not Applicable Scram Trip Bypassed Recirculation Flow Converter - Upscale 1 120/125 of Full Scale i
Recirculation Flow Converter - Not Applicable Inoperative go Recirculation Flow Converter - I C Flow Deviation f Comparator Mismatch <-4 4 ~'IFF5We @ N ' 151 Flo M eviation for 1 BO% Aated Pouct - (1) The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT. (2) When the reactor mode switch is in the refuel or startup positions, the APRM rod block trip setpoint shall be less than or equal to 137. of rated thermal power, but always less than the APRM flux scram trip setting.
~
G (3) The RBM bypass time delay (td2) shall be < 2.0 seconds.
,7,N.
h c.-w{ L'h u.u_2__ Amendment No. 42,170,729,722,13)FT
-s 55a
\sf
.g (W.:
PNPS TABLE 4.2.A MINIMUM TEST AND CALIBRATION FREQUENCY FOR PCIS Instrument Channel (51 Instrument functional Test Calibration Frequency Instrument Check
- 1) Reactor liigh Pressure (1) Once/3 months None
- 2) Reactor Low-low Hater Level (1) (7) (7) Once/ day
- 3) Reactor liigh Hater Level (1) (7) (7) Once/ day
- 4) Main Steam liigh Temp. (1) o Once/3 months None
- 5) Main Steam liigh Flow (7) Mo % (7) Once/ day
- 6) Main Steam low Pressure (1) (7) (7) Once/ day
- 7) Reactor Hater Cleanup liigh Flow (1) Once/3 months Once/ day
- 0) Reactor Hater Cleanup liigh Temp. (1) Once/3 months None logic System functional Test (4) (6) Frequency
- 1) Main Steam Line Isolation Vvs. Oncel '7" Main Steam Line Drain Vvs.
Reactor Hater Sample Vvs. Y- f,'
- 2) RilR - Isolation Vv. Control Once Shutdown Cooling Vvs. M llead Spray / d Discharge to Radwaste F
- 3) Reactor Water Cleanup Isolation Once h i @ <
- 4) Drywell Isolation Vvs.
Once/(B-monil% a )-j1 IIP Hithdrawal Atmospheric Control Vvs.
<[) Q [ 'l i
Sump Drain Valves 4 f
- 5) Standby Gas Treatment System Reactor Building Isolation Once M # ~
)
sau m w Amerliinierit ~ Ho.~ 10/ ,G( tM i r,n _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ ___________________m- _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ ___m .- +
.1) '
PI4Ps TABLE 4.2.B HINIMUM TEST AND CALIBRATION FREQUENCY FOR CSCS Instrument Channel Instrument functional Test Calibration Frequency
- 1) Reactor Water Level Instrument Check (1) (7) (7) Once/ day
- 2) Drywell Pressure (1) (7) (7) Once/ day
- 3) Reactor Pressure (1) (7) (7) Once/ day
- 4) Auto Sequencing Timers NA i
Once/ Operating Cycle None
- 5) ADS - LPCI or CS Pump Disch.
Pressure Interlock (1) Once/3 months None
- 6) Start-up Transf. (4160V)
- a. Loss of Voltage Relays Monthly Once/ Operating Cycle None
- b. Degraded Voltage Relays Monthly Once/ Operating Cycle None
- 7) Trip System Bus Power Monitors Once/ operating cycle NA Once/ day 4 8) Recirculation System d/p (1) Once/3 months Once/ day m.
- 9) Core Spray Sparger d/p NA hc Once/ day
- 10) Steam Line High Flow (HPCI & RCIC) (1)
Once/3 month [gf6## None
- II) Steam Line High Temp. (HPCI & RCIC) (1) Once/3 months None
- 12) Safeguards Area High Temp. (1) Once/3 months None
- 13) RCIC Steam Line low Pressure (1) Once/3 months None h
- 14) HPCI Suction Tank Levels (1) % ..
' Once/3 months None
- 15) Emergency 4160V Buses A5 & A6 Monthly loss of Voltage Relays Once/ Operating Cycle None
^
c:Wehfer Em'endiiient"31o.42 99, M 1 7 61
- = . - = . - _ . - - . - - . - -
%.. % ,f '
PNPS TABLE 4.2.0 MINIMUH TEST AND CAllBRATION FRE00ENCY FOR CSCS f Loalcal Sy_ stem Functional Test (4) (6) Frequency Remarks
- 1) Core Spray Subsystem Once @ - 'Q h ,
- 2) Low Press. Coolant Injection Subsystem Once/ -'
i '
- 3) Containment Spray Subsystem Once e
/ '
- 4) HPCI Subsystem Once/ % .h{
- 5) IIPCI Subsystem Auto Isolation Once f ,
- 6) ADS Subsystem Once/ f-
- 7) RCIC Subsystem Auto Isolation Once k
- 8) Diesel Generator Initiation Once M gsi j
- 9) Area Cooling for Safeguard System Once 1 I
f i i i i 4 , h m.evision .Nor::14 7 { AmendmentNo.4f,y),s f2 t i - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
bw (hei#21 . PNPS TAllLE 4.2.C MINIMUM TEST AND CALIBFATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATION Inst rument Chatutel Inntrument Functinnal Calibration innt t oment check Test april - Downscale once/3 Mont hs once/3 Months once/ Day O* APIUI - Upscale Once/3 fionths once/3 ffonths once/ hay ' APRM - Inoperative once/3 flonths Not Applicable once/ pay 1I01 - Upscale (2) (3) Startup or Control Shurdown (2) IRf1 - Downscale (2) (3) St ar t up or Control Shut down (2) IIUt - InoperatIvo (2) (3) Not Applicable (2) RBf! - Upscale once/3 Months once/6 Months once/ pay RBit - Downscale 'Once/3 Mont hs once/6 Months once/ pay ' RBM - Inoperative once/3 Month ; Not Applicable once/ Day SRit - Upscale " (2) (3) Startup or Control Shutdown (2) SR11 - Inoperative (2) (3) Not Applicable (2) SPJ1 - Detector Not in Startup Position (2) (3) Not Applicable (2) SRM - Downscale (2) (3) Startup or Control Shutdown (2) IIUt - De t ec to r No t. in Startup Position (2) (3) Not Appilcable (2) Scram Discharge lustrument Volume once/3 Months Re f'ue l Not Applicable Water Level-Ill gh Scram Discharge Instrument once/3 ffont hs Not Applicable
- Not Applicable Volume-Scram Trip Bypassed ,,,--
Recirculation Flow Converter Not Applicable Once[ Cycle Once/ pay Recirculation Flow Converter-Upscale once/3 flonths once/3 Months once/ Day Recirculation Flow Converter-Inoperative Once/3 ffonths Not Appitcable Once/ Day Recirculation Flow Converter-Comparator once/ 3 11ont hs Once/3 Months once/pny P cf , ion Flow Process Instruments Not Applicable Once @# Once/ Day Lonic System Functional Test (4) (6) System Logic Check Once s a 4Revisitm No. -164C., . , . Amentimen t No. I10;-110,(IX// w? 63
7 L ,J , l PNPS TABLE 4.2.0 tilNIMUM TEST AND CALIBRATION FREQUENCY FOR RADIATION MONITORING SYSTEMS Instrument Channels Instrument functional Calibration initrument Che_c1 (2) Test
- 1) Refuel Area Exhaust Monitors - Upscale (1) Once/3 months Once/ day
- 2) Refuel Area Exhaust Monitors' - Downstale (1) Once/3 months Once/ day Loaic System Functional Test (4) (6) Frequ_e_ncJ
- 1) Reactor Building Isolation Once(B-men %- d' ;
- 2) Standby Gas Treatment System Actuation Once/QI manths-)<-T, '
<;Gvisian Amenilment No.ilovis[.Pdi , t30,'> - fel
BASE l ;. 3.2 In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which ' initiates action to mitigate the consequences of accidents which are ' beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance. l When necessary, one channel may be made inoperable for brief intervals l to conduct required functional tests and calibrations. , Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the l high and low values are both critical and may have a substantial effect ; I on safety. The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations. Actuation of primary containment valves is initiated by protective i instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available
,) whenever primary containment integrity is required.
The instrumentation which initiates primary system isolation is I connected in a dual bus arrangement. The low water level instrumentation 6tc trip at 128.25 Sches abRiD ' 1 we tan nf +he active ?ucpcloses all isolation valves except those in E l Groups 1, 4 and 5. This trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second h I valve closing time. Requ' ed closing times are less than this. Th_e low low reactor water level instrumentationfi: :et +n +rin when rentchleve! is '".26 "ches above the icp of thE'abtike fi F M9" on the hstru r t) TH ; g oses Main Steam Line Isolation
-/
& (h Jq f y
\
I l s .;icr @ E Amendment No. 105, 68
3.2 BASES (Cont'd) P alves, Main Steam Drain Valves, Retirc Sampie Valves (Group 1) activates the CSCS subsystems, starts the emergency diesel generators [M3(S pp 6 and trips the recirculation pumps. This trip setting level was chosen Itobehighenoughtopreventspuriousactuationbutlowenoughto I initiate CSCS operation and primary system isolation so that no fuel [( damage will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line (andwiththetripsettinggivenabove,CSCSinitiationandprimary system isolation are initiated in time to meet the above criteria.
}
The high drywell pressure instrumentation is a diverse signal to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation at about the same time as the low low water level instrumentation; thus the results given above are applicable here also. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of Group 1 isolation valves. ' Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel y during a steam line break accident. The primary function of the jM instrumentation is to detect a break in the main steam line. For hes hj ', worst case accident, main setting @-44ggLratsfitMm-qow]in steamwith conjunction linethebreak flow p^ the outside limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures remain approximately 1000*F and release of radioactivity to the environs is well below 10 CFR 100 guidelines.
)
Temperature monitoring instrumentation is provided in the main' steam line tunnel and the turbine basement to detect leaks in these areas. Trips are provided on this instrumentation and when exceeded, cause ; closure of isolation valves. The setting of 170*F for the main steam l line tunnel detector is low enough to detect leaks of the order of 5 to / 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation. , High radiation monitors in the main steam line tunnel have been provided to detect gross fuel failure as in the control rod drop acci-1 v , ,.__t % , 0 1 Amendment No. 24,,LL1', 69
4.2 BASES (Cont'd) The automatic pressure relief instrumentation can be considered to be a 1 o'ut of 2 logic system and the discussion above applies also. , The instrumentation which is required for the recirculation pump trip and alternate rod insertion systems incorporate analog transmitters. The ; transmitter calibration frequency is once per refueling outage, which is ' consistent with both the equipment capabilities and the requirements for similar equipment used at Pilgrim. The Trip Unit Calibration and Instrument Functional Test is specified at monthly, which is the same frequency specified for other similar protective devices. An instrument check is specified at once per day; this is considered to be an appropriate frequency, commensurate with the design applications and the fact that the recirculation pump trip and alternate rod insertion systems are backups to existing protective instrumentation. ; 1 Control Rod Block and PCIS instrumentation common to RPS instrumentation l have surveillance intervals and maintenance outage times selected in - accordance with NEDC-30851P-A, Supplements 1 and 2 as approved by the NRC l and documented in SERs (letters to D. N. Grace from C. E. Rossi dated ' l September 22, 1988 and January 6, 1989). - '
/ ~g 24
_,e A logic system functional test interval of ^ months was selected to minimize the frequency of safety system inoperability due to testing and to minimize , the potential for inadvertent safety system trips and their attendant ! transients. mascd'cr 19~dustry er.pseience end' BUR 5tansarn TnchniFii l 4pect acations. er 38 month testing interval prnvides adequm+o accoranen nf mperabil4ty he this equipment. - _J l 1 i l _ _ (h;disi3nMc.15k_,
~
E Amendment Ro7~iI,121,130, y47 77 l
LIMITING CONDITIONS FOR OPERATTON SURVEllLANCE REOUIREMENTS 3.6.C.2 Leakaae Detection Systemi /i)4.6.C.2 Leakace Detection Systems (Cont'd) ; t-II (Cont'd) l
- 2. Onechannelofadrywelllis ei
- 2. An instrument channel 'I' .
atmospheric particulate I' calibration at lea t (I radioactivity monitoring system, or j i once per-C,fynM cy Q
%s Al' {
1' [g b. Foreachrehahedg,drywell E\
- 3. One channel of a drywell J I atmospheric radioactivity il atmospheric gaseous '
radioactivity monitoring l monitoring system perform: '1 x 1 system. lg 1. An instrument check at 't iS least once per day, ji
- b. 1. At least one drywell -
l }! i sump monitoring system l 2. An instrument shall be Operable; otherwise, be in Hot If functional test at
/'1 lN least once per 31 Li Shutdown within the next I days, and /i /
12 hours and in Cold l
.[
Shutdown within the .I 3. following 24 hours. An instrument channel i calibration at least
- 2. At least one gaseous or .Il (I once per @ T M s] > th-i m- -
I particulate I radioactivity monitoring l I C ) * '*%' ' . c y r; channel must be '-{ Operable; otherwise, I reactor operation may ji continue for up to 31 11 days provided grab ;l i samples are obtained and analyzedevery24 hours,iti ,lQ or be in Hot Shutdown l within the next 12 hours ll and in Cold Shutdown !! within the following 24 ll hours. yl
!I :
- c. With no required leakage ;[ l detection systems Operable, il l be in Cold Shutdown within i-4 t 24 hours.
( l
~- ,
?i n 4 y Amendment No. d3t y 125b c
l
_ _ ~ __ . _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _m. _ _ _
- L!MITIN3 CONDITIONS FOR OPERATION SURVEILLANCE REOb!REMENTS i
3.6.I snock Suppressors (Snubters) 4.6.'I Shock Suppressors'(50UDDe's)
' ~ l 1 During all meces of operation The following surveillance l e= cept Cold Shutdown and Refuel, requirements apply to all safety l 4 .
all safety-related snubbers related nydraulic and metnanical ! j 'f
, listed in PNPS Procedures shall snubbers listed in PNPS Procedures. l 1 De operacle except as noted l l /
V in 3.6.I.2 through 3.6.I.3 below. The required visual inspection- ' i interval varies inversely with An Inoperable Snubber is a the observed cumulative numcer properly fabricated. installed of inoperable snubbers founc ^ 4 and sized snubber which cannot during an inspection. 1 pass its functional test. Inspections performed before that interval has elapsed may be l l Upon determination that a snubber used as a new reference point to l is either improperly fabricated, determine the next inspection. ' installed or sized, the However, the results of such corrective action will be as early inspections performed i specified for an inoperable before the original time ; snubber in Section 3.6.I.2. - interval has elapsed may not be l used to lengthen the required !
- 2. From and after the time that a interval.
1 snubber is determined to be f inoperable, replace or repair Number of snubbers found j the snubber during the next 72 inoperable during inspection or , hours, and initiate an engineering during inspection interval: l evaluation to determine if the components supported by the Subsequent -
, snubber (s) were adversely affected Inoperable Visual Inspec-by the inoperability of the Snubbers tion Interval !
snubbers and to ensure that the supported component remains- 0 '2.V Months + 257. . capable of meeting its intended I g _.,. Monthsi257 function in the specific safety 2 p. Months 251 system involved. 3,4 m/ grl@b i25% ; 5,6,7 m/-Qy Days + 25% Further corrective action for '/ p;mg$ F-Q Days i25% this snubber, and all generically jo o ne ce. 3, ogy3 y 2.r fo susceptible snubbers, shall be .The required-inspection interval determined by an engineering shall not be lengthened more evaluation. than one step at a time.
. 3. From and after the time a snubber Snubbers may be categorized in
- is determined to be inoperable, two groups, " accessible" or improperly fabricated improperly " inaccessible" based on their installed or improperly sized, if accessibility for inspection the requirements of Section(s) during reactor operation. These 3.6.I.1 and 3.6.I.2 cannot be met, two groups may be inspected i then the affected safety system, independently according to the or affected portions of that above schedule, system, shall be declared inoperable, and the limiting 1. Visual Inspection Acceptance condition for t~atn system Criteria
~ entered, as appropriate. A. Visual inspections shall verify: 0 % R *9 P
- NO N, hn. Q
l L:v! :NG CCND: TIONS FOR CDERatICN _
$URVEILLANCE REOU p[u[vg _ _ _ .
3.6.! Shoct Suopressors (Snubbers) 4.6.1 Shock Suppressors (Snuccer.i l 4 Snubbers may be addec to, or 1 That there are no visitle removed from, per 10 CFR 50.59, indications of damage or safety related systems witnout impaired coerability. j,P prior NRC apprCval. the addition
) or deletion of snubbers shall be 2. Attachments to the i,P"k recorted to the NRC in accordance foundation or support fi - with 10 CFR 50.59 structure are sucn inat tne D functional capability of the snubber is not suspect.
I i B. Snubbers which appear ! INOPERABLE as a result of l visual inspections may be ! determined OPERABLE for tne purpose of establishing the next visual inspection interval l , provided that:
- 1. The cause of the rejection l
is clearly established and remedied for that particular snubber, and
- 2. The affected snubber is functionally tested, when necessary, in the as found N- condition and determined OPERABLE per specifications 4.6.I.2.B., 4.6.I.2.C., as applicable.
C-@#For any snubber determined inoperable per specification 4.6.I.2 clearly establish the cause of rejection and remedy the problem for that snubber, , and any generically l susceptible snubber. !
- 2. Functional Tests (Hydraulic and Mechanical Snubbers)
A. Schedule Atleagoncepergrating c cyc e g..~muy a 12 representative sample ( .Y of the total of each type: hydraulic, mechanical) of snubbers in use in the plant shall be functionally tested, either in place or in a bench test. For each snubber that does not meet the functional test
,r. , acceDtance criteria of
LIMITING CONDITION FOR OPERATION SL7NEILIJSCE REQUIREMENT ,
.[34.6.I Shock Suppressors (Snubbers)
\) l b,/; Specificarica 4.6.I.2.B. or
,t t, 4.6.I.2.C. applicable, an h,/ adcitiona %)fthattypeof
) snubber sh be functionally j ! tested. , l g , g e7e
/ / )
'/ / B. General Snubber Tunctional Y Test Acceptance Criteria I,
, f } (Hydraulic and Mechanical)
) y The general snubber functional h/
( h) test shall verify that: i (; 1. Activation (restrain-
'I
) /( ing action) is achieved j# ) within the specified l l range of velocity or l / acceleration in both l,
I h'l / tension and compression. l l (' 2. Snubber release, or bleed-l ( rate, as applicable, i i where requiredt is wit". sin D d/ [) the specified range in conpression or tension. (/ ) For snubbers specifi-(' /{ cally required not to 3
)
[ displace under contin-
/ I uous load, the ability
- (, '
of the snubber to with-stand load without dis-(I .
, placement shall be veri-l fied.
.h, n
; O! (' C. Mechanical Snubbers Functic 1 Test Acceptance Criteria
( i ,
/
i i/ The mechanical snubber
/ functional test shall
[{/ ,)
\
verify that: i 1.
',l (j/ k The force that initiates v
/ free movement of the
/ snubber rod in either
, /;(/)! ,)}(
~
tension or compression is less than the specified Il / maxietun drag force. Drag
;} ./) force shall not have ine-
/ i reased more than 50% since
/ ,f .(/ ( the last functional test.
lt! Snubber Service I,t fe tonitoring -w j.3
'j A. d record of the service life
,c 7 Q).
kendment tb.(60- ( 137c
l . LIMIT 1t4G C0tJDIT10NS FOR OPERAT10t1 SURVElttAt4CE RE001 REMET 4TS 3.7.B Standby Gas Treatment System 4.7.B Standby Gas Treatment System i l . and Control Room M .g.,p g h and Control Room Hiah - i Efficiency Air Fi W ation Efficiency Air Filtration i System System j
- 1. Standby Gas Treatment System Standby Gas Treatment System l Except as specified in b ~1. ,-- -
I
- a. l 3.7.B.I.c below, both trains 4cJ^ 9 a.(1.) C,.deAt least oncbe k M T it shall i of the standby gas treatment of demonstrated that i system and the diesel pressure drop across j generators required for the combined high i operation of such trains efficiency filters and !
shall be operable at all charcoal adsorber ! times when secondary banks is less than 8 l containment integrity is inches of water at ! required or the reactor shall 4000 cfm. ; be shutdown in 36 hours. ! (2. At least oncec6ve m @ -C ,
- b. (1.) The results of the in- f d5TiiD demonstrate place cold 00P tests on .
/ that the inlet heaters i HEPA filters shall show on each train are l 299% DOP removal. The operable and are !
results of halogenated. capable of an output i hydrocarbon tests on of at least 14 kW. l charcoal adsorber banks shall show 199% (3.) e tests and analysis j halogenated hydrocarbon o Specification i removal. 3. B.I.b. shall be
- per rmed at least j once me.1 m mun un ;
(2.) The results of the l laboratory carbon sample or following painting, j analysis shall show 295% fire or chemical methyl iodide removal at release in any I a velocity within 10% of ventilation zone ! systemdgsign,0.5to communicating with the i 1.5 mg/m inlet methyl system while the ' iodide concentration, system is operating 270% R.H. and 1190 F. that could contaminate i The analysis results are the HEPA filters or to be verified as charcoal adsorbers. acceptable within 31 __ days after sample (4.) "At 1 ast once @/9y ' Q removal, or declare that th , automatTc-train inoperable and initiation of each take the actions branch of the standby specified 3.7.B.I.c. gas treatment system shali be demonstrated,
' with Specification 1 3.7.B.I.d satisfied.
y
/ ,
h /
/
C! s
/
/
Gevi:;icr M" ( % AmendinEntY 50,51,52,112,h, 158
"' ' % 0'OIT!0!5 70; OPERt.T!Of. 5U;VEILLANCE REQUIREMENTS 3.'.E (Continued) ' 7.b (Continued) l .
y og 4
- 2. ; er.* col Roc- H1on Ef#iclerc, Air 2. Control Room High Efficiencv Ai[
;-! ! ration S n t p- rilt rat ion Systen
[
- e. Excett as specified in At least once b !9
}' a. e specification 3.7.B.2.c the pressure drop across each Delow. both trains o' the combined filter train shall b' Lontrol Room High Ef ficiency demonstrated to be less thap 6 Air Filtration System used inches of water at 1000 cfr/ or for the processing of inlet the calculated equivalert. j air to the control ro0~
- l under accioent conditions b. (1.)Thetestsandana/vsisof and the diesel generator (s) Specification 3/.S.2.b recuired for operation of shall be perfc .ned once each train of the system
' 2v .u. or shall be operable whenever [o 1llowing painting, fire or l secondary containment chemical release in any integrity is required and ventilation zone l during fuel handling communicating with the operations. system while the system is ;
operating.
- b. (1.) The results of the in-place cold DOP tests on (2.) In-place cold DOP testing I HEPA filters shall show shall be performed after 299'; DOP removal. The each complete or partial results of the replacenient of the HEPA halogenated hydrocarbon filter bank or after any tests on charcoal structural maintenance on adsorber banks shall the system housing which show 299% halogenated could affect the HEPA hydrocarbon removal filter bank bypass leakage.
when test results are i extrapolated to the (3.) Halogenated hydrocarbon initiation of the test. testing shall be performed . after each complete or (2.) The results of the partial replacement of the laboratory carbon charcoal adsorber bank or sample analysis shall after any structural show 295"; methyl iodide maintenance on the system removal at a velocity housing which could affect l within 10% of system the charcoal adsorber bank design,0.05to0.15 bypass leakage. mg/m inlet methyl iodide concentration, (4.) Each train shall be 270% R.H., and 2125 F. operated with the heaters The analysis results in automatic for at least are to be verified as 15 minutes every month. acceptable within 31 days after sample (5.) The test and analysis of l removal, or declare Specification 3.7.B.2.b.(2) l that train inoperable shall be performed after and take the actions every 720 hours of system specified in 3.7.B.2.c. operation. r d
- E During RF0 89, one train can be without it s safety-related bus and/or its emergency diesel generator without entering the LC0 action statement f provided the conditions listed on page 158A are met.
l l
?Y W nd ent No. 50. 51. 52. 101. 112. I,d' 15oB
, LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7.B (Continued) 4.7.B (Continued) fu o
- f.
- c. From and after the date that c. At least once M 1, 15 mugK one train of the Control demonstrate that the inlet Room High Efficiency Air heaters on each train are Filtration System is made or operable and capable of an found to be incapable of output of at least 14 kw.
supplying filtered air to the control room for any d. Perform an instrument reason, reactor operation or functional test on the refueling operations are humidistats controlling the permissible only during the heaters once perfjyS ,cnihQ succeeding 7 days providing 4 that within 2 hours all / active components of the &fadq other CRMEAF train shall be d~ l demonstrated operable. If the system is not made fully operable within 7 days, reactor shutdown shall be initiated and the reactor - shall be in cold shutdown within the next 36 hours and l l irradiated fuel handling l operations shall be l terminated within 2 hours. ! Fuel handling operations in progress may be completed.
- d. Fans shall nperate within 110% of 1000 cfm.
i l i i
. l
/(
- During RF0 #9, one train can be without its safety-related bus and/or its c emergency diesel generator without entering the LC0 action statement pa provided the conditions listed on page 158A are met.
i l I i 1 i'icn-i b Amendmeirr'No . 50. 51. 57 H 2.k3 lw ,
MSIS: 3,7. A & 4.7. A Primary Containment Group _6 - process lines are normally in use and it is therefore not desirable
.o cause spurious isolation due to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system. Also, since the vessel could potentially be drained through the cleanup system, a low level isolation is provided.
Grppp_2 - The HPCI vacuum breaker line is designed to remain operable when the ' HPCI system is required. The signals which initiate isolation of the HPCI - vacuum breaker line are indicative of a break inside containment and reactor pressure below that at which HPCI can operate. The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment. In satisfying this design intent an additional margin has been included in specifying maximum closure times. This margin permits identification of degraded valve performance, prior to exceeding the design closure times. In order to assure that the doses that may result from a steam line break do not exceed the 10CFR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding O erforations would be avoided for main steam valve closure times, including
" instrument delay, as long as 10.5 seconds.
These valves are highly reliable, have low service requirements and most are normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolstion. The test interval of once per operating cycle for automatic inith tion results in a . failure probability of 1.1 x 10-' that a line will not '.solate. More frequent testing for valve operability results in a greater a m rance that the valve will be operable when needed. The main ster line isolation valves are functiona.i:' tested on a more frequent interval to e.tablish a high degree of reliiH lity. The primary containmrnt is penetrated by several small diameter instrument i lines connected to tne reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment. A program for ; periodic testing and examination of the excess flow check valves is in place. / 1 i Primary Containment Paintina p l The interiors of the drywell and suppression chamber are painted to prevent . _ rusting. The inspection of the paint during each major refueling outage ' Q pro e tel y every r ~enthR) assures the paint is intact. Experience with this type of paint GMoH4+61M-genenting-stati6hhindicates that the (h inspection interval is adequate. SSMgdn AGk >
%&id iR}e Amendment No. ax4 dar &Je]169
BASES: 3.7.B.1 and 4.7.B.1 - Standbv Gas Treatment Sntem l The Standby Gas Treatment System is designed to filter and exhaust the reactor '
'uilding atmosphere to the stack during secondary containment isolation l onditions. Upon containment isolation, both standby gas treatment fans are designed to start ,t'o b6ing the reactor building pressure negative so that all leakage should bef i n-leakage. After a preset time delay, the standby fan ,
automaticallyshutsgwnsothereactorbuildingpressureismaintained ' approximately 1/4 inch of water negative. Should one system fail to start, the redundant system is designed to start automatically. Each of the two i trains has 100% capacity. l High Efficiency Pa-ticulate Air (HEPA) filters are installed before and after i the charcoal adsorcers to minimize potential release of particulates to the l environment ard to prevent clogging of the iodine adsorbers. The charcoal : adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and .j a HEPA filter efficiency of at least 99 percent removal of cold D0P ; particulates. The laboratory carbon sample test results should indicate a- ! methyl iodide removal effici y of at least 95 percent for expected accident ! conditions. The specified ficiencies for the charcoal and particulate : filters is sufficient to p O de exceeding 10 CFR 100 guidelines for,the ; accidents analyzed. The a c/ 1 sis of the loss of coolant accident assumed a !
- charcoal adsorber efficien y of 95% and TID 14844 fission product source i
! terms, hence, installing two banks of adsorbers and filters in'eath train l provides adequate margin. A 14 kW heater maintains relative humidity below j 70% in order to ensure the efficient removal of methyl iodide on the ! impregnated charcoal adsorbers. Considering the re_lative simplicity of the : M eating circuit, the test frequency of once p' r ! -tyisadeu.ateto i
. .emonstrate operability. g_ gt" !
l Air flow through the filters and charcoal adsorbers for 15 minutes each month I assures operability of the system. Since the system heaters are automatically I controlled, the air flowing through the filters and adsorbers will' be 170% } relative humidity and will have the desired drying effect. l; Tests of impregnated charcoal identical to that used in the filters indicate ! that shelf life of five years leads to only minor decreases in methyl iodide - l removal efficiency. Hence, the frequency of laboratory carbon sample analysis 1 is adequate to demonstrate acceptability. Since adsorbers must be be removed' ! to perform this analysis this frequency also minimizes the system out of i l service time as a result of surveillance testing. In addition, although the j i halogenated hydrocarbon testing is basically a leak test, the adsorbers have i charcoal of known efficiency and holding capacity for elemental iodine and/or j methyl iodide, the testing also gives an indication of the relative efficiency of the installed system. The 31 day requirement for the ascertaining of test ' results ensures that the ability of the charcoal to perform its designed function is demonstrated and known in a timely manner. The required Standby Gas Treatment System flow rate is that flow, less than or equal to 4000 CFM which is needed to maintain the Reactor Building at~a 0.25 inch of water negative pressure under calm wind conditions. This capability is adequately demonstrated during Secondary Containment Leak Rate Testing performed pursuant to Technical Specification 4.7.C.1.c. 42,h 172
RiiiS: 3.7.5.1 and 4.7,B.1 ( @ tinu_ed) The test frequencies are adequate to detect equipment deterioration prior to significant defects, but the tests are not frecuent enough to load the filters or adscrbers, thus reducing their reserve capacity too quickly. The filter testing is performed pursuant to appropriate procedures reviewed and approved t, the Operations Revie. Committee pursuant to Section 6 of these Technical Specifications. The in-place testing of charcoal filters is performed by injecting a halogenated hydrocarbon into the system upstream of the charcoal adsorbers. Measurements of the concentration upstream and downstream are mace. The ratio of the inlet and outlet concentrations gives an overall indication of the leak tightness of the system. A similar procedure substituting dioctyl phthalate for halogenated hydrocarbon is used to test the iEPA filters. Pressure drop tests across filter and adsorber banks are performed to detect plugging or leak paths though the filter or adsorber media. Considering the relatively short times the fans will be run f r test purposes, plugging is unlikely and the test interval of once per W .J c nthdis reasonable. m s- g ns anc housing gasket doors are desig'ned such le age Systemdr[iryleakagefromtheStandbyGasTreatmentSystemRoom. wouldbe[ , This ensures that thete vill b2 no bypass of process air around the filters or adsorbers. Only one of the t*, Standby Gas Treatment Systems (SBGTS) is needed to maintain the set W ry containment at a 0.25 inch of water negative pressure upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling activities may continue while repairs are being made. In thi event one SBGTS is inoperable, the redundant system's active components will be tested within 2 hours. This substantiates the availability of the I-operable system and justifies continued reactor or refueling operations. If both trains of SBGTS are inoperable, the plant is brought to a condition snere the SBGTS is not required. l l
~n ,-
sion ii5; ^ m l Amen ment No. 42,l<M2, 173 1 w
..r n m (h
dj ) TABLE 4.8-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,
,, v Channel l '
Instrument Source Channel functional I, / Instrument Calibration Check Check Test h / r
- 1. Gross Beta or Gamma Radioactivity ,
Monitors Providing Alarm and Auto- f f , matic Isolation /
' /
- a. Liquid Radwaste Effluents Line NA Once per Quarterly /
J8Mo.S. / i
- 2. Flow Rate Measurement Devices /
- a. Liquid Radwaste Effluent Line NA Once per Quarterly /
*~ /8Mos, )' /
f I
/
/
f W
/
"D'uring or prior to release via this pathway. -
/
'Previously established calibration procedures will be used for these requirements. i i
.j Amendment No. 190 i
t . w l TABLE 4.8-4 RADIOACTIVE GASEOUS EFFLUENT HONITORING_ INSTRUMENTATION SURVEll. LANCE qEQUIREMENTS 1 / Instrument - Instrument Source Instrument Functional ' Instrument Check Check Calibration Test /t I. Main Stack Effluent Monitoring System !
- a. Noble Gas Activity Monitor Daily' Monthly Once per_ Quarterly ,
(Two channels) _eperatth-T , l0M0S g -
- b. Iodine Sampler Cartridge NA NA NA NA
/
- c. Particulate Sampler Filter NA NA NA NA l d. Effluent System Flow Rate Dally' NA Quarterly '
- Heasuring Device g ra i ... - -
Mos,
- e. Sampler flow Rate Measuring Dally' NA Quarterly i
Once 4 Dnice gg i
- 2. Reactor Building Ventilation W/8405 ' ,
. !/ ,,
Effluent Monitoring System / l a. Noble Gas Activity Monitor Daily' Monthly Once. per -Quarterly , 5
~ 44"1 / ,
LY / l8gOS, ,.
- b. Iodine Sampler Cartridge NA NA NA NA /
J
! . i jf'J
- q. c. Particulate Sampler Filter NA NA NA NA ,f
; d. Effluent System Flow Rate Dally' NA Once per Quarterly ' -
Heasuring Device ^ *; e n t ! n ' I' , l / ' 4
~ muss. -
Amendment No. 193 4
i I i b v' TADLE 4.8-4 (Continued) \ RADIOACTIVE GASEOUS EFFLUENT HONITORING_ INSTRUMENTATION SURVElllANCE REQUIREMENTS [ v ! Instrument , Instrument Source Instrument Functional C Instrument Check Check Calibration Test '
- e. Sampler Flow Rate Measuring Dally' s 'k NA Once per Quarterly -
Device
/0N05. -
- 3. Steam Jet Air Ejector Radioactivity ,
/
Monitor
- a. Noble Gas Activity Monitor Daily' NA Once per Quarterly e operating ,
cycle * '
/
- 4. Augmented Offgas Treatment /
System Explosive G's a Monitoring , System '
- a. Hydrogen Monitor Dally' NA Quarterly' Monthly ,
/ '
/
/'
'During releases via this pathway. f
/
'During augmented offgas treatment system operation.
'During operation of the steam jet air ejector. /
'Previously established calibration procedures will be used for these requirements.
/
' Calibrate at 2 points with standard gas samples differing by at least 11 but not exceeding 41. /
/
4 b a a
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREHENTS
;, 4.9.A Auxiliary Electrical Eauipment y Surveillance (Cont'd) @
{l; i '
- 1. Verifying de-energization of the emergency buses and load shedding from the emergency buses.
- 2. Verifying the diesel starts i from ambient condition on the auto-start signal, @
energizes the emergency buses with permanently connected loads, energizes the auto-connected emergency loads through the load f sequence, and operates for @ 1 5 minutes while its generator is loaded with the emergency loads. During performance of this surveillance verify that HPCI and RCIC inverters do not trip. /: The results shall be logged.
- c. Once per operating cycle with the diesel loaded per 4.9.A.l.b l verify that on diesel generator !
trip, secondary (offsite) AC ' power is aut titally connected within to seconds to the ! em ency vice buses and merge loads are energized
//. 6 thr gh the load sequencer in e same manner as described in ,
4.9.A.l.b.l. j f3 t The results shall be logged. ; 1 in [ gg AmendmentNo.42,67,/// 194)(e l
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9.B Operation with Inoperable 4.9.A Auxiliary Electrical Eauipment Eauipment Surveillance (Cont'd) Whenever the reactor is in Run 3. Emergency 4160V Buses AS-A6 g@ Mode or Startup Mode with the reactor not in a Cold Condition, Degraded Voltage Annunciation System. the availability of electric power shall be as specified in a. Once each operating cycle, 3.9.B.1, 3.9.B.2, 3.9.B.3, calibrate the alarm sensor. 3.9.B.4, and 3.9.B.5.
- b. Once each 31 days perform a
- 1. From and after the date that channel functional test on the incoming power is not available alarm system. ,
from the startup or shutdown transformer, continued reactor c. In the event the alarm system operation is permissible under is determined inoperable under this condition for seven days. 3.b above, commence logging During this period, both diesel safety related bus voltage generators and associated every 30 minutes until such l emergency buses must be time as the alarm is restored demonstrated to be operable. to operable status.
- 2. From and after the date that 4. RPS Electrical Protection incoming power is not available Assemblies from both startup and shutdown transformers, continued a. Each pair of redundant RPS operation is permissible, EPAs shall be determined to provided both diesel generators be operable at least once e and associated emergency buses per 6 months by performance are demonstrated to be operable, of an instrument functional 4'5 4- all core and containment cooling test.
sysi m are operable, reactor /g mJs power level is reduced to 25% of b. Once per 60eratina cycle, design and the NRC is notified each pair of redundant RPS within one (1) hour as required EPAs shall be determined to by 10CFR50.72. be operable by performance l of an instrument ! 3. From and after the date that one calibration and by l of the diesel generators or verifying tripping of the j associated emergency bus is made circuit breakers upon the or found to be inoperable for simulated conditions for any reason, continued reactor automatic actuation of the operation is permissible in protective relays within accordance with Specification the following limits: 3.5.F if Specification 3.9.A.1 and 3.9.A.2.a are satisfied. Overvoltage 5 132 volts Undervoltage 2 108 volts
- 4. From and after the date that one Underfrequency 2 57Hz of the diesel generators or l
! associated emergency buses and j either the shutdown or startup transformer power source are i made fr h
& (_?R jav4 fon=163 7 A Tmendment No. 88,127(14h 196 l )
l l
l LfMfTING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS I 3.9.B Ooeration with Inocerable Ecuioment (Cont'd) or found to be inoperable for any l
, reason, continued reactor ' / operation is permissible in / accordance with Specification j 3.5.F. provided either of the L/ following conditions are satisfied:
I
- a. The startup transformer and both offsite 345 kV transmission lines are available and capable of i automatically supplying auxiliary power to the emergency 4160 volt buses.
- b. A transmission line and associated shutdown #
transformer are available and capable of automatically supplying auxiliary power to the emergency 4160 volt buses. ' 5. From and after the date that one of the 125 or 250 volt battery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding three days within electrical safety considerations, provided repair work is initiated in the most expeditious manner to : return the failed component to an operable state, and Specification 3.5.F is satisfied. 39 gg, g
- 6. With the emer ncy bus voltag M67 i less than V but above V !
(excluding transients) during normal operation, transfer the ) safety related buses to the diesel generators. If grid voltage continues to degrade be in at least Hot Shutdown within the next 4 hours and in Cold Shutdown within the ' following 12 hours unless the < grid conditions improve. I l v m1 %f f 8 Amendment No. 42,67.88,120,h 197 l
BASE 6: (Cont'd) l 4.9 deliver full flow. Periodic testing of the various components, plus a functional test once per cycle, is sufficient to maintain adequate reliability. Although station batteries will deteriorate with time, utility experience ' indicates there is almost no possibility of precipitous failure. The type of surveillance described in this specification has been demonstrated over the years to provide an indication of a cell becoming irregular or unserviceable long before it becomes a failure. The Service Discharge Test provides indication of the batteries' ability to satisfy the design requirements (battery duty cycle) of the associated dc system. This test will be performed using simulated or actual loads at the rates and for the duration specif.ed in the design load profile. A once per cycle testing interval was chosen to coincide with planned outages. l l The Performance Discharge Test provides adequate indication and assurance that ! the batteries have the specified ampere hour capacity. The results of these 1 tests will be logged and compared with the manufacturer's recommendations of acceptability. This test is performed once evbry five years in lieu of the Service Discharge test that would normally occur within that time frame. The diesel fuel oil quality must be checked to ensure proper operation of the i diesel generators. Water content should be minimized because water in the ' fuel could contribute to excessive damage to the diesel engine.
' The electrical protection assemblies (EPAs) on the RPS inservice power ' supplies, either two motor generator sets or one motor generator and the alternative supply, consist of protective relays that trip their incorporated circuit breakers on overvoltage, undervoltage, or underfrequency conditions.
There are two EPAs in series per power source. It is necessary to periodically test the relays to ensure the sensor is operating correctly and to ensure the trip unit is operable. Based on experience at conventional and nuclear power plants, a six-month frequency for the channel functional test is established. This frequency is consistent with the Standard Technical Specifications. The EPAs of the power sources to the RPS shall be determined to be operable by .th During calibratica, a transfer to the alternative power) sourc@e ii performance of a channel calibration of the relays once perEq t$9 however, prior to switching to alternative feed, de-energization of the applicable MG set power source must be accomplished. This results in a half scram on the channel being calibrated until the alternative power source is connected and the half scram is cleared. Based on operating experience, drift of the EPA protective relays is not significant. werefore, 10
'(spur 4 css sci ~aimsa c.alibret%iFfrequenciT cn;c [perwycae-is-established. avoid-possible (e 4[4[n Dw ~
1 Am mentfioI 127, 135, 141, 201 v , i
Attachment D to BEco letter 93 156 i i Complete Copy of Amended Technical Specification Paaes Page: 4 l 5a I
- 27 '
1 29 ; 32 j 40 45 - 46a l 47 48 49 < 50 ' 50a ! 53 i 53a ! 55a l 59a ! 60 l 61 ; 62 63 l' 64 68 l 69 77 125b ! 137a !
- 137b l 137c '
158 l 158B l l 158C 169 172 173 190 193 193a 194a 196 197
. 201 ,
i i l I i
1.0 DEFINITIONS (Cont'd)
- 1. At least one door in each access opening is closed. ,
- 2. The standby gas treatment system is operable.
- 3. All automatic ventilation system isolation valves are operable '
or secured in the isolated position. ;
- 0. Operatino Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage. ;
P. Refuelino Frecuencies
- 1. Refuelina Outaae - Refueling outage is the period of time l between the shutdown of the unit prior to a refueling and the (
startup of the plant after that refueling. For the purpose of
- designating frequency of testing and surveillance, a refueling l outage shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous refueling outage, the required surveillance testing i need not be performed until the next regularly scheduled ;
outage (Definitions V and V apply). l
- 2. Refuelina Interval - Refueling interval applies only to ASME !
Code, Section XI IWP and IWV surveillance tests. For the purpose of designating frequency of these code tests, a j refueling interval shall mean at least once every 24 months. ' i Q. Alteration of the Reactor Core - The act of moving any component in ( the region above the core support plate, below the upper grid and within the shroud. Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration. Normal movement of in-core instrumentation is not defined as a core alteration. R. Reactor Vessel Pressure - Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured , by the reactor vessel steam space detectors. S. Thermal Parameters
- 1. Minimum Critical Power Ratio (MCPR) - the value of critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.
- 2. Transition Boilino - Transition boiling means the boiling i regime between nucleate and film boiling. Transition boiling i is the regime in which both nucleate and film boiling occur j intermittently with neither type being completely st:bla.
- 3. Total Peakina Factor - The ratio of the fuel rod surface heat flux to the heat flux of an average rod in an identical geometry fuel assembly operating at the core average bundle power.
Amendment No. 15 149 4 I i
i l l 1.0 DEFINITIONS (Continued) i U. Surveillance Freauency - Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval. The Surveillance Frequency establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of Definition "U" is based on engineering judgment and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval. V. Surveillance Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the instrument, component, or system is not required to be operable, but the instrument, component, or system shall be tested prior to being declared operable. The operating cycle interval is 24 months and l the 25% tolerance given in Definition "U" is applicable. The refueling interval is 24 months and the 25% tolerance specified in definition "U" is applicable. W. Fire Suppression Water System - A fire suppression water system shall consist of: a water source (s); gravity tank (s) or pump (s); and distribution piping with associated sectionalizing control or isolation valves. Such valves shall include hydrant post indicator valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser. X. Staqqered Test Basis - A staggered test basis shall consist of: (a) a test schedule for n systems, subsystems, trains, or otner i designated components obtained by dividing the specified test ! interval into n equal subintervals; (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval. Y. Source Check - A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. I Amendment No. 42, 89, 128, 149 Sa u
PNPS Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Operable Inst. Modes in Which Function Channels per Trip Function Trip Level Setting Must Be Operable Action (I) Trip System (1) Refuel (7) Startup/ Hot Run MinimumlAvail. Standby 1 1 Mode Switch in Shutdown X X X A 1 1 Manual Scram X X X A IRM 3 4 High Flux 5120/125 of full scale X X (5) A 3 4 Inoperative X X (5) A APRM 2 3 High Flux (15) (17) (17) X A or B 2 3 Inoperative (13) X X(9) X A or B 2 3 High Flux (15%) 515% of Design Power X X (16) A or B 2 2 High Reactor Pressure $1063.5 psig X(10) X X A 2 2 High Drywell Pressure $2.22 psig X(8) X(8) X A 2 2 Reactor low Water Level 211.7 In. Indicated Level X X X A SDIV High Water Level: 138 Gallons X(2) X X A 2 2 East 2 2 West 2 2 Main Condenser low Vacuum 223 In. Hg Vacuum X(3) X(3) X A or C 2 2 Main Steam Line High 57x Normal Full Power Radiation Background (18) X X X(18) A or C j 4 4 Main Steam Line Isolation Valve Closure $10% Valve Closure X(3)(6) X(3)(6) X(6) A or C 2 2 Turbine Control Valve 2150 psig Control Oil Fast Closure Pressure at Acceleration Relay X(4) X(4) X(4) A or D 4 4 Turbine Stop Valve $10% Valve Closure X(4) X(4) X(4) A or D Closure Amendment No. 15,-421-86 5 -92 5 -117, 133, 147, 27
l l NOTES FOR TABLE 3.1.1 (Cont'd) , 1
- 2. Permissible to bypass, with control rod block, for reactor protection system l reset in refuel and shutdown positions of the reactor mode switch.
- 3. Permissible to bypass when reactor pressure is $576 psig. l
- 4. Permissible to bypass when turbine first stage pressure is 5112 psig. 1
- 5. IRM's are bypassed when APRM's are onscale and the reactor mode switch is in the I run position. l
- 6. The design permits closure of any two lines without a scram being initiated. '
- 7. When the reactor is subcritical, fuel is in the reactor vessel and the reactor water temperature is less than 212 F, only the following trip functions need to i be operable: i A. Mode switch in shutdown B. Manual scram l C. High flux IRM D. Scram discharge volume high level ,
E. APRM (15%) high flux scram
- 8. Not required to be operable when primary containment integrity is not required.
- 9. Not required while performing low power physics tests at atmospheric pressure during or af ter refueling at power levels not to exceed 5 MW(t). i
- 10. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
- 11. Deleted
- 12. Deleted
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per ;
level or there is less than 50% of the normal complement of LPRM's to an APRM. '
- 14. Deleted I
- 15. The APRM high flux trip level setting shall be as specified in the CORE I OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated thermal power.
- 16. The APRM (15%) high flux scram is bypassed when in the run mode.
- 17. The APRM flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes.
- 18. Within 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing control rods at reactor power levels below 20% rated power.
l Amendment No. 6, 15, 27, 42, 86, 117, 118, 133, 147, 29 l l l
TABLE 4.1.2 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CAllBRATION MINIMUM CALIBRATION FRE0VENCIES FOR REACTOR PROTECTION INSTRUMENT CHANNELS Instrument Channel Calibration Test (5) Minimum Freauency (2) IRM High Flux Comparison to APRM on Controlled Note (4) Shutdowns Full Calibration Once per Operating Cycle APRM High Flux Output Signal Heat Balance Once every 3 Days Flow Bias Signal Calibrate Flow Comparator and At least Once Every Flow Bias Network 18 Months Calibrate Flow Bias Signal (1) Every 3 Months LPRM Signal TIP System Traverse Every 1000 Effective Full Power Hours High Reactor Pressure Note (7) Note (7) High Drywell Pressure Note (7) Note (7) Reactor low Water Level Note (7) Note (7) High Water Level in Scram Discharge Tanks Note (7) . Note (7) Turbine Condenser Low Vacuum Note (7) Note (7) Main Steam Line Isolation Valve Closure Note (6) Note (6) Main Steam Line High Radiation Standard Current Source (3) Every 3 Months Turbine First Stage Pressure Permissive Note (7) Note (7) Turbine Control Valve Fast closure Standard Pressure Source Every 3 Months Turbine Stop Valve Closure Note (6) Note (6) Reactor Pressure Permissive Note (7) Note (7) 32 Amendment No. 147, __.________.__._._.__.__.-__.___.____.___.___._.__________.__________________v-s_.- - ---. .,____._a_m _ _-__2 _ _ _ _ _m _ - -s ___._ _ _ ___-_s _____m_- - ___a - m ~
I i 3.1 BASES (Cont'd) i Scram Discharae Instrument Volume i The control rod drive scram system is designed so that all of the water that is i discharged from the reactor by a scram can be accommodated in the discharge j piping. The two scram discharge volumes have a capacity of 48 gallons of water . each and are at the low points of the scram discharge piping. l During normal operation the scram discharge volume system is empty; however, should it fill with water, the water discharged to the piping could not be i accommodated which would result in slow scram times or partial control rod I insertion. To preclude this occurrence, redundant and diverse level detection ; devices in the scram discharge instrument volumes have been provided. The ! instruments are set to alarm, initiate a control rod block, and scram the l ! reactor at three different progressively increasing water levels in the volume. l As indicated above, there is sufficient volume in the piping to accommodate the i scram without impairment of the scram times or amount of insertion of the : control rods. This function shuts the reactor down while sufficient volume i remains to accommodate the discharged water and precludes the situation in ! which a scram would be required but not be able to perform its function properly. ! i 4.1 BASES f f The reactor protection system is made up of two independent trip systems. l There are usually four channels to monitor each parameter with two channels in i* each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of , both trip systems will produce a reactor scram. The system meets the intent of j IEEE-279 for nuclear power plant protection systems. Specified surveillance i intervals and surveillance and maintenance outage times have been determined in I accordance with General Electric Company Topical Report NEDC-30851P-A, i
" Technical Specification Improvement Analysis for BWR Reactor Protection System," as approved by the NRC and documented in the safety evaluation report (NRC letter to T. A. Pickens from A. Thadani dated July 15,1987).
A comparison of Tables 4.1.1 and 4.1.2 indicates that two instrument channels , have not been included in the latter table. These are: mode switch in ; shutdown and manual scram. All of the devices or sensors associated with these ; scram functions are simple on-off switches and, hence, calibration during operation is not applicable (i.e., the switch is either on or off). l The sensitivity of LPRM detectors decreases with exposure to neutron flux at a j slow and approximately constant rate. This is compensated for in the APRM ~ system by calibrating every three days using heat balance data and by I calibrating individual LPRM's every 1000 effective full power hours using TIP ! traverse data. i Amendment No. 42, 133, 138, 147 40 l l l
PHPS TABLE 3.2.A INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION Operable Instrument Channels Per Trio System (1) Minimum Available Instrument Trip level Settina Action (2) 2(7) 2 Reactor Low Water Level 211.7" indicated level (3) A and D l 1 , 1 Reactor High Pressure ,
$110 psig D ,
2 2 Reactor Low-low Water Level at or above -46.3 in. A indicated level (4) 2 2 Reactor High Water Level 545.3" indicated level (5) B 2(7) 2 High Drywell Pressure $2.22 psig A 2 2 High Radiation Main Steam 57 times normal rated B Line Tunnel (9) full power background 2 2 Low Pressure Main Steam Line 2810 psig (8) B 2(6) 2 High Flow Main Steam Line 5136% of rated steam flow B l 2 2 Main Steam Line Tunnel Exhaust Duct High Temperature $170 F B 2 2 Turbine Basement Exhaust Duct High Temperature $150 F B 1 1 Reactor Cleanup System High Flow $300% of rated flow C 2 2 Reactor Cleanup System High Temperature $150 F C Amendment No. 86, 147,,150 45
d ! l l ! i
- 3. Instrument set point corresponds to 137.96 inches above top of active fuel. l
- 4. Instrument set point corresponds to 79.96 inches above top of active fuel. l ;
i
- 5. Not required in Run Mode (bypassed by Mode Switch).
j i { !
- 6. Two required for each steam line. !
- 7. These signals also start SBGTS and initiate secondary containment isolation. -!
- 8. Only required in Run Mode (interlocked with Mode Switch). i
- 9. Within 24 hours prior to the planned start of hydrogen injection with the !
reactor power at greater than 20% rated power, the normal full power radiation i background level and associated trip setpoints may be changed based on a ! calculated value of the radiation level expected during the injection of i hydrogen. The background radiation level and associated trip setpoints may be ! adjusted based on either calculati6ns or measurements of actual radiation levels l resulting from hydrogen injection. The background radiation level shall be l determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and l prior to withdrawing control rods at reactor power levels below 20% rated power. : ! e i I f i l i i Amendment No. 147 46a i l j
- . - ~ _ . _ _ _ - - -. - _ - _ _ _ _ . . . . . ~ _ . .-- .- .-
l PNPS TABLE 3.2.B IJJSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS I Minimum # of Operable Instrument Channels Per Trip System (1) Trin Function Trio Level Settinst Remarks 2 Reactor Low-Low Water Level at or above -46.3 in. 1. In conjunction with Low l indicated level-(4) Reactor Pressuro, initiates
. Core Spray and LPCI.
- 2. In conjunction with High Drywell Pressure, 94.4 -
115.6 second time delay and LPCI or Core Spray pump interlock initiates Auto Blowdown (ADS).
- 3. Initiates HPCI; RCIC.
- 4. Initiates starting of Diesel Generators.
2 Reactor High Water Level 5 +45.3" indicated Trips HPCI and RCIC turbines. l 1evel 1 Reactor Low Level >-151" indicated Prevents inadvertent operation l (inside shroud) level of containment spray during accident condition. (Indicative of 2/3 core coverage) 2 Containment High Pressure 1.55 5 p $ 1.82 psig Prevents inadvertent operation l of containment spray during accident condition. Instrument is set to trip at or before 1.82 increasing and reset at or before 1.55 decreasing. Amendment No. 90 47
PNPS TABLE 3.2.B (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trip System (1) Trip Function Trip Level Settinn Remarks 2 High Drywell Pressure $2.22 psig 1. Initiates Core Spray; LPCI; HPCI. l
- 2. In conjunction with Low-Low Reactor Water Level, 94.4 - 115.6 second time l delay and LPCI or Core Spray pump running, initiates Auto Blowdown (ADS)
- 3. Initiates starting of Diesel Generators 4 In conjunction with Reactor Low Pressure initiates closure of HPCI vacuum breaker containment isolation valves.
1 Reactor Low Pressure 400 psig i 5 Permissive for Opening Core Spray and l LPCI Admission valves. 1 Reactor Low Pressure $110 psig In conjunction with PCIS signal permits closure of RHR (LPCI) injection valves . 1 Reactor Low Pressure 400 psig i 5 In conjunction with Low-Low Reactor l Water Level initiates Core Spray and LPCI. 2 Reactor Low Pressure 900 psig i 5 Prevents actuation of LPG 1 break detection circuit. 2 Reactor Low Pressure 80 psig i 5 Isolates HPCI and in conjunction with l High Drywell Pressure initiates closure of HPCI vacuum breaker containment isolation valves. Amendment No. 42;-113, 148 48 5 i __.--___--_...__---__.-_____________.---_..c_ . . _ . - . . . _ , -~ .c -- , - - - - , -- - , _. -, _ n_ . - - . - . - - - . _ - . ,. . ...
PHPS TABLE 3.2.B (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument
. Channels Per Trip System (1) Trip Function Trio Level Setting Remarks 1 Core Spray Pump Start 0.21 < t < 1 sec. Initiates sequential starting of Timer CSCS pumps on any auto start.
1 LPCI Pump Start Timer 4.16 < t < 5.84 sec. 1 LPCI Pump Start Timer 9.5 < t < 11.5 sec. 1 Auto Blowdown Timer 2 94.4, $115.6 sec. In conjunction with Low Low Reactor Water level, High Drywell Pressure and LPCI or Core Spray Pump running interlock, initiates Auto Blowdown. 2 ADS Drywell P.ressure 9 $ t $ 15.4 min. Permits starting CS and LPCI l Bypass Timer pumps and actuating ADS SRV's if RPV water level is low and drywell pressure is not high. 2 RHR (LPCI) Pump Discharge 150 10 psig Defers ADS actuation pending Pressure interlock confimation of low Pressure core cooling system operation. 2 Core Spray Pump Discharge 150 10 psig (LPCI or Core Spray Pump Pressure Interlock running interlock.) 2 Emergency Bus Voltage 20-25% of rated 1. Permits closure of the Diesel Relay voltage resets Generator to an unloaded at less than or emergency bus, equal to 50% l
- 2. Permits starting of CSCS 4
. kV motors.
Amendment No. 40,-196, 120 49
-----.------,------_..----------------_w--- - - - - - - . - _ - - - , - _ , - - - _ _ - _ _ --___m ,- e-- -
- w. --r- w- -+-,-
_ _ --v-- - -- ,
PNPS TABLE 3.2.8 (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trip System (1) Trio Function Trip Level Setting Remarks 2 Startup Transformer At 0 Volts between 1. Trips Startup Transformer to Loss of Voltage 0.96 5 t $ 1.34 seconds Emergency Bus Breaker. Time Delay.
- 2. Locks out automatic closure of Startup Transformer to Emergency Bus.
- 3. Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
- 4. Prevents simultaneous starting of CSCS components.
- 5. Starts load shedding logic for Diesel Operation conjunction with (a) Low Low keactor Water level and Low Reactor Pressure or (b) High drywell pressure or (c) Core Standby Cooling System components in service in conjunction with Auxiliary Transformer breaker open.
Amendment 50
PHPS TABLE 3.2.B (Cont'd) INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS f Minimum # of Operable Instrument
' Channels Per Trip System (1) Trio Function Trip level Settina Remarks 2 Startup Transformer 3878.7V .51% with 10.24 1. Trips Startup Transformer to Degraded Voltage 0.36 seconds time delay. Emergency Bus Breaker.
- 2. Locks out automatic closure of Startup Transformer to Emergency Bus.
- 3. Initiates starting of Diesel Generators in conjunction with loss of auxiliary transformer.
- 4. Prevents simultaneous starting of CSCS components.
- 5. Starts load shedding logic for Diesel Operation in conjunction with a) Low Low Reactor Water Level and Low Reactor Pressure or b) High drywell pressure or c) Core Standby Cooling System components in service in conjunction with Auxiliary Transformer breaker open.
b Amendment No. 42, 61, 108, 120 50a
l NOTES FOR TABLE 3.2.B
- l. Whenever any CSCS subsystem is required by Section 3.5 to be operable, there l shall be two (Note 5) operable trip systems. If the first column cannot be met '
for one of the trip systems, that system shall be repaired or the reactor shall be placed in the Cold Shutdown Condition within 24 hours after this trip system is made or found to be inoperable. ! i
- 2. Close isolation valves in RClf, subsystem.
- 3. Close isolation valves in HPCI subsystem.
I
- 4. Instrument set point corresponds to 79.96 inches above top of active fuel. l l
- 5. RCIC has only one trip system for these sensors. !
i i i 1 l : 1 I P i
}
}
l l 1 53 Amendment No. 105, 148
PNPS TABLE 3.2.B.1 INSTRUMENTATION THAT MONITORS EMERGENCY BUS VOLTAGE Minimum e of Operable Instrument Channels Per Trip system Function Settinn Remarks 1 Emergency 4160V Buses A5 3958. 5V + 0.5% , -0.24% Alerts Operator to possible
& A6 Degraded Voltage , with 10.241 0.36 seconds degraded voltage conditions.
Annunciation (1) seconds time delay Provides permissive to initiate load snedding in conjunction with LOCA signal. (1) In the event that toe alarm system is determined inoperable, commence logging safety related bus voltage everyM hour until such time as the alarm is restored to operable status. Amendment No. 42r-617 -1981 -120 53a . - - - . . - - _ . - . - - - - - . - . . - ~ . - - . . - -. ..- - - ..---..-...-.. _ . . - . - . - . . . . . - . , , . . . . . . . - . - . . - . . . - . - . . . . . . - . - - . .
PNPS TABLE 3.2.C-2 CONTROL R0D BLOCK INSTRUMENTATION SETPOINTS Trio Function Trio Setooint AFRM Upscale (1) (2) APRM Inoperative Not Applicable APRM Downscale 2 2.5 Indicated on Scale Rod Block Monitor (Power Dependent) (1) (3) Rod Block Monitor Inoperative Not Applicable I Rod Block Monitor Downscale (1) (3) IRM Downscale 2 5/125 of Full Scale IRM Detector not in Startup Position Not Applicable l IRM Upscale s 108/125 of Full Scale l l IRM Inoperative Not Applicable SRM Detector not in Startup Position Not Applicable SRM Downscale 1 3 counts /second SRM Upscale 5 i s 10 counts /second SRM Inoperative Not Applicable l Scram Discharge Instrument Volume s 17 gallons i Water Level - High Scram Discharge Instrument Volume - Not Applicable Scram Trip Bypassed Recirculation Flow Converter - Upscale s 120/125 of Full Scale Recirculation Flow Converter - Not Applicable Inoperative Recirculation Flow Converter - 5 8% Flow Deviation Comparator Mismatch l (1) The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT. (2) When the reactor mode switch is in the refuel or startup positions, the APRM rod block trip setpoint shall be less than or equal to 13% of rated thermal power, but always less than the APRM flux scram trip setting. (3) The RBM bypass time delay (td2) shall be < 2.0 seconds. Amendment No. 42, 110, 129, 133, 138 55a 1 l _ _
PNPS TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUMP TRIP AND , ALTERNATE ROD INSERTION Minimum Number of Operable or Tripped t Instrument Channels Per Trio System (1) Trio Function Trio Level Settina 2 High Reactor Dome 1175 5 PSIG l Pressure , 2 Low-Low Reactor >-46.3" l Water Level indicated level , i Actions (1) There shall be two (2) operable trip systems for each function. 1 (a) If the minimum number of operable or tripped instruraent l channels for one (1) trip system cannot be met, restore the l trip system to operable status within 14 days or be in at l least hot shutdown within 24 hours. l (b) If the minimum operability conditions (l.a) cannot be met j for both (2) trip systems, be in at least hot shutdown ' within 24 hours. 1 Amendment 105 59a
-_______________ _ _ _ _ _ _ _ _- , , . , , . - . , , .o
PNPS TABLE 4.2.A MINIMUM TEST AND CALIBRATI0ii FREQUENCY FOR PCIS Instrument Channel (5) Instrument Functional Test Calibration Freauency Instrument Check
- 1) Reactor High Pressure (1) Once/3 months None
- 2) Reactor low-Low Water Level Once/3 Months (7) (7) Once/ day
- 3) Reactor High Water Level Once/3 Months (7) (7) Once/ day
- 4) Main Steam High Temp. (1) Once/3 months None
- 5) Main Steam High Flow Once/3 Months (7) (7) Once/ day
- 6) Main Steam Low Pressure Once/3 Months (7) (7) Once/ day
- 7) Reactor Water Cleanup High Flow (1) Once/3 months Once/ day
- 8) Reactor Water Cleanup High Temp. .
(1) Once/3 months None Loaic System Functional Test (4) (6) Freauency
- 1) Main Steam Line Isolation Vvs. Once/ operating cycle l Main Steam Line Drain Vvs.
Reactor Water Sample Vvs. s
- 2) RHR - Isolation Vv. Control Once/ operating cycle l Shutdown Cooling Vvs.
Head Spray Discharge to Radwaste
- 3) Reactor Water Cleanup Isolation Once/ operating cycle l
- 4) Drywell Isolation Vvs. Once/ operating cycle l TIP Withdrawal Atmospheric Control Vvs.
Sump Drain Valves
- 5) Standby Gas Treatment System Once/ operating cycle Reactor Building Isolation l Amendment No. 107, 130 60
PNPS TABLE 4.2.B MINIMUM TE,ST AND CAllBRATION FREQUENCY FOR CSCS Instrument Channel Instrument Functional Test Calibration Frecuency Instrument Check
- 1) Reactor Water Level (1) (7) (7) Once/ day
- 2) Drywell Pressure (1) (7) (7) Once/ day
- 3) Reactor Pressure (1) (7) (7) Once/ day
- 4) Auto Sequencing Timers NA Once/ operating cycle None
- 5) ADS - LPCI or CS Pump Disch.
Pressure Interlock (1) Once/3 months None
- 6) Start-up Transf. (4160V)
- a. Loss of Voltage Relays Monthly Once/ operating cycle None
- b. Degraded Voltage Relays Monthly ' Once/ operating cycle None
- 7) Trip System Bus Power Monitors Once/ operating cycleNA Once/ day
- 8) Recirculation System d/p (1) Once/3 months Once/ day
- 9) Core Spray Sparger d/p NA Once/18 months Once/ day l
- 10) Steam Line High Flow (HPCI & RCIC) (1) Once/3 months None
- 11) Steam Line High Temp. (HPCI & RCIC) (1) Once/3 months None
- 12) Safeguards Area High Temp. (1) Once/3 months None
- 13) RCIC Steam Line Low Pressure (1) Once/3 months None
- 14) HPCI Suction Tank Levels (1) Once/3 months None
- 15) Emergency 4160V Buses A5 & A6 Monthly Once/ operating Cycle None Loss of Voltage Relays i 61 Amendment No. 42 7 -611-99, 148
PNPS TABLE 4.2.B MINIMUM TEST AND CALIBRATION FRE0VENCY FOR CSCS Logical System Functional Test (4) (6) Frequency Remarks , -+
- 1) Core Spray Subsystem Once/ operating cycle
- 2) Low Press. Coolant Injection Subsystem Once/ operating cycle
- 3) Containment Spray Subsystem Once/ operating cycle
- 4) HPCI Subsystem Once/ operating cycle
- 5) HPCI Subsystem Auto Isolation Once/ operating cycle
- 6) ADS Subsystem Once/ operating cycle .
- 7) RCIC Subsystem Auto Isolation Once/ operating cycle ,
- 8) Diesel Generator Initiation Once/ operating cycle
- 9) Area Cooling for Safeguard System Once/ operating cycle 4
1 4 i l Amendment No. 130 62 i
PNPS TABLE 4.2.C MINIMUM TEST AND CALIBRATION FREQUENCY FOR CONTROL ROD BLOCKS ACTUATION Instrument Channel Instrument Functional Calibration Instrument Check Test APRM - Downscale Once/3 Months Once/3 Months Once/ Day APRM - Upscale once/3 Months Once/3 Months once/ Day APRM - Inoperative , Once/3 Months , Not Applicable Once/ Day, IRM - Upscale, (2) (3) Startup or Control Shutdown (2) IRM - Downscale (2) (3) Sta rtup or Control Shutdown (2) IRM - Inoperative (2) (3) Not .Sonlicable (2) RBM - Upscale once/3 Months once/6 Months once/ Day RBM - Downscale once/3 Months once/6 Months once/ Day RBM - Inoperative once/3 Months Not Applicable Once/ Day SRM - Upscale (2) (3) Startup or Control Shutdown (2) SRM - Inoperative (2) (3) Not Applicable (2) SRM - Detector Not in Startup Position (2) (3) Not Applicable (2) SRM - Downscale (2) (3) Startup or Control Shutdown (2) . IRM - Detector Not in Startup Position (2) (3) Not Applicable (2) Scram Discharge Instrument Volume once/3 Months Refuel Not Applicable Water Level-High Scram Discharge Instrument Once/3 Months Not Applicable Not Applicable Volume-Scram Trip Bypassed Recirculation Flow Converter Not Applicable Once/ Operating Cycle Once/ Day l Recirculation Flow Converter-Upscale Once/3 Months Once/3 Months once/ Day Recirculation Flow Converter-Inoperative Once/3 Months Not Applicable once/ Day Recirculation Flow Converter-Comparator Once/3 Months once/3 Months Once/ Day Off Limits Recirculation Flow Process Instruments Not Applicable once/ Operating Cycle Once/ Day g i Lonic System Functional Test (4) (6) System Logic Check once/ Operating cycle l
- Amendment No. ~110,-130, 147 63 i
PNPS TABLE 4.2.0 MINIMUM TEST AND CALIBRATION FRE0VENCY FOR RADIATION MONITORING SYSTEMS Instrument Channels Instrument Functional Calibration Instrument Check (2) Test
- 1) Refuel Area Exhaust Monitors - Upscale (1) Once/3 months Once/ day
- 2) Refuel Area Exhaust Monitors - Down : cale (1) Once/3 months Once/ day Logic System Functional Test (4) (5) Freauenc_y i
- 1) Reactor Building Isolation Once/ operating cycle
- 2) Standby Gas Treatment System Actuation Once/ operating cycle I
i } i Amendment No. 89, 130 64 _ _ _ _ . . - . . _ _ _ _ . _ . . -..___,_. _ _ _ _ _ _ __ . - _ . . - ~ . _ . _ . . _ _ _ . . . _ _ _ _ _ _ _ . - . _. .__ . ._ . . _ . . - _ - _
i BASES: l 3.2 In addition to reactor protection instrumentation which initiates a reactor l scram, protective instrumentation has been provided which initiates action to r mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious ! consequences. This set of specifications provides the limiting conditions of ; operation for the primary system isolation function, initiation of the core ! cooling systems, control rod block, and standby gas treatment systems. The , objectives of the Specifications are, (i) to assure the effectiveness of the protective instrumentation when required by preserving its capability to ; tolerate a single failure of any component of such systems even during periods ! when portions of such systems are out of service for maintenance, and (ii) to , prescribe the trip settings required to assure adequate performance. When , necessary, one channel may be made inoperable for brief intervals to conduct ' required functional tests and calibrations. Some of the settings on the instrumentation that initiate or control core and : containment cooling have tolerances explicitly stated where the high and low ! values are both critical and may have a substantial effect on safety. The set ' points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal l operating range to prevent inadvertent actuation of the safety system involved ! and exposure to abnormal situations. i Actuation of primary containment valves is initiated by protective > instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required. Such instrumentation must be available whenever i primary containment integrity is required. l The instrumentation which initiates primary system isolation is connected in a I dual bus arrangement. ; The low water level instrumentation closes all isolation valves except l l those in Groups 1, 4 and 5. This trip setting is adequate to prevent core uncovery in the case of a break in the largest line assuming a 60 second valve j closing time. Required closing times are less than this. , The low low reactor water level instrumentation closes the Main Steam Line l Isolation Valves, Main Steam Drain Valves, Recirc Sample Valves (Group 1) I activates the CSCS subsystems, starts the emergency diesel generators and l trips the recirculation pumps. This trip setting level was chosen to be ' high enough to prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that no fuel damage will occur and so that post accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be violated. For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system isolation are initiated in time to meet the above criteria. i Amendment No. 105, 113 68
i l 3.2 BASES (Cont'd) l The high drywell pressure instrumentation is a diverse signal to the l water level instrumentation and in addition to initiating CSCS, it ! causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumentation will initiate CSCS operation ~at about the ' same time as the low low water level instrumentation; thus the results j given above are applicable here also. The low low water level ; instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of Group 1 isolation valves. I Venturis are provided in the main steam lines as a means of measuring i steam flow and also limiting the loss of mass inventory from the vessel i during a steam line break accident. The primary function of. the instrumentation is to detect a break in the main steam line. For the ! worst case accident, main steam line break outside the drywell, the ; steam flow trip setting in conjunction with the flow limiters and main j steam line valve closure, limits the mass inventory loss'such that fuel , is not uncovered, fuel temperatures remain approximately 1000 F and ! release of radioactivity to the environs ,is well below 10 CFR 100 l guidelines. ! Temperature monitoring instrumentation is provided in the main steam l line tunnel and the turbine basement to detect leaks in these areas, i Trips are provided on this instrumentation and when exceeded, cause : closure of isolation valves. The setting of 170 F.for the main steam > line tunnel detector is low enough to detect leaks of the order of 5 to i 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, the high steam flow instrumentation is a backup to the ; temperature instrumentation. ; i High radiation monitors in the main steam line tunnel have been provided [ to detect gross fuel failure as in the control rod drop acci- - r i t i i l l l I i i Amendment No. 34, 113 69 !
-- , ,---,-,,w- , . - - - . - , - - - - , - - , - . ,ee,..-,-,.,r, ,, r, aw- n --...e. m.
t i I 4.2 BASES (Cont'd) ; The automatic pressure relief instrumentation can be considered to be a 1 out of 2 logic system and the discussion above applies also. > The instrumentation which is required for the recirculation pump trip l and alternate rod insertion systems incorporate analog transmitters. < The transmitter calibration frequency is once per refueling outage, ! which is consistent with both the egn.pment capabilities and the l requirements for similar equipment , sed at Pilgrim. The Trip Unit Calibration and Instrument Functionil Test is specified- at monthly, ' which is the same frequency specified for other similar protective . devices. An instrument check is specified at once per day; this is r considered to be an appropriate frequency, commensurate with tne design ' applications and the fact that the recirculation pump trip and alternate l rod insertion systems are backups to existing protective instrumentation. 1 Control Rod Block and PCIS instrumentation common to RPS instrumentation ; have surveillance intervals and maintenance outage times selected in ' accordance with NEDC-30851P-A, Supplements 1 and 2 as approved by the i NRC and documented in SERs (letters to D. N. Grace from C. E. Rossi l dated September 22, 1988 and January 6, 1989). l A logic system functional test interval of 24 months was selected to. l , minimize the frequency of safety system inoperability due to testing and ! to minimize the potential for inadvertent safety system trips and their i attendant transients. ! l i l Amendment No. 42, 121, 130, 147 77
l l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS l ! 3.6.C.2 Leakaae Detection Systems 4.6,C.2 Leakaae Detection Systems l (Cont'd) (Cont'd) , i
- 2. One channel of a drywell 2. An instrument channel atmospheric particulate calibration at least ,
radioactivity monitoring once per operating i cycle. system, or
- 3. One channel of a drywell b. For each required drywell
- atmospheric gaseous atmospheric radioactivity radioactivity monitoring monitoring system perform
- :
system.
- 1. An instrument check at :
- b. 1. At least one drywell sump least once per day, l monitoring system shall be !
Operable; otherwise, be in 2. An instrument functional ! Hot Shutdown within the test at least once per ! next 12 hours and in Cold 31 days, and ! Shutdown within the ; following 24 hours. 3. An instrument channel , calibration at least !
- 2. At least one gaseous or once per operating ;
- particulate radioactivity cycle. !
4 monitoring channel must be i Operable; otherwise, reactor operation may continue for up to 31 days provided grab samples are : obtained and analyzed every ; 24 hours, or be in Hot i Shutdown within the next 12 l hours and in Cold Shutdown ! within the following 24 hours. 4
- c. With no required leakage !
- detection systems Operable, be 1 in Cold Shutdown within 24 hours.
1 ] Amendment No. 139 125b
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.I Shock Sunnressors (Snubbers) 4.6.1 Shock Suppressors (Snubbers)
- 1. During all modes of operation The following surveillance except Cold Shutdown and Refuel, requirements apply to all safety ,
all . safety-related snubbers related hydraulic and mechanical [ listed in PNPS Procedures shall snubbers listed in PNPS Procedures. > be operable except as noted in ! 3.6.1.2 through 3.6.I.3 below. The required visual inspection ;
~
l interval varies inversely with the An Inoperable Snubber is a observed cumulative number of , properly fabricated, installed inoperable snubbers found during an , I and sized snubber which cannot inspection. Inspections performed ' pass its functional test. before that interval has elapsed may be used as a new reference point to Upon determination that a determine the next inspection. ; I snubber is either improperly However, the results of such early , fabricated, installed or sized, inspections performed before the ; the corrective action will be as original time interval has elapsed specified for an inoperable may not be used to lengthen the , snubber in Section 3.6.1.2. required interval. r
- 2. From and after the time that a Number of snubbers found inoperable {
snubber is determined to be during inspection or during ! inoperable, replace or repair inspection interval: i the snubber during the next 72 hours, and initiate an Subsequent ; engineering evaluation to Inoperable Visual Inspec- , determine if the components Snubbers tion Interval : supported by the snubber (s) were ; adversely affected by the 0 24 Months 25% inoperability of the snubbers 1 18 Months 25% ! and to ensure that the supported 2 12 Months 25% ! component remains capable of 3,4 6 Months i 25% i meeting its intended function in 5,6,7 124 Days 25% i the specific safety system 8,9 62 Days 25% ; involved. 10 or more 31 Days 25% l Further corrective action for The required inspection interval ! this snubber, ana all shall not be lengthened more than ; generically susceptible one step at a time. I snubbers, shall be determined by i an engineering evaluation. Snubbers may be categorized in two i groups, " accessible" or ;
- 3. From and after the time a " inaccessible" based on their )
snubber is determined to be accessibility for inspection during , inoperable, improperly reactor operation. These two groups i fabricated, improperly installed may be inspected independently or improperly sized, if the according to the above schedule. requirements of Section(s) 3.6.I.1 and 3.6.I.2 cannot be 1. Visual Inspection Acceptarce ' i met, then the affected safety Criteria system, or affected portions of that system, shall be declared A. Visual inspections shall inoperable, and the limiting verify: condition for that system entered, as appropriate. ; 1 Amendment No. 20, 60, 93 137a i
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 3.6.1 Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snubbers)
- 4. Snubbers may be added to, or 1. That there are no visible removed from, per 10 CFR 50.59, indications of damage or safety related systems without impaired operability.
prior NRC approval. The addition or deletion of snubbers 2. Attachments to the foundation ! shall be reported to the NRC in or support structure are such accordance with 10 CFR 50.59. that the functional capability of the snubber is not suspect. ' B. Snubbers which y pear INOPERABLE as a result of visual , inspections may be determined : OPERABLE for the purpose of establishing the next visual inspection interval provided that:
- 1. The cause of the rejection is ;
clearly established and , remedied for that particular snubber, and
- 2. The affected snubber is I functionally tested, when necessary, in the as found condition and determined OPERABLE per specifications i 4.6.I.2.B., 4.6.I.2.C., as applicable.
C. For any snubber determined l ; inoperable per specification 1 4.6.I.2, clearly establish the cause of rejection and remedy the problem for that snubber, and any generically susceptible snubber.
- 2. Functional Tests (Hydraulic and Mechanical Snubbers) .
I A. Schedule At least once per operating cycle, a representative sample l l (12.5% of the total of each type: : hydraulic, mechanical) of snubbers in use in the plant shall be . functionally tested, either in place I or in a bench test. For each snubber that does not meet the functional test acceptance criteria of , Amendment No. 20, 60, 93 137b l l
LIMITING CONDITIONS F6R OPERATION SURVEILLANCE REQUiREMEitT_S 4.6.1 Shock Suppressors (Snubbersl Specification 4.6. I.2.8, or' , 4.6.I.2.C, as applicable, an ; additional 12.5% of that l , type of snubber shall be functionally tested. B. General Snubber Functional Test Acceptance Criteria - (Hydraulic and Mechanical) The general snubber functional , test shall verify that: l
- 1. Activation (restraining action) is achieved within ,
the specified range of ' velocity or acceleration in both tension and compression. -
- 2. Snubber release, or bleedrate, as applicable, where required, is within ,
the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be verified. ; C. Mechanical Snubbers Functional Test Acceptance Criteria , The mechanical snubber ; functional test shall verify that:
- 1. The force that initiates '
free movement of the snubber rod in either tension or compression is l less than the specified maximum drag force. Drag force shall not have increased more than 50% since the last functional test.
- 3. Snubber Service Life Monitorina '
A. A record of the service life ] Amendment No. 60 137c > i
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS - l l 3.7.B Standby Gas Treatment System 4.7.B Standby Gas Treatment System L l and Control Room Hiah and Control Room Hiah l Efficiency Air Filtration Efficiency Air Filtration System System
- 1. Standby Gas Treatment System 1. Standby Gas Treatment System ;
- a. Except as specified in a. (1.) At least once per !
3.7.B.l.c below, both trains operating cycle, it shall of the standby gas treatment be demonstrated that system and the diesel pressure drop across the , generators required for combined high efficiency ; operation of such trains shall filters and charcoal : be operable at all times when adsorber banks is less than l , secondary containment 8 inches of water at 4000 l integrity is required or the cfm. : l reactor shall be shutdown in l 36 hours. (2.) At least once per operating cycle,
- b. (1.) The results of the in- demonstrate that the ;
inlet heaters on each train l place cold DOP tests on HEPA filters shall show are operable and are capable
- 199% DOP removal. The of an output of at least 14 ;
results of halogenated kW. l hydrocarbon tests on J charcoal adsorber banks (3.) The tests and analysis of > shall show 299% Specification 3.7.B.I.b. : halogenated hydrocarbon shall be performed at least removal. once per operating cycle l or following painting, fire (2.) The results of the or chemical release in any laboratory carbon sample ventilation zone ; analysis shall show 195% communicating with the ; methyl iodide removal at system while the system is a velocity within 10% of operating that could i system design, 0.5 to I 5 contaminate the HEPA filters l mg/m3 inlet methyl iodide or charcoal adsorbers. concentration, 270% R.H. : and 2190 F. The analysis (4.) At least once per l results are to be operating cycle, automatic ! verified as acceptable initiation of each branch of within 31 days after the standby gas treatment ; sample removal, or system shall be declare that train demonstrated, with inoperable and take the Specification 3.7.B.1.d actions specified satisfied. 3.7.B.l.c. Amendment No. 50, 51, 52, 112, 144 158
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7.B (Continued) 4.7.B (Continued)
- 2. Control Room High Efficiency Air 2. Control Room Hiah Efficiency Air Filtration System Filtration System t
- a. Except as specified in a. At least once per operating i Specification 3.7.B.2.c cycle the pressure drop across below, both trains of the each combined filter train shall ;
Control Room High Efficiency be demonstrated to be less than ' Air Filtration System used 6 inches of water at 1000 cfm or f for .he processing of inlet the calculated equivalent. ! air to the control room under , accicent conditions and the b. (1.) The tests and analysis of i diesel generator (s) required Specification 3.7.B.2.b for operation of each train shall be performed once of the system shall be per operating cycle or operable whenever secondary following painting, fire or containment integrity is chemical release in any required and during fuel ventilation zone handling operations. communicating with the system while the system is ,
- b. (1.) The results of the in- operating.
place cold D0P tests on HEPA filters shall show (2.) In-place cold D0P testing 299% DOP removal. The shall be performed after results of the each complete or partial halogenated hydrocarbon replacement of the HEPA tests on charcoal filter bank or after any adsorber banks shall structural maintenance on ! show 199% halogenated the system housing which I hydrocarbon removal when could affect the HEPA ! test results are filter bank bypass leakage. extrapolated to the initiation of the test. (3.) Halogenated hydrocarbon ; testing shall be performed , (2.) The results of the after each complete or laboratory carbon sample partial replacement of the analysis shall show 195% charcoal adsorber bank or ! methyl iodide removal at after any structural < a velocity within 10% of maintenance on the system , system design, 0.05 to housing which could affect , 0.15 mg/m3 inlet methyl the charcoal adsorber bank iodide concentration, bypass leakage. 270% R.H., and 2125 F. The analysis results are (4.) Each train shall be to be verified as operated with the heaters acceptable within 31 in automatic for at least days after sample 15 minutes every month. removal, or declare that train inoperable and (5.) The test and analysis of l take the actions Specification 3.7.B.2.b.(2) specified in 3.7.B.2.c. shall be performed after every 720 hours of system l operation. During RF0 #9, one train can be without its safety-related bus and/or its emergency diesel generator without entering the LCO action statement provided the conditions listed on page 158A are met. Amendment No. 50, 51, 52, 101, 112, 144 158B
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0UIREMENTS l 3.7.8 (Continued) 4.7.B (Continued) l
- c. From and after the date that c. At least once per operating one train of the Control Room cycle demonstrate that the ;
High Efficiency Air inlet heaters on each train i Filtration System is made or are operable and capable of . found to be incapable of an output of at least 14 kw. supplying filtered air to the control room for any reason, d. Perform an instrument reactor operation or functional test on the ; refueling operations are humidistats controlling the { permissible only during the heaters once per operating i succeeding 7 days providing cycle. : that within 2 hours all ; active components of the ! other CRHEAF train shall be ; demonstrated operable. If ; the system is not made fully i operable within 7 days, reactor shutdown shall be initiated and the reactor shall be in cold shutdown j within the next 36 hours and ; irradiated fuel handling operations shall be terminated within 2 hours. t Fuel handling operations in i progress may be completed. ;
- d. Fans shall operate within ;
10% of 1000 cfm. . i f t l l l During RF0 #9, one train can be without its safety-related bus and/or its emergency diesel generator without entering the LC0 action statement i provided the conditions listed on page 158A are met. l 1 l l l l Amendment No. 50, 51, 57, 112, 144 158C
i BASES: l 3.7.A & 4.7.A Primary Containment t t Grouc 6 - process lines are normally in use and it is therefore not desirable to cause spurious isolation due to high drywell pressure resulting from non-safety related causes. To protect the reactor from a possible pipe break in the system, isolation is provided by high temperature in the cleanup system area or high flow through the inlet to the cleanup system. Also, since the ! vessel could potentially be drained through the cleanup system, a low level ; isolation is provided. i Grouc 7 - The HPCI vacuum breaker line is designed to remain operable when the ! HPCI system is required. The signals which initiate isolation of the HPCI ' vacuum breaker line are indicative of a break inside containment and reactor l pressure below that at which HPCI can operate. ; The maximum closure time for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in , consideration of the design intent to prevent core uncovering following pipe , breaks outside the primary containment and the need to contain released , fission products following pipe breaks inside the primary containment. : In satisfying this design intent an additional margin has been included in i specifying maximum closure times. This margin permits identification of ' degraded valve performance, prior to exceeding the design closure times. In order to assure that the doses that may result from a steam line break do not exceed the 100FR100 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate that fuel rod cladding ; perforations would be avoided for main steam valve closure times, including instrument delay, as long as 10.5 seconds. These valves are highly reliable, have low service requirements and most are , normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation. The test interval of once per operating cycle for automatic initiation results in a failure probability of 1.1 x 10 7 that a line will not isolate. More frequent I testing for valve operability results in a greater assurance that the valve i will be operable when needed. The main steam line isolation valves are functionally tested on a more frequent interval to establish a high degree of reliability. The primary containment is penetrated by several small diameter instrument lines connected to the reactor coolant system. Each instrument line contains a 0.25 inch restricting orifice inside the primary containment. A program for ! periodic testing and examination of the excess flow check valves is in place. ; I Primary Containment Paintina The interiors of the drywell and suppression chamber are painted to prevent rusting. The inspection of the paint during each major refueling outage assures the paint is intact. Experience at Pilgrim Station and other BWRs with this type of paint indicates that the inspection interval is adequate. Amendment No. 113 169
i J BASES: j 3.7.B.1 and 4.7.B.1 - Standby Gas Treatment System j The Standby Gas Treatment System is designed to filter and exhaust the reactor
- building atmosphere to the stack during secondary containment isolation conditions. Upon containment isolation, both standby gas treatment fans are
{; designed to start to bring the reactor building pressure negative so that all 1 leakage should be in-leakage. After a preset time delay, the standby fan automatically shuts down so the reactor building pressure is maintained approximately 1/4 inch of water negative. Should one system fail to start, t the redundant system is designed to start automatically. Each of the two l trains has 100% capacity. l High Efficiency Particulate Air (HEPA) filters are installed before and after i the charcoal adsorbers to minimize potential release of particulates to the ; environment and to prevent clogging of the iodine adsorbers. The charcoal j ) adsorbers are installed to reduce the potential release of radioiodine to the j
- environment. The in-place test results should indicate a system leak
- tightness of less than 1 percent bypass leakage for the charcoal adsorbers and l a HEPA filter efficiency of at least 99 percent removal of cold DOP ;
; particulates. The laboratory carbon sample test results should indicate a ,
l methyl iodide removal efficiency of at least 95 percent for expected accident
; conditions. The specified efficiencies for the charcoal and particulate !
} filters is sufficient to preclude exceeding 10 CFR 100 guidelines for the ;
- accidents analyzed. The analysis of the loss of coolant accident assumed a j charcoal adsorber efficiency of 95% and TID 14844 fission product source ;
9 terms, hence, installing two banks of adsorbers and filters in each train ! j provides adequate margin. A 14 kW heater maintains relative humidity below : 70% in order to ensure the efficient removal of methyl iodide on the !
] ; impregnated charcoal adsorbers. Considering the relative simplicity of the ! ; heating circuit, the test frequency of once/ operating cycle is adequate to j j demonstrate operability.
Air flow through the filters and charcoal adsorbers for 15 minutes each month assures operability of the system. Since the system heaters are automatically l controlled, the air flowing through the filters and adsorbers will be 170% i relative humidity and will have the desired drying effect. Tests of impregnated charcoal identical to that used in the filters indicate that shelf life of five years leads to only minor decreases in methyl iodide - removal efficiency. Hence, the frequency of laboratory carbon sample analysis l l is adequate to demonstrate acceptability. Since adsorbers must be be removed i to perform this analysis this frequency also minimizes the system out of { service time as a result of surveillance testing. In addition, although the , i halogenated hydrocarbon testing is basically a leak test, the adsorbers have l ) charcoal of known efficiency and holding capacity for elemental iodine and/or methyl iodide, the testing also gives an indication of the relative efficiency of the installed system. The 31 day requirement for the ascertaining of test results ensures that the ability of the charcoal to perform its designed function is demonstrated and known in a timely manner. The required Standby Gas Treatment System flow rate is that flow, less than or equal to 4000 CFM which is needed to maintain the Reactor Building at a 0.25 inch of water negative pressure under calm wind conditions. This capability is adequately demonstrated during Secondary Containment Leak Rate Testing performed pursuant to Technical Specification 4.7.C.I.c. Amendment No. 42, 112 172
BASES: : l 3.7.B.1 and 4.7.B.1 (continued) The test frequencies are adequate to detect equipment deterioration prior to significant defects, but the tests are not frequent enough to load the filters or adsorbers, thus reducing their reserve capacity too quickly. The filter ! testing is performed pursuant to appropriate procedures reviewed and approved by the Operations Review Committee pursuant to Section 6 of these Technical Specifications. The in-place testing of charcoal filters is performed by , injecting a halogenated hydrocarbon into the system upstream of the charcoal ; adsorbers. Measurements of the concentration upstream and downstream are ; made. The ratio of the inlet and outlet concentratiens gives an overall l indication of the leak tightness of the system. A similar procedure - substituting dioctyl phthalate for halogenated hydrocarbon is used to test the l HEPA filters. ' l Pressure drop tests across filter and adsorber banks are performed to detect , plugging or leak paths though the filter or adsorber media. Considering the t relatively short times the fans will be run for test purposes, plugging is unlikely and the test interval of once per operating cycle is reasonable. l System drains and housing gasket doors are designed such that any leakage would be in-leakage from the Standby Gas Treatment System Room. This ensures that there will be no bypass of process air around the filters or adsorbers. Only one of the two Standby Gas Treatment Systems (SBGTS) is needed to maintain the secondary containment at a 0.25 inch of water negative pressure ! upon containment isolation. If one system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling activities may continue while repairs are being made. . In the event one SBGTS is inoperable, the redundant system's active components will be tested within 2 hours. This substantiates the availability of the operable system and justifies continued reactor or refueling operations. l 4 If both trains of SBGTS are inoperable, the plant is brought to a condition ! where the SBGTS is not required. l l l Amendment No. 42, 112, 173 ! l
--,_i
.- . _. - _ -= . _ _ .. . . . . _ _ . -~ - -. .. -
TABLE 4.8-2 RADI0 ACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE0VIREMENTS Channel Instrument Source Channel Functional Instrument Check Check Calibration Test
- 1. Gross Beta or Gamma Radioactivity Monitors Providing Alarm and Auto-1 matic Isolation
, a. Liquid Radwaste Effluents Line i NA Once per Quarterly l 18 months 2
- 2. Flow Rate Measurement Devices
- a. Liquid Radwaste Effluent Line i SA Once per Quarterly l 18 months t
i 1During or prior to release via this pathway. 2Previously established calibration procedures will be used for these requirements. 1 I i i Amendment No._89 190
TABLE 4.8-4 1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Source Instrument Functional Instrument Check Check Calibration Test
- 1. Main Stack Effluent Monitoring System
- a. Noble Gas Activity Monitor Daily l Monthly Once per Quarterly l (Two Channels) 18 Months 4
- b. Iodine Sampler Cartridge NA NA NA NA
- c. Particulate Sampler Filter NA NA NA NA
- d. Effluent System Flow Rate Daily l NA Once per Quarterly l Measuring Device 18 Months
, e. Sampler Flow Rate Measuring Daily l NA Once per Quarterly l Device 18 Months
- 2. Reactor Building Ventilation Effluent Monitoring System .
- a. Noble Gas Activity Monitor '
Daily l Monthly Once per Quarterly l 18 Months 4 i i
- b. Iodine Sampler Cartridge NA NA NA NA
- c. Particulate Sampler Filter NA NA NA NA
! d. Effluent System Flow Rate Daily l NA Once per Quarterly l Measuring Device 18 Months l l Amendment No. 89 193
- i
l TABLE 4.8-4 (continued) RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Instrument Instrument Source Instrument Functional Instrument Check Check Calibration Test
- e. Sampler Flow Rate Measuring Daily l NA Once per Quarterly l i Device 18 Months
- 3. Steam Jet Air Ejector Radioactivity Monitor
- a. Noble Gas Activity Monitor Daily 3 NA Once per Quarterly operaging cycle
- 4. Augmented Offgas Treatment System Explosive Gas Monitoring System
- a. Hydrogen Monitor Daily 2 NA Quarterly 5 Monthly 1
During releases via this pathway 2 Duiing augmented offgas treatment system operation. 3 During operation of the steam jet air ejector. 4 Previously established calibration procedures will be used for these requirements. 5 Calibrate at 2 points with standard gas samples differing by at least 1% but not exceeding 4%. I Amendment No. 89 193a
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 4.9.A Auxiliary Electrical l ] Eauipment Surveillance ! (Cont'd) I
- 1. Verifying de-energization of I the emergency buses and load i shedding from the emergency l buses. l t
- 2. Verifying the diesel starts .
from ambient condition on the auto-start signal, energizes the emergency buses with ! permanently connected loads,
- energizes the auto-connected ;
emergency loads through the load sequence, and operates i for 2 5 minutes while its generator is loaded with the { emergency loads. l During performance of this surveillance verify that HPCI and RCIC inverters do not , trip. ; The results shall be logged.
- c. Once per operating cycle with ;
4 the diesel loaded per 4.9.A.1.b verify that on diesel generator trip, secondary (offsite) AC power is automatically connected r within 11.8 to 13.2 seconds l to the emergency service ! buses and emergency loads are energized through the load
- sequencer in the same manner .
as described in 4.9.A.1.b.1. l, The results shall be logged. : h Amendment No. 42, 61, 141 194a i i
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9.B O J eration with Inoperable 4.9.A Auxiliary Electrical Eouipment Eauipment Surveillance (Cont'd) , Whenever the reactor is in Run 3. Emergency 4160V Buses AS-A6 Mode or n artup Mode with the Degraded Voltage Annunciation I reactor not in a Cold Condition, System. I the availability of electric power shall be as specified in a. Once each operating cycle, 3.9.B.1, 3.9.B.2, 3.9.B.3, calibrate the alarm sensor. 3.9.B.4, and 3.9.B.5.
- b. Once each 31 days perform a
- 1. From and after the date that channel functional test on the '
incoming power is not available alarm system. ; from the startup or shutdown ! transformer, continued reactor c. In the event the. alarm system 1 operation is permissible under is determined inoperable under ' this condition for seven days. 3.b above, commence logging l During this period, both diesel safety related bus voltage i generators and associated every 30 minutes until such ! emergency buses must be time as the alarm is restored ! demonstrated to be operable. , to operable status. ! l 2. From and af ter the date that 4. RPS Electrical Protection : incoming power is not available Assemblies i from both startup and shutdown ! transformers, continued a. Each pair of redundant RPS operation is permissible, EPAs shall be determined to provided both diesel generators be operable at least once and associated emergency buses per 6 months by performance , are demonstrated to be operable, of an instrument functional ; all core and containment cooling test. systems are operable, reactor ! power level is reduced to 25% of b. Once per 18 months, each l ! design and the NRC is notified pair of redundant RPS EPAs ! within one (1) hour as required shall be determined to be by 10CFR50.72. operable by performance of an instrument calibration l
- 3. From and after the date that one and by verifying tripping I of the diesel generators or of the circuit breakers associated emergency bus is made upon the simulated or found to be inoperable for conditions for automatic '
any reason, continued reactor actuation of the protective l operation is permissible in relays within the following l accordance with Specification limits: 3.5.F if Specification 3.9.A.1 and 3.9.A.2.a are satisfied. Overvoltage $; 132 volts Undervoltage 2 108 volts
- 4. From and after the date that one Underfrequency 2 57Hz l of the diesel generators or I associated emergency buses and either the shutdown or startup transformer power source are made l Amendment No. 88, 127,145 196 l
I l l
l LIMITING CONDITION FOR OPERATION SORVEILLANCE REQUIREMENTS l 3.9 AtrXILI ARY ELECTRICAL SYSTEM (Cont) J t B. Operation with Inoperable Equipment j (Cont) i or found to be inoperable for any reason, continued reactor operation is permissible in ' accordance with Specification 3.5.F. provided either of the i following conditions are satisfied; The startup transformer and i a. both offsite 345 kV transmission lines are available and capable of j automatically supplying , l auxiliary power to the ! emergency 4160 volt buses- l l
- b. A transmission line and l associated shutdown l transformer are available and !
capable of automatically , supplying auxiliary power to the emergency 4160 volt j buses. ;
- 5. From and after the date that one of the 125 or 250 volt battery systems is made or found to be inoperable for any reason, continued reactor operation is permissible during the succeeding three days within ;
electrical safety considerations, provided repair work is initiated in the most expeditious manner to return the ' failed component to an operable state, and Specification 3.5.F 1 is satisfied. ( 6. With the emergency bus voltage , l 1ess than 3958,5V but above 3878.7V(excluding transients) during normal operation, transfer the safety related buses to the diesel generators.. If grid voltage continues to l degrade be in at least Hot Shutdown within the next 4 hours and in Cold Shutdown within the following 12 hours unless the grid conditions improve. Amendment No. 422-61;-88;-120;-127 197
d ! BASES: (Cont'd) i ! 4.9 l l deliver full flow. Periodic testing of the various components, plus a functional i test once per cycle, is sufficient to maintain adequate reliability. ,
/
Although station batteries will deteriorate with time, utility experience indicates there is almost no possibility of precipitous failure. The type of l surveillance described in this specification has been demonstrated over the years ! to provide an indication of a cell becoming irregular or unserviceable long I before it becomes a failure. I Tne Service Discharge Test provides iridication of the batteries' atility to satisfy the design requirements (battery duty cycle) of the associated dc system. This test will be performed i. sing simulated or actual loads at the rates and for the duration specified in the design load profile. A once per cycle testing interval was chosen to coincide with planned outages. l The Performance Discharge Test provides adequate indication and assurance that the batteries have the specified ampere hour capacity. The results of these tests will be logged and compared with the manufacturer's recommendations of : acceptability. This test is performed once every five years in lieu of the Service Discharge test that would normally occur within that time frame. . i The diesel fuel oil quality must be checked to ensure proper operation of the diesel generators. Water content should be minimized because water in the fuel could contribute to excessive damage to the diesel engine. The electrical protection assemblies (EPAs) on the RPS inservice power supplies, either two motor generator sets or one motor generator and the alternative ! supply, consist of protective relays that trip their incorporated circuit breakers on overvoltage, undervoltage, or underfrequency conditions. There are l two EPAs in series per power source. It is necessary to periodically test the i relays to ensure the sensor is operating correctly and to ensure the trip unit is operable. Based on experience at conventional and nuclear power plants, a six-month frequency for the charmel functional test is established. This frequency ic consistent with the Stancard Technical Specifications. ; The EPAs of the power sources to the RPS shall be determined to be operable by ! performance of a channel calibration of the relays once per 18 months. During l calibration, a transfer to the alternative power source is required; however, prior to switching to alternative feed, de-energization of the applicable MG ! set power source must be accomplished. This resv'ts in a half scram on the I channel being calibrated until the alternative power source is connected and the ; half scram is cleared. Based on operating experience, drift of the EPA i protective relays is not significant. 1 i l l l I l Amendment No. 127, 135, 141, 145 201 l l 1
4 l s l i instrument Summaty by Technical Specihcation Section Attachment E i o Proposed l Current Surveillance Calculation Number Technical Existmg Techn cal Covermng Requirement j or Dualitative , lastrurnent or Comparent Specification Associated Component Specification Proposed Technical Surveillance Calibration Justification F unction Referent.e er Instrument Number ($stpoint) Value Specification Value Raouirement Interval Provided (DJP) RPS . Section 3.1 T 3.1.1 1. Each RF0 .Each RF 0 Mode Svetch in Shutdovvn T411.1 .na __
.r.a._ functensi test- .na - f unctional test f18 movithsj-- (24 monthst Q)P T 3.1 1.2.
Manual Scram 1_411_2. .na - - .na tunctona_I test _ ,na . f_unctmnal t est__ ,Every 3 months _ ,3 months, _ fa . _ _ . _ . T3.1.1.3a. l IRM HWh Flus Comperison . ContrM i 14113a. ContmW , T41,2.1 _ _ . ,na _ __ _ _ _ _ _ _ . _ . _ na __ . [na Shutdo wn Shutdown na j
- I
- IRM High Flus . Full T 4.113a. . < - 120l125 of f ull i < - 120:125 of full 'Onceicycle 0ncercycle Calibration T4.12.1 ' na scale scale 118 months) (24 monthst .QJP T3.1136.
IRM inoperative T4.113b na 'ns corrective acten na correctiva action Conditanal 'Conditionef na T3.1.1.4a. I APRM-th0 hFlux(flowbias . 5 T 411.4c, , < - 120% of rated j < - 120% of rated Each RF 0 no calibrate on. signal calibration) 8pR 18es b ' 13.114b. ! , APRM . inoperative ( T4114b 'na na . functional t est na functmnaltest Conditional 'Conditenal na T3.1.1.4c. l < - 15 % of design 1 < - 15 % of design Once'm eek. Once'm eek. APRM HighFlux(15%) T4.11.4d na power power lConditenell IConditenal) 'ns T 3.115. 4
. f High Reactor Pressure T411.5. i j [ach RF 0 f ach FtF0 ;
T4.1.2.4 P:S 263-51 A. B. C. D i < _1085 Psig _ _ } < -1063.5 psig (18 months) (24 months) . ..!C002 ._ __ , T3.116 i l l l High Drywell Pressure T4.116. '
.Each EF0 Each RF0 ,
) T4.12.5 ,PiS512A.B.C.D ; < - 2.5 pseg ' < -222 psig (18 months) . ,24( months) 'C 026 T311.7. Reactor Low Waterlevel T411.7 LS 263 57A 1. 57B 1
. tach RF0 Each RF0 l T412.6_ ,and 5B A-1. 56B _1 , > - 9 mches i
- > -11.7 mehes -(18 months)__ _ J24 monthst ;C010
.+
T3.116a. l l Scram Discharge Volume . . T4.1.1.8. LIS 3C2 - 82 A. 82B. and - Each RF0 Each RF0 High Water Level l T4122__ _B3C.B30 _, < _30 gallons __ j < - 38 gations (la monthsl (24 months) C015 T 31 1.8b. Scram Discharge Volume .
. I I T 4.1.1.B. LtS 302-82C 82D and .Each RF0 Each RF0 I N h 8** 'I T4_12 7. - .B3A,83B < -39 pallons 118 months) (24 months) C016
,' < - 38 gallons l T3.119. j no 4hange Main Condenser Low vacumn T4119. .Each RF0 proposed to remove T412.8 . ,LIS 503 A. B.C.D _ j >_- 23* Hg Vacumn{> - 23* Hg Vacumnll1B months)_ 18 months from TS T311.10. l no . change Main Steam Line High T 4.1.1.10. Each RF0 proposed to remove T41.2.10 .RM 1705 2 A. B. C. D , < - 7x badground : < -_7x hackground r ~ months) _ 18 months from TS
-t 1 T31.1.11. '
I Mair Steam Lire isolation T4.1.1.11. 2031 A.B.C.D (LS Si ; < - 10% Valve : < - 10% valve Durmg the RF0 .Durma the RF0 i T4.1.2.9 203 2 A.B.C.D (LS 6) Closure closure 118 months) .!?4 m_onthsE QJP _ _ _ _ T311.12. > -150 psig control > -150 psig conteol i Turbine Control Valve Fast d pesure at 'W press at Aceleraten 14.12.12 _ PS 37. 38, 39. 40 }Aceleraten relay Jelay__ Every 3 months Every 3 months 1NI35 _ , . _ . T311.13. Turbine Stop Valve Closure T411.13. I' f . and 1st Stage Pressure T 4.1.1.14. < - 10% Valve ' < - 10% Valve t i Permissive T 4.12.11, Closure (Dypass @ ' Closure (bypass @ 112 E ach RF 0 Each RF0 141213 Pts 504 A. B. C. D 305 psigt psign (18 months) (24 months) C018 Page 1
i
- instrument Summary by Technical Specification Section Attachment E Proposed j Current Surveillance Calculation kumher Techmcal Governing Requirement i or Dunhtative
) Existmg Technecal 2 instrument or Component Specification Associated Component Specification Proposed Technical Surveillance Calibration Justihcotion Function Referenre er Instrument Wumtier (Serpointi Value Specification Value Requirement intervat Previded (DJP) i RPS Section 3.1(continued) 4 l .) LPRM Signal Calibration i Every 1000 EFP ,Every 1000 EFP T4.1.2.5 na na na hours hous ne l . < - 10% Valve i
< - 10% Valve i i Reactor Pressure Permissive T4.1.1.15. PS 263 51 A 1. 5181 ' Closure (bypass p Closure (bypass @ 576 E ach RF 0 Isch RF0 ;
T4.1.2.14 51C-1. and 51D 1 600 psig) psign (18 months) 124 months) C003 i l j RPS instrument Response l ! Time Run Mode j T 4.1.2.14 _ ,re_ lay _ .. _ ___ _. 50 m, sec _ _ __ ;ns . _ _ _ , once m eek_ once seen_ _na ___ , i RPS instrument Response i 1' T4.1.1.15. ! t Time Other than run i T41214 telav 50 mser na onre tmeek onre meek na 4 l ( i ? i i 4 i I s i I I ,l 3 J l l 4 1
)
]l E
; Page 2 i
Instrument Sumnny ' oy Technical Specification Section Attachment E Proposed Current Surveillance Calculation Number l Technical Existing Techmcal Governing Requirement i er Qualitative l Instrument or Component Specification Associated Component Specification Proposed Technical Surveillance Calibration Justification l l Function Referorce er instrument IWumber (Setunint) Value Specification Value Requirement Interval Provided (DJP) , t 3.2 Protective instrumentatios ' Primary Containment isolation Functions (Table 3.2.A) [LS 26157A 1,57B 1 & l > - 9* mdrated !Each RF0 Esch RF0 Reactor low Water level T3.211 iSEA-1. SBB-1 I level > -11.7 inches (18 months) (24 months) lC010 T3.2.A1 l Reactor High Pressure T4.2A1 ]PS 261 23A.B < - 110 psig < - 110 pstg 0ncel3 months ;0ncel3 months na f Reactor Low-low Water T3.2A3. :LIS 2G57 A,57B & SEA, at er above -49' i .Esch RF0 iEach RF0 level T4.2A2 58B lindmated level !st or above!!18- months) indicate:llevel 46.3* !(24 monthal C009 T3.214, LS 263 574 2,57B 2, < - 48* indicated l <- 45.3* mdmated jEsch RFL lEach RFC l Reactor High Water level T4.213 5B A 2,58B-2 l level
- level i(18 months) :(24 months) 'C011 l 'Esch RF0 l lEach RF0 High Drywell pressure T3.215. iPIS 512A.B.C.D l < - 2.5 psig < -2.22 psig (24 monthsl CO26 l(18 months) _
l no .chanDe High Radiation main Steam ;Each RfD (18 proposed to remove , Line Tunnel T3.2 A 6 .RM1705 2A. BAD <- 7 background < - 7 buapround ! months) 18 months from TS . f l l Low Pressure Main Steam T3.2.A.7 :Esch RF0 iEsch RF0 Line T4.210 PIS 261-30A.B.C.D > - 810 psig > - 810 psig !(18 months) !(24 months) ! C001 T3.210 l <- 136% of rated Each RF0 l < - 140% of rated lEsch RF0 High Flow Main Steam Line T4.215 DPIS 261-2A through S steam flow tflow t(18 months) i(24 months) C019 Main steam line Tunnel Exhaust Duct High 13.2.A.9, Temperature T4.2.A 4 na < -170 F < -170 F Oncel3 months ;0nceT3 months na Turbine Basement Exhaust ! l Duct Hiph Temperature T 3.2._A.10 na _ < -150 F < -150 F Oncel3 months jonce:3 months .ns . Reestor Cleanup $rstem T3.2111 j < - 300% of rated < - 300% of rated l High Flow T4.217 ina flow iflow ,0ncef3 months 'Oncef3 months na ; I- l Reactor Cleanup system T3.2A12, l l High Temperature T4.218 !na : < - 150 F < - 150 F j0nce 3 months :Once!3 months ,na l ! f Reactor Low Lowwater T3.2A1.1, LIS 263 72A,72B,72C, et or above-49' .at or above - 46.3* Each RF0 ;Each RF0 , Level initiate CS & LPCI T4.2.B.1 l720 mdicated level !indmated level (18 months) !(24 months) -C004 C005 Reactor Low Low Water j l l Level. initiates Anto T3.2A1.2, ,LIS 263 72A,72B,72C,:at or above-49" ,at or above 46.3" ;Each RTO lEach RF0 blowdown T4.2.B.1 !720 mdested ievel !indmated ievel !(18 months) (24 months) C004
;tS 263 72A. 72B,72C.
l;s t or above .46.3* l,Each RF0 Reactor Low Low Water T3.2.B.1.3. !2D, & 72A.1, B-1,C 1,& at or above 49' level, initiates HPCI. RCic T4.2A1 'D 1 indicated level indmated level j(Esch r 10 months) RF0 !(24 months) :0004, C005 LS 263 72 A. 72B,72C,
!2D, & 72A 1, B 1,C-1.& l'at orabove abow", l l )
Reactor Low Low Water T3.22.1.4. ,at er 46.3* 49":Esch RF0 ,Each RF0 l l Level, Initiates D0 starting T4.2.B.1 'D 1 !mdmatolevel !indmated level !(18 months) !(24 monthst 'C004. C005 j I
,LS 26172 A 2, B 2 & LS l l Reactor High Water Level T 3.2.B.2.
l263-72 A 3. B-3 & LS e * (8" mdested < - + 45.3* Each RF0 ;Each RF0 (HPCI, RCIC) 14.2.B.1 2G72 A 4. B-4 level indested level 118 months) '(24 months) C006,C00',
> - 307* above jvesseltere-approx.
Reactor Lowlevel(inside T3.2.B.3, '2i3 core height (- Each RF0
>- 151* indicated jEach RF0 4 shroudi T4.2.B.1 LS 263 73A.B 175.5*) level l(18 months) :(24 months) CO27
't T3.2A4, PS 10018BA 1,B-1.C- 1.55 < - P < - :Esch RfD lEsch RF0 l Containment High Pressure T4.2.B.2 i1.D 1 1 < p < 2 psig 1.82 psig (18 months) 'CO29 J24 months)
,PS 1001894 2.B 2.C. l High Drywell Pressure, T 3.2.B.5.1, ;2,D1. & BB A 3. B 3. C-3. ;Each RFD lEsch Rf D Initiates CS, LPCI & HPCI T4 2 B.2 D-3 < - 2.5 psig < - 2.22 psic 418 months) 124 months) C030.C031 Page 3
I l Instrument Summery by Technical Specification Section Attachment E l Proposed l Current Surveillance Calculetion Ernbar l Technical Emsting Technical Governing Requirementi or Dualitatsve Instrument or Component Specification Associated Component Specification ProposedTechnical Surveillence Calibration Justification Function Reference or Instrurnent Number (Setpoint) Value Specification Value Requirement Interval Provided (DJP) 3.2 Protective Instrumentation ( continued) instrumentation that initietes er controls the core and containment coohng systems (Table 3.2.B) High Drywell Pressure, 73.2.B.S.2. - Each RF0 :Esch RF0 l ,C028 initiates Auto Blowdown T4.2.B.2 PIS 100189 A.B.C.D ! < - 2.5 psrg < - 222 psig !(18 months) !(24 months) - High Drywell Pressure. T 3.2.B.5.3, Each RF0 Esch RF0 [PS 1001-89A-2.B 2.C] l j initiates DQ start T42.B.2 .02 < - 2.5ysq l < - 222 pse_j(18 months) IQ4 months) .0030 High Drywell Pressure, j. initiates HPCI vacuum l j breaker containment T3.2.B.S.4. PS 1001-89 A-3.B-3.C. Each RF0 Each RF0 i isolation valve closure T4.2.B.2 I3.D 3 < - 2.5 psig < - 2.22 psig. !(18 months) !(24 months) C031
! l Reactor Low Pressure, CS & T3.2.B.6, ;PIS 263 52A 1.B.1 & PiS ;Esch RF0 ;Esch RF0 LPCI valve permissive T4.23.3 400 psg +.25 psig 00 psg +. 5 psq K18 months) ;(24 months) ;C045,0046 J261-52A.B Reactor Low Pressure, RHR T3.2.B2, mjection valve permissive 74.2.B.3 PIS'261-23 A.B < - 110 psig < - 110 psig l0ncel3 months 0nceI3 months na Reactor Low Pressure.
initiates CS & LPCI T3.23.8, T4.2.B.3
,PS 263 50A.1, PS 263-
- 50B-1,50A2 & B2 400 psg + 25 s9f4400 psg + 5 psig l Inch RF0 (18 months)
;Each RF0 c(24 monthst f:C042 C043 i 3
j i
! i Reactor Low Pressure. l l Prevents LPCI break T3.2.B.B. ; ,Eech RF0 f ach RF0 l
detection circuit actuation T4.22.3 PIS 263 49 AS & 50A.Bg 800 psg +. 25 psig 900_psig +.Spsig ;(18 months) I(24 months) lC040,C041 Reactor Low Pressure, T3.2.B.10. ;PS 2635CA3.B3 & PS Each RF0 Each RF9 l Isolates HPCI T4.2.B.3 :263 52 A-2.B 2 iC044. C047 100 > P > 50 psigJ0psg +.Spsig f(18 monthat (24 months) l ! T3.22.11, l l0nceicycle
! Oncefcycle i j
Core Sprey Pump Start Timer T4.2.B.4 14A . K14A & B 0 < t < 1 sec. (24 enenths) C075 J 0.21 < t < 1 sec l(18 months) T3.23.12, Once! cycle j0ncelcycle LPCI Pump Start Timer T4.2.B.4 l10A . K7CA & B_J< 1 < 6 sec. l 4.16 < t < 5.84 secy i months)C075 C4 months) T3.2.B.13, [8.5 < - t < - 11.5 j0nceleycle l l j0nce! cycle LPCI Pump Start Timer T4.2.B 4 110A K75A &B ! 8 < t < 11 sec I sec i(18 months) i(24 months) lC075 T3.23.14. j j > - 80, < - 120 [ > - 94.4, < - j0ncefcycle {0nceit ycle Auto Blowdown Timer T4.2.B.4 2E K24A & B .sec !115.6 set (18 months) 1(24 months) C032 ADS Drywell Pressure 73.2.B.15, 3E -K20 A & B. l 8 < - t < - 15.4 ;Esch Rf D ;Each RF0 l Bypass Timer T4.2.B.4 2E.K21 A & B ! 11 +. 2 mm. ' min !(18 months) !(24 months) ;C032 l RHR {LPCD Pump Discharge T32.B.16, j i Pressure interlock T4.22.5 na _ jl50 +.10 psig 1150 10psig ;0ncef3 months lOnce'3 months na CS pupm discharge pressure T32.B.17 j interlock l l lns T4.23.5 na 150 +- 10 psig 150 +.10ysig ;0ncet3 months ;0ncer3 months _ 20 25% of rated 20 25% of rated 13.2.B.10, l l l Emergency Bus Voltage i voltage resets at less voltage resets at < or Oncelcycle l0ncaicycle Retey(A5 & A61 T4.2.B.15 jl27-A5l1,2 & 127-A6l1.2 than 50% t- to 50% i(18 monthst j24 months) C033 l l0 volts between 0.86 Startup Transformer Loss of T3.2.B.18 jl27-5041,2 & l0nce! cycle 127- At 0 volts between l < - t < - 1.34;0ncefcycle Voltage Relay T4.2.Bf A 604l1.2 11.1 and 1.2 secondsjeco ds time deisy !(18 months) (24 months) CO22A 4 j l3868V +.C.5% w 8.2 j38782V +. 0.51% w I startup Transformer T3.22.20. ji27A-504:1,2,3,4 & +. 0.5 second time j1024 0.36 second l0nceicycle 0nce! cycle Degraded Voltage T4.2.B.6B 1127 A-604!1.2,3,4 l delay time delay (18 months) 3
;(24 months) .PS147 l 1 l
RHR, CS, ADS, HPCI, eral l 1 HPCI Trip System Bus Power T3.23.21,22,23, lCSCS relays ref. 'Once! cycle 0nce! cycle l 4 Monitor 24.25. T4.23.7 procedure BM2 2.10.10 ins . functmnat test 'na . functional test lf18 months) C4 months) .QJP Page 4
l l 1 Instrument Summary by Technical Specification Section Attachment E l I
! I l Proposed I Current Surveillance Calculation Number 1 Technical Faisimg Technical Governing Requirement i or Qualitative i i instrument or Component Specification Associated component Specification ProposedTechnical Surveillance Calibration Justification i Function Reference er Instrument Number (Setpointi Value Specification Value Requirement interval Provided (DJP1 l
Instrumentation that initiates or controls the core and containment cooling systems (Table 3.23 continued) , T3.2.B26 { ! j ! j Recirculation Pump A dip T4.2.B.B ___ na . _ ! < - 2 psd < - 2 psyd _ Once!3 months _ Onr.e:3 months jna _ { T32.B.27, ! l ; l Rscirculation Pump B dip T42.B.8 ns [ < - 2 psid < - 2 psid 'Oncef3 months Once'3 man.t.hs_ ,na __ __ , Recirculation Jet Pump Riser T3.2328, j { l l } l d!p A > B T4.2.B.9 :na ' 05 < p < 15 psidj 03 < p < 1.5 psed Oncel3 months Once!3 months na
~
Core Spray Sparger to T3.2.B29, no calibrate on f } ,0nce I Cycle Reactor Pressure Vessel dip T4.2.B.9 DPIS 1459A & B ] 11 -15)pse 1t + 15;psul )18 months) 18 months hne Condensate Storage Tank T3.2.B.30. l l > - 18' above tank > - 18' above tank ] Lowlevel T423.14 na tiero zero 0ncei3 monthsjncet3 months _ ;ns Suppression Chamber HiDh T3.2.B.31, ! i < - 1'11* betew l < - 1*11* below 1 l l ,
. Level T423.14 na torus iero torus rero oncet3 months Once!3 months na I RCic Turbine Steam Line T3.2.B.32. j j < - 300% of rsted j <- 300% of rated l !
l High Flow T4.2.B.10_,ns___ _ ._jsteam flow steam flow Once!3 months Once:3 months na RCic Turbine Compartment T3.23.33, l l l l Wall T4.2.B.11 na ! < -170 F : < - 170 F Oncef3 month Once/3 months 'na T3.2.B.35. , I i I l Torus Cavity Exhaust Duct T42.B.12 'ns ! < - 150 F < - 150 F 'Oncet3 months !0ncef3 months 'na T32.B.36. j j l l l d RCIC Valve Station Area Wall 14.2.B.13 na ! < - 200 F i < - 200 F Oncef3 months Oncel3 months ns l T3.22.37, RCIC Steam Line to pressure T_4.2.B 10 j n a _ _ _ _ _ _ __ j l l l
! 100 > P > 50 psig ! 100 > P > 50 psig Once!3 months 'Oncel3 months
]_
'na i
HPCI Turbine Steam Line T3.2.B.38, l { < - 300 % of rated l < - 300 % of rated l l High Flow T4.22.10 na flow flow _ Once!3 momhs lOnce:3 months jna l HPCI Turbine Compartment T32.3.38.
\ l Exhaust Ducts l < - 170 F
{
! < - 170 F j l
- Once!3 months Oncel3 months na f'
T4.2.B.11. _ .na j T3.2 B.39. j j j Torus Cavity Exhaust Duct T4.2.B.12_ na_ j 190 200F 190 200 F Oncel3 months Oncer3 months no HPCIl RHR Valve Station T32.B.40, j
; j Area f xhaust T4.2112 'ns i < - 170 F < - 170 F Once 3 months Once?3 months na Page5
i instrument Summary by Technical Specification Section Attachment E t i Proposed Current Surveillance Calculation Number Technical Existmo Techment Governmg Requirement i or Qualitative instrument er Component Specification Associated Component Specification PropcsedTechnical Surveillance Calibration Justification a Function Reference or Instrument Number (Setpointi Value Specification Value Requirement Imervet Provided 10JP) 5 instrumentation that Monitors Emergency Bus VoltageTable 3.2.B.1
- ' i
)
I I i i Emer0ency 4160V Buses AS l '3950V + 0.5% with i3950.5V + 0.5% ! {
-[ 1 i
& A6 Degraded Voltage T 3.2.B.1.1, ;127A
; A5:1.2 & 9.2 + 0.5 sec time 0.24% with 10.24 +- Once i Cycle Oncef cycie j Annunciation T4.2.8.15 127A ABl1.2 ' dela y 0.36 see time delay (18 monthst (24 monthst 'PS148 l
i i r 1 l t 1 a l a Page 6 l
instrument Summary by Technical Specification Section Attachment E Proposed Current Surveillent.e Calculation Number Technical Esistmg Tec hnical Governing Requirement i or Quahtative Instrument or Component Specification Associated Component Specification PropusedTechnical Surveillance Calibretion Justification Function Reference er instrument Wumber (Setpointlif akes Specification Value Requirement Interval Provided f0JP) Instrumentation that initiates Rod Blocks (Table 3.2.C.1) T3.2.C.1.1. l I l l f APRM Upscale (flow Bias) T4.21.2_ _ na na :na -Oncef3 months Oncel3 months na T3.2I.1.2. l l l l l l APRM Upscale T4.2.C.2 :na 4na _ _ na__ ;0nceI3 months _ j0ncel3 months na T3.2I.1.3. j i l { l APRM inoperative T4.2.C.3 'na jna Conditional Condnional ins
. [ns T3.2.C.1.4. j j j l l APRM Downstale T4 2.C.1 na na ;na ;0ncef3 mo_nths Oncef3 months 'na I
Rod Block Momtor i Power T3.2.C.1.5 j
} l:na l Dependent) .T4.24.7 na lns Oncel6 months Once!6 months Fna Rod Block Monitor T3.2.C.11 l l j j l Inoperative T4.21.9 'ns 'na 'na : Conditional ' Conditional lna Rod Block Monitor T3.2.C.1.7 l
Downscale .T4.2.CL_ na _ na jna a l0ncal.6 monthsj0ncef6 months f na ; Stanup or Startup or i ! l T3.2.C.1.8 Controlled Controlled l l f IRM Downscale .T4.21.5 na na na iShutdown , Shutdown na IRM detector ret in Startup T3.2I.1.9 j Position .T4.2.C.14 l na .ns l l l l na
.n s_ _ ,na_ ,na
' Startup or
! Startup or T 3.2I.1.10 Controlled li [ Controlled IRM Upscale .T4.2.C.4 l Shutdown Shutdown T3.2I.1.11
{ l l l j IRM inoperative .T4.2I.6 ra na na !Conditenal Conditional 'na SRM Detector notin Startup T3.2I.1.12 i { l l l l Pos: tion .T4.2I.12 'na ns :na .na 'na !na l ;Startup or ;Startup or T3.2I.1 13 l !Controled .Cornrolod i i SRM Downscale .T4.2.C.13 na m na na Shutdown ) Shutdown ne j
] ;Startup or .Startup or I T3.2.C.1.14 l Controlled l.na l Controlled SRM Upscale .T4.2.C.10 na na Shutdown Shutdown na T32.C.1.15 f l 1 l
.' i I
! I SRM inoperative .T4.21.11 ns ny na jConditional jConditenal na
! : I I j Scram Discharge lnstrument T3.21.1.16 '
; j l Volume Water tevel, High .T4.2I.15 1S 302-B2E. 83E na na Refuel Refuel C017 Scram Discherge instrument Volume . Scram Trip T 3.2I.1.17 l I l
[
- Bypassed .T4.2.C 16 ,na na l Oncel3 month
-na Oncel3 months .ns j
Recirculation Flow Converter T3.24.1.18. ; Upscale
} l T4]I.17 _ na na na 1 Oncel3 months 'ns ;
Recirculation Flow Converter T3.21.1.19
. inoperative na j
na na l0ncel3 Conditenal monthsl _f 'ne
.T4.21.19 _ C 4 ondamnal l Recirculation Flow Converter T3.2.C.120 l
_ l I I Comparetor Miamatch .T4.2.C.20 na , na ;na Once!3 months 'Oncel3 months 'na FC 27 A. FC.ZB B. FY i
~
l l 260-BA.B & FT 2G1 i l ; r Recirculation Flow Process 6A BI.D & SQRT 260 l j i ;0nceicycle instruments T4.2 C.21 9A.B C.D 'na 'na Oncelcycle f 24monthal :CO20.0021 Page 7
l l l i i instrument Summary by Technical Specification Section Attachment E 1 l I Proposed Current Surveillance Calculation Number Techmcal Existmg Techrucel Governing RegunementI or Ouststatsve instrument or Component Specification Associated Component Specification Proposed Techmcal Surveillance Calibration Justification Function Reference or instrument Number (Setpoint) Value Specification Value Requirement Intervat Provided (0JPl Control Rod Block Instrumentation Setpoints (Table 3.2 C 2) l' l T32.C.2.1. i j , 3 l APRM Unscale 14.2.C.2 na ' Limas m COLR : Limits m COLR _ ._;0nce!3 months Once!3 months 'na T3.2.C.22, j
, { ;
APRM inoperative T4.2.C.3 na .na no Conditmnal Condamnet na T3.2.C.2.3. j > - 2.51ndicated on i > - 2.51ndicated on l ; j APRM Downstale T. 4. 2. C 1.-_._,na._._ 4 Scale . -._ ..- --, Scale .Oncel3 m_onths .q'O.n.ce:3 mo.nths 'ns _m -- - Rod Block Monitor (Power T3.2.C.2.4 l t l l } l dependent) 14 2.C.7 na Limits m COLR tumts m C_OLR_ -.. 'Once'_6 m_on_ths
-.y- !Once:S months
- ria Rod Block Monitor T3.2.C.2.5, l l l l f inoperative T42.C.9 na _ _ _ _ _ na __ _ _ na Condamna! lConditmnal . na _ _ _ _ , !
Rod Block Monitor T 32.C.2.6, i j j l Downscale T4.2.C.8 'na Lumts in COLR Limasm C0tR Oncel3 months Once'3 months na ;
! Startup or ;Startup or j T 3.2.C.2.7, 1 . > - 51125 of full > - 5/125 of full Controlled Controlled i
IRM Downseale j T4.2.C.5_ _ ina___ _ _ scale _ _ scale Shutdown Shutdown ' ne IRM Detector not in startup T3.2.C2.8. j i j l ; l l ; j position T4.2.C.14 na na na na na na l jStartup or Startup or l T3.2.C.2.9. ; < - 1081125 of full l < - 1081125 of full Controlled Controlled ; IRM Upscale T4.2.C.4 na ,na_ _ _ . scale ___ scale _ _ _ _ Shutdown ___ Shutdown _ T32.C.2.10, i ! j I l i > IRM inoporative T4.2.C.8__ . ;na _ na na Conditional Condnenal 'na i SRM Detectot not in Startup T3.2.C.2.11, i ! ! i < 1 Position T4.2.C.12 na na na ns ins r,a ,
; ,Startup or Startup or l T3.2.C.2.12, ! l Controlled I l . Controlled SRM Downsesle T4,.2.C.13 na 1 > - 3 counts I sec l > - 3 counts J sec ' Shutdown Shutdown na ,
! ! Startup or Startup or i T3.2.C.2.13, i < - 10 5th < - 10 5th : Controlled Controlled
{ SRM Upscale countsjec_ _ T4.2.C.10_ na, _ _ _ _ _ counts:sec _ Sh.utdown_ Shutdown _ ne__ _ j T3.2 C.2.14. l { } , SRM inoperative T4.2.C.11 na na n Condamnal ' Conditional na Scram Discharge Instrument T3.2.C.2.15. ' j r Volume Water level . High T42.C.15 ,LS 302;B2E. 83[_ . , _ _ ; < - 18 gallons < - 17 pallons Fefuel Refuel C017 l Scram Discharge instrument ( l
.J l
l Volume Water level Scram T32.02.16. !
) l Trip Bypassed T4.2.C.18 ne _
na __ _ ns Oncel3 monihs Onca'3 menths 'na Recirculation Flow Converter T3.2.C.2.17 j < - 120/125 of full l < - 120l125 of full l l ! Upscale T42.C.18 na scale Oncel3 months Once'3 months lna Recirculation Flow Converter T3.2.C.2.18, (scale ___ __ i i i ! I i _ _ _ _ _ ;n s _ __ na_ i t 1
- I'"P ative T4.2.C.19 __ ns Oncel3 months Once'3. months na
! < - 1C% d > B0%
- s
; rated power. < -
l l Recirculaticn Flow Converter T3.2.C2.19, !15% if < - 80% 0nce l Cycle ( OnceiCyc6 ; Comparator Mismatch T 4.2.C 20 ,FC 27 A.FC 78 B rated power < - 0% !18 months 1 (24 months) 0020.0021 1 1 Page 8 !
Instrument Summary by Technical Specification Section Attachment E Proposed Current Surveillance Calculation Nwnher Technical Existmg Technical Governing Requirement i or Qualitative instrument or Component Specification Associated Component Specification Proposed Technical Surveillance Calibration Justification Function Reference or Instrument Number (Setpoint) Value Specification Value Requirement Interval Provided 10JP) Radiation Monitoring Systems that initiate and/or isolate (Table 3.2.D) Refuel Area Exhaust ; f , Momtors .T34D1,,_ ___ n t _ _ upsca < 100 mehr ;Uptge, < 100 mdhr_,0ncel3 months Once!3 months na Refuel Area Erhaust i ; Monitors T.3.2.0.2 na Downstale Downstale Oncef3 months Once'3 months na l l l 1 Page 9
Instrument Summary by Technical Specification Section Attechment E 1 j Proposed
; Current Surveillance Calculaimn Number ; Technical Esistmg Techrucal Governin9 Requirementi or Qualitative lustrument or Component Specification Associated Cornpanent Specification Proposed Technical Surveillance Calibration Justification Function Reference or Instrurnent kumher (Serpoint) Value Specification Value Requirement Interval Provirled (0JP1
) j Surveillance Instrumentation l Table 3.2.F)
! T3.2J.1. j l 1 ,Each RF0 ;Each RF0 [
j Reactor Water level 14.2J.1_ _ ;64a29A & B__ _ _ indmator 0160* Indmator D 6L _j(18 men.ths) (24 months C062 _ i T 3.2J.2. l l ;Each RF0 ,Esch RF0 i l j Reactor Pressure T_4.2J.2_ _ _ ,640__25 A&B t __jindmator 01200 psig indicator 01200 psig 118 months) :(24 months 1 C063
' l T3.2.F.3. ,
; Each RF 0 jf ach RF0 i
$ Drywell Pressure T4.2 F.3 !TRU 9044,45 Recorder -80. psia Ruorder .80 psia 118 months) 1(24 monthat
- 13.2J.4 '
l
',C052
; Recorder. Indmator D. Recorder. indrator D. l i Drywell temperature T4.2J.4 TRU 9044. TI 9019 400 F !400 F i0ncef6 months Once!6 months 'na j Suppression Chamber Air 13.2J.5. ,Ruorder.Indmator 0. Recorder.lndicator a f j j Temperature T4.21.2.5 TRU 9045. Tl 9018 400F 400 F lOncel6 months Once!6_ months na _
j Suppression Chamber Water T3.2J.6. j l i l Level T4.2J.6 LR 5038. LR 5049 Recorder 0-32' l RecorderlOnce!6 432*monthsjnce'6 months na j T 3.25.7, j 2B Volt Indmatmg j28 Volt Indmatmg l Lights l l l Control Rod Position T4.2J.7 na jLights _ ;na gna na p Neutron Manitoring T 3.2.F.8. T4.25.8 j na
-SRM.lRM.LPRM O jSRM.lRM.LPRM O ; Controlled
!to _ to 100% power
; Controlled f.
jna j Shutdown jShutdown Suppression chamber Water ;TI-502101 A.1RU 5021d _100% power {na l Temperature 3.2J.9 11 A !30 230 F IBuk!LocaD 30 230 F (Bukflacali na na j Suppression Chamber Water ' il 5022-018. TRU 5022 ] } } l.no Temperature 3.2J.10 01B '30 230 F (Bukflocair 30 230 F (Buktlocal) na na Drywelll Torus Ditterential 13.2J.11 I l l l l 4 Pressu o T4.2J.9 PI5021 ; 025,3.0 psig -0.25, 3.0 psig l0ncet6 months :Oncef 6 months na j Dywell Pressure, Torus Pressure T 3.2J.12 l l -025. 3.0 pseg.1.0 025,3.0 psig.1.0 ] Oncef6 months Ontat6 months na l l { T42J.10__ _ ;PI 5067,A.B __ _ J2.0 psig_ _f .02 psg, l T3.2J.13. j ;Esch RF0 ,Each RF0 i Setetyfrelief valve position T4.2J.11 :Z! 203 3A.B.C.0,4A.B 'na
, l
_ _ jna !(18 months) '(24 months) f,0054 Safety Valve position T32J.14 IE & 2T 203 3A.B.C.D ! ,Each RF0 Each RF0
} l Indicator tprimary)
Tail pipe temperature T4 21.12 _]4_A. 4B __ _ _ _,na __ ja___ (18 monthsl J24_ months) C054
- 13.21.15 lTE 6271. 72. 73. 76. 74- l Each RF0 l lEsch RF0 !
- indication (secondaryl T4 2J.12 B. 75 B , ____na________ na _ _ _ ;(18 months) '(24 months) C064 j Torus Water level wirls T3.2J.16, Each RF0 i j ,Esch RF0 l range T4 2J.13 Li 1001604 A.B i0300"H2O i D 300" H2O !(18 months) 124 months) .C005 Torus Water level Wide Range T3.21.17 T 4.2J.13 LR 1001604 A.B f
] 'Esch RF0 'Esch RF0 l
4300" H20_ _ J300* H2O [118 months)]24 monthsL:C006 Containment Pressure high T321.18. ;Esch RF0 :Esch RF0
, l l range T42J.14 _ ,PI 100140CA,B_ _iG 225 psig S225 psig 118 months) (24 months) 'C067 Containment Pressure - High 13.2J.19. Each RF0 l l l jEach RfD l PR 1001600A,B
- Range T4.2J.14 i O . 225 psip . l 0 225 psig j(18 months ( _j24_mo_nths) CM8_ _ _
Containment Pressure Low T3.2J20. ! Each RF0 ,Each RF0 l l 5 to 5 psig__,(18 months) lC077 l Range T4.2J.14 Pl 1001 601 A.B i-5 to Sysig i- '(24 months) Containment Pressure Low T3.2J.21, ) i
;Each RF0 :Esch RF0 l
Range T4.25.14 .PR 1001601 A.B f S to 5 psig -S to 5 psig (2,4_ months) jt18 months) ]C069 Containment High Radiation ( T3.2J22. Each RF0 j RIT.1001 - 606 A B & RR l jEsch RF0 drywel0 142.F.15_j1001 606A B_ _ 11e 1:10 7th 11e 1:10 7th 118 months) d24 months) lC055.C056 1 T32J.23. . Rl 1001 - 609 RR i } Each RF0 Each RF0 j Reactor Building Vent T4.2J.16_ {1_001 60B__j 10 11e 10 4th Rfhr l 101 to 10 4th R!hr '(18 months) l(24 months) ;C057.0059 T 3.2J.24. .Rl 1001 608 RR { ;Esch RF0 ;Each RF0 l l Main Stack went T4.2J.17 1001 608 124 monthsl iC058,0059
} 101 to 10 4th RIhr q 101 to 10 4th Rthr !(18 months)
T3.2J25 RI 1001 610 & Each RF0 Each RF0 l Turbine Building vent T42J.18 '1001 608 RR { 101 to 10 4th R!hr 1101 to 10 l 4th RThr 124 months) C060I(18 C059 month) Page 10
instrument Summary by Technical Specification Section Attachment E l Proposed > Current Surveillance Calculation Numbo. I Technical Existmg Technical Govermng Requirement l or Duahtstive 7 4 instrument or Component Specification Associated Component Specification Proposed Technical Surveil:ence Cahbration Justification F uretion Reference or Instrument number (Setpomt) Value Specification Value Requirement Interval Provided (DJPI instrumentation stat initiates Recirculation Pump Trip and Alternate Rod insertion (Toble 3.2 Q
'Esch RF0 f ach Rf D l
f High Reactor Dome Pressure T3 4.2 G PIS 263123A.B,C.D_ _4117_5 + 15 psia _ jl175 . 5 psi;_. ;t1B monthsl (24 months) C025 towlovv Reactor Water ' > - 77.26" above ! > - 48.3* indicated f ach RF0 Each RF0 ! Level T3'4.2.G LIS 263121 A.B C.D TOAF icvel (18 monthsl t:4 months) 'C024 I 2 7 -l i 1
)
A
)
l r l l l Page 11
l l i i instrument Summary by Technical Specification Section Attachment E , Proposed Current surveillance Calculation kunter Technical ExistinDTechmcal Governing fLequirement I or Qualitative lustrument or Conporant Specification Associated Component Specification Proposed Technical Surveillance Calsbration Justification Function Referent.e or Instrument Nunter (Setpointi Value Specification Value Requirement Interval Provided (DJP) Dryvvell Temperature Surveifiance instrumentation ( Table 3.2.H) '
'Esch EF0 Each RF0 Above 40' elevation T3 41H.1.2.3 .TE 505CA. BAD E.&F na na (18 months) (24 months) C061 E ach RF 0 f ach RF0 j l
Below 40' elevation 13 4 2 H 4 5 TE 5050G H.J K na na (19 months) (24 months) C061 l h i j ! I a l l l I 1 l Page 12
Instrument Summary by Technical Specihcation Section Attachment E a j Proposed j Current Surveillance Calculation Number Technical Existmp Technical Governing Requirement l or Quahrative Instrument or Component Specification Associated Component Specification Proposed Technical Surveillance Calibration Justifiention Function Reference or Instrument Number (Setpointi Value Specification Value Requirement interval Provided (DJPl ECCS Section 3.3 Reactivity Margin Core Ref uel, pnor to ; Refuel pnor to i l toading 3,4.3. A.1 ne na startup QJP
. no__ _ _ _ _ q'st a rt up Reactivity Margin irep i I i l control rods 3,4.3. A.2_ na
_ _ __na__ _ _ ne _ _ __ _ _ 0nce' week _ 0_ncel week ___jaa l l Subsequent to t Subsequert ie control Rods 3.4.3.B _ _ na_ _ , _ _ _ _ ,na ______na____ __hF 0__ RF,0 QJP
! Followmg each .Followmg each Scram insertion Time 3:4.3[ _ ,n a _ _ _ _ _ __ _ ,n s ne _ _ _ RF0_ ___ RF.0_ ._ QJP_ __ _ __
i 1 l l > i l ControlRod Accumulators 3g_.3.0_ 4 _ c a _ _ ___ _._ ,n a __ __ __ ;na __ _ _ j0nceish_if t,_ l j ;Followmg each OncelshjL__}na Followmg esch Reactivity Anomoties 3,4.3.E na na na RF0 RF0 QJP l Each Refuehng Each Refuelms no revsed m TS Scram Discharga Volume 3:4.3.G -na lna lna Interval [ Interval ' Amend.149 l 3 4 Standby Liquid Control i lInp full open < - 'inp full open < - l l l 4 Normal System Availability, j1800 psig. resent 1000 psig, resent ] Isch Refueling Each Refuelms ;no revsediniS rehof valve trip 3 4.4 4.2a 'na 11275 psig 1275 psig Int erval_ laterval j Amend.149_ _. Normal System Availability, 1 i > 39 GPM @ 1275 l > 39 GPM @ 1275 ,Each Refuelmg Each Ref ueling 'no revised m TS Pump Flavv Rate 3: 4.4. A.2b na psig ___ psig._ ' Interval interval ' Amend.149 l l l ! Both m the course Both m the ] ] l ! j Normal System Availabihty, ! of 2 refuelmg course of 2 [ manual initiation 3 4.4.A.2c na 'na 'na _ intervals __ refuelmg Intervals na _ __ Surveillance veth inoperable I Components 3_:4 4.B.1 _ __ lna _ _ ___ na _ _ _ _ _ na _ ___ _ __ _,immediate unmedi_a_te__ __ lna __ _ __ l ( When Baron s When Boron s 'added & each ! Sodium Pentaborate Solution 3 4.4.C.4 na 'ns 'na 'added & each RF0 RF0 '0)P 1 1 I 1 4 Page 13
1 i i Instrument Summary by Technical Specification Section Attachitwnt E i 4 Proposed j Current Surveillance Calculation Number
, Technical Iristmg Technical Govermng Requuement 1 or Quahtative i
{ instrument er Component Specification Associated Component Specification Proposed Technical Surveillance Calibration Justification j f unetion Reference or instrument Number (Satpomti Value Specifiention Value Requi ement Interval Provided (DJP) I 3.5 Core and Containment Cooling ! J , j CS System Testing. l l l i simulated automatic j ;0ncalcycle Onceleycle actuation 3 4.5.A.1.s _ , {n a _ _ _ _ _ __ _ na_ _ __ na _ _ _ _ _ , ,08 months) _ ,,24_ ( months} [QJP _ _ _. ._ , LPCI System Testing, ! , simulated automatic Onceicycle 'Oncazcycle l , 1 actuaten 3.41 A_.3.a _ _ ,na _ _ _ , _ __ _ _ ,na _ ,ns, _ _ _ _ _ ___ ,08 monthsL J24 mo_nthst QJP _ _ _ _ i I { f Contairunent Cooling Systern, ; ! l a air test drywell and torus ! ! [ headers & nonles 3;4 5.B.c 'ns 'na _ _ , _ _ .. ns _ _ _ _ _ _ Oi!ce'5 years _ ,0nce!5 years ns _ _ _ _ _ . _ 3 i I t HPCI System, simulated ; j Once! cycle ,0nce! cycle j d sutomatic actuation 3_4.5.C.1.a _.no _ _ _.____,na _ _ ___ na 08 months) '(24 months) QJP _ _ __ __ HPCI System, flow rate at l 'Once! cycle Oncelcycle I , 150 psip 3 4 5.C.1.e . .ns na na (24_ months) _ _ .. ,0JP_ _ _. t 08 months [__ l_____,I i j i {i RCic System, simulated 1 .0ncercycle ,0nce/ cycle automatic actuation 3 410.1_.a ___ _ na ,_ _ __ _ ___ n: _ _ _ _ _ _ na _ _ _ (18,msnthsl_ (24 months)_ DJP_ _ _ j t l 1 ! ! RCIC System, flow rate at i j .0nce rcycle ;0nce: cycle j 150 psic 3,45.D.1.e _,,ns _ _ _ ___ na _ _ _ __na __ _ _ jtl_8 months)_ ,_ _ J24_monthsi_ _ 0)P _ _ _ _ ADS, simulated automatic 1
.Onceleycle ,0nceitycle I a:tuation 3.4 5 E.1.a na na -na '(18 months) (24 months) OJP , A05, manually open relief ,0ncaicycle ;0nce! cycle i ,
valves 3 44L1.b na na na 'l18 monthst '(24 munths) QJP 1 DieselGenerator Availability j-
, l linoperable) 3,41F.1 na na as Conditional Conditional !na i
[ Mamtenance of filied ,i !. l i l discharge pips - press switch ' r 1 I cal 34.5 H na na na Every 3 mrnths Everv 3 months 'ne l l s d d I I Page 14 i
Instrument Summary by Technical Specification Section l Attachment E
)
Proposed Current Surveillance Calc ulation Number Techmcel f ristmg Technical Governing Requirement i er Qualitatrve insts ument or Compoient Specification Associated Component Specification Proposed Technicet Surveillance Calibration Justification Fuertion Reference or Instrurrent humber (Setpoint) Value Specification Value Requiremerit interval Provided (DJP) 3.6 Primary System boundary Thermal & Pressure limitations 3-4.6. A ,na na na
,dady (15 mm) _ ,dady (15 mmt ,na Coolant Chemistry 3.4.6 B na na na ,0nce' month ,0nce month na Coolant Leakage Drywell Once Cycle Sump Momformg 3 4 6.C.2.a2 ,na ,na na
,0ntsl18 months ( months 1
, 24 ,GJP CoolantLeakage Drywell Once7C ycle atmos rad monitor 3,4 6.C.2.b.3 ,na na na .,0nce 18 months ,24
( monthsi ,0JP Safety & Relief Valves Per TS section Per iS sectwn Checked or replaced 3,4 6.D.1 ya ,no , na ,3.4.13 ,3 4.13
.. , na Safety & Relief Valves-Each RF0 f ach RFC disassembled & inspected 3 4.6.D.2 na ,na na _ (18 months) , (24 monthsl , QJP Safety & Rehof Valve-matrument calibration 34.6.0.4 , per TS T4.2.F _ _ ,per TS T4.2.F _ , er p TS T4.2.F , per TS T4.2.F lper TS T4.2.F_ _ 0054,0064 Jet Pumps 34611 , na ,na na ,Dady_ ,Dady_ ,na Jet Pump FlowIWhsmatch 3.461.2 ,na ,na ,na ,Dady ._ Dady ,na Structural integnty 3 4.6.G y a __ . ,na ,na Shock Suppressors, See ISI program See IS! program ,na inaccessible hydraubc &
mecimnical 3 4.6.L1 na na na Shock Suppressors, See TS schedule See TS schedule ,0JP Once! cycle Oncetcycle functional test 3 4.6.t.2 ya ya .no ,08 months) ,C4 months) Shock Suppressors . Snubber _0JP Once cycle Once cycle Service Life review 346.l.2 na . na na (18 monthd (24 monthsl QJP Page 15
i I i 1 instrument Summary by Technical Specification Section Attachment E l I Proposed l Current Surveillante Calculation IWumher Technical Existmg Technical Governing Requirement i or Quahtative instrument er Component Specification Associated Component Specification Proposed lechnical Surveillares Calibration Justification , Function Reference or instrument IWumber (Setpointi Value Specification Value Requirement interval Provided f0JPl l 3,4.7 Containment systems . I Suppression Chamber - ! 'f ach RF0 f ach RF0 I 1 visualinspection 3:4.7. A.1.e 'na na na ' i 18 munths) Q4 months) _ QJP _____ Dryvvell & Torus - visual Fach RF0 Each RF0 j mspection 3.4.7. A.2.d _ ca __ ria _ . na ._. . _ _ _]I18 months) _ ;Q4_ months) QJP_ _ , _,_ , Primary Containment - Cont. ; Atmos Dilution Sys. ; ,0nce! cycle 0nceicycle j functional test 3 4.7.A.7.a _ _ ;na na na (18 months) C4 months) OJP_ _ _ _ 2 i SBGT test fibers and l Once! cycle Adsorber banks, delta P 3n4.7.B t.a.1 ,na__ _ __ n a __ _ _ na _ ^ Oncel18 months C4 months) QJP_ _ _ , SBGT testinlet heaters 1 ;0nce! cycle f14Kvd 3 4.7.B 1.a.2_.. ,na _ _ lna _ _ na Oncei18 months :Q4 months! 'QJP j SBGT system test filter Oncelcycle eff ciency 3 4.7 B.1.s.3 na na na Once 16 months C4 monthsl CJP [ SBGT system test auto , Once!cycie i mitiatmn 3 4 7.B.1.a.4 na na na Oncel16 months Q4 months) QJP CRHEAFS testpressure j ; ,0nce! cycle l drop 3 4.7.B.2.s ,na ;ns. .___ _ ina_ _ _ _ ___ j0ncellBmonths C4 monthsL ,QJP _ _ _ _ CRHE AFS - test filter j efficiency, heater & inst. , ! 0ncelcycle function 3.4.7.B.2.b.1 na
. na _ __ na _ _ __ Once3Bmonths [C4 months)__ QJP, _ __
1 1 . i i I CRHE AFS - demonstrate inlet i l j Once! cycle l heaters operable (14Kuvi 3 4.7.B.2.c na na 'na Oncen8 months C4 monthst QJP CRHEAFS performinst. I i Once! cycle j functional tests 3 4.7.B.2.d na na na Oncen8 months 'C4 months) 'QJP Each Rf D G4 Secondary Containment - j f ach AFD. prmr to , months). pnor to containment integrity 3 4.7.C.1.c na na na refuelmg refurimp QJP Page 16
I I Instrument Summary by Technical Specification Section Attachment E I Proposed Current Surveillance Calculation Number Technisal Existmg Techmcal Governing Requirementi or Quaktative
- instrument or Component Specification Associated Component Specification Proposed Technical Surveillance Calibration Justification Function Reference or instrument Number (Setpoint) Value Specification Valoe Requirement Interval Provided (DJP) 1 3 8 Radioactive Etfluents
, 1
! ' Bat ch, m eekly. . Batch, m eekly. j Liquid Effluent concentration 3:4.8.4. T4.8.1 na 'na .na monthly, grtly.__ monthly grily 'na l L Radioactive liquid Effluent 3 4.8.B. T3.8.1.1 no calibtate on-
[ j Gror.s Radiation Momtors T4.8.2.1.a_ __._ RM.1705 30 _ _ _ jn a ___ _ _ .ns Once! cycle 18 months line i i t Radioactive liquid Effluent 3.4.B.B. T3.8.1.2. ! no . calibrate en i Flow Rate Instrumentation T4.B.2.2.a FR 7214A.B 'na 'ne Oncetcycle 18 montha line
, I 1 l' Liquid Radweste Treatment 3:4.B.C. na 'ns '. n u Oncef31 days OnceI31 days na j
~
Contmuous. Conimunut ' meekly. monthly,q weekly, monthly.q Gaseous Effluents Dose Rate 3;4.8.DJ B.3__ na_ _ ___ na __ na cartetty 'uarterty ' na , I ! Radioactive Caseous Effluent 3.4.B.E. 73.8.2.1 I l Once: cycle no calibrate om j Main Stack (noble gas) l T4 B.4.1 RM 170516 A,& B ' na 'na 118 months) 18 months line f' f Radioactive Gaseous Effluent l } Main Stack (effluent flow 3:4.81 T3 8.2.1 ! 'Once! cycle no calibrateon l = ratel T4.B.4.1 FR 8368 na 'na (18 months) 18 monttts line Radioactivs Gaseous Effluent l i Main Stack (Samp!s Flow 3 4.81. T3.8.2.1 - I l ;0#.celcycle no calibrateon 8 Ratel T4.8.4.1 ' S 2247-Fi.1 na na 't18 monthsl 18 months line
' l Radioactive Case <us Effluent 3.4.81. T3.B.2.2 ;0nce! cycle no calibrateon
), RB Vent (noble gas) T4.8.4.2 RM1705 32A & B 'na 'na '(18 months) 18 months line i Radioactive Gaseous Effluent 3 4.8.E, T3.6.22 Once! cycle no calibrateom RB Vant lefflunni flow rate) T4.B.4.2 FI 8116A & B na na .(18 months) 18 months line 4 i i t
! i f Radioactive Gaseous Effluent 3 4.81 T3.8.2.2 ; , no -calibrate on
,0nce! cycle r
. RB Vent (Sampler flow Ratel T4.8.4.2 6 2264-Fi-1 na ns . (18 months) 18 months line t 3 4 t
} }
1 ' l 4 ! 1 l i ! j Radioactive Gaseous Effluent 3 4.8.E. i3.8.2.3 500,000 micro 500,000 micro onceitycle 'Once. cycle i i SJAE T4.8.4.3 ,RM1755 3A. B cunes'sec cunesisec ____J18 months) (24 months)____ _ 0051 _ _ r i l Radioactive Gaseous Effluent A4.8L T3.8.2.4 - l ; i 4 Hydrogen Monitor T4.B.4.4_ _ n e_ _ __ _ _ _ na __ _ _ _ 1 na___ _ quarterly quarterty na
. { } l i I Gaseous Effluent Treatment 3 4_8.F. T3.8.2[na_ _ _ _ _ na _ _ _ . _ na ___ _ __ _ Contmuou_ sly Contynuously na _ _ _ __
i ) i i Main condenser 3,4.8.G na na 'ns 'Once!31 days Oncel31 days jna _. _
,0nceltycle ;0ncelcycle {
Mechanical Vacuum Pump 3 4 8.H 'na na na (18 months) 't24 months) 'C013 QJP l l Page 17 1 i
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Instrument Summary by Technical Specihcation Section Attachment E Proposed Current Surveillarco Calculation Illumber Technical Existmg Technical Governing Requirementi or Qualitative instrument er Component Specification Associated Component Specification Proposed Techruce! Surveillance Calibration Justification f unction Reference er Instrument Number (Setpointi Value Specification Value Requirement intervat Provided toJP) l l Auxiliary Electrical Systems l Aux Electrical Eq - DG start & 'Oncetcycle Once! cycle ! ! load 3 4.9 A.1_.h 182 509. 60B__ na __ ___ _ jna _ _ _ _, __ j(la monthsi_.,C4 months! C073,0JP _ _ _ j Aux Electrical Eq. DCload , i Once> cycle ,0nceicycle . j transfer 3_4.9. A.1.c_ ;162-501.601 _ _ _ _ 127 14 seconds _ _ J11J 13_.2 sec_ _J18 months) _. C4 months C074._QJP _ ___ Aux Electrical Eq . Station & ! ( { ! Svvitchyard Batteries- .Oncercycle Once.icycie l i I discharge test 3 4.9.A.2.c ' na na na f _ _ _ jl1B monthst_ 'C4 months!_ __ QJP _ i I EmerDency 4160V Buses A5- j A6 Degraded Voltage 127 A- A5/1.2 & j ,0nt-e! cycle 0ncelcycle j Annunciation 3-4.9.4.3.a l127A AS 1.2 'na na '(18 months) 'C4 months) C070 RPS Electrical Protection , Once! cycle Arweliee 309A41 ra ne na '(15 months) 18 months no calibrate online I J t (( t 1 l 2 i l l 8 i i i 1 'f v ) i 4 Page 18 1
Instrument Summary by Technical Specification Section Attachment E Proposed Current Surveillance Calculation humber Technical Existmg Technical Governing Requirement i or Qualitative int ument or Component Specification Associated Component Specification Proposed Technicet Surveillance Calibration Justific.ation Functmn Reference or Instrument humber (Setpoint) Valus Specification Valun Requirement interval Provided (DJPl 310 Core Alterations i r
! I .
Pnor to f uel ! 1 : i Refueling Interlocks - ; i j i Pnor ta fuel ,handhng (24 ; functmnal test 3 4.10.A na 'na na handimg months) QJP Pnor to fuel Core Monitoring - $RM Pnor to fuel handima (24 l functional test 3 4.10.B ,na_
,na __ _ . . _ . __ na __ __. _ _,handlms __ _ , months! _ _ _ _ QJP _ _ _ _ _
i I n c Spent F ual Pool Water level 3.4.10 C na . 'na na Daily Daily na Multiple Control Rod i I Pnor to control Pnor to control Removal 3.4.10.D na na : .e rod removal rod removal 'na 311 Reactor Fuel Assembly s - 1 L
't !
Ave. Planar Linear Heat l l 1 Generation Rate APLHGR 3 4.11. A r,a na na Daily Daily _ na Linear Heat Generation Rate (LHGR) 3 4.11.8 na na na Daily _ ,[ Daily jna Mmimum Critical Power l i Ratio l { 3_4.._11_.C_ ,na _ _ _ _ _ na_ _ _ _ _ ne __ Daily __ _ Daily _ _ _ na __ __ i t i ' Power i Flow Relationship 3 411.D na na 'na Daily Daily 'na Page 19
f 1 l i instrument Summary by Technical Specification Section Attechment E l Proposed Current Surveillance Calculetion Nundier Technical Existmg lechnical Governing Requirement l er Qualitative instrument or Component Specif; cation Associated Component Specification Proposed Technical Surveillance Calibration Justification Function Reference er instrument Number (Setpoint) Value Specification Vehm Regiarement interval Provided (0JP) 3.12 fire Protection @lternata Shutdovvn Penels)
,Once: cycle .'Once cycle i i ! i Core Spray System MOV's 3 4.12.1 na na {18_mongs)_ 1(24 months na_ __ _ _
QJP _ _ Onceleycle ;0nceleycle l l RHR System MOVs 314.1 2.2 ~na 'na na -(18 months) R4 monthst QJP
; ; .OnceIcycle 'Oncercycle i SSW Pumps
[4].3 n a_ __ _ jaa ns___ _ (18 months) (24 months) !QJ_P _ _ __ __ r l ! 0nceicycle ;0ncetcycle l RBCCW pumpa and MOVs 3.4.12.4 fna 'na na_ ^(18 monthst :(24 monthsl QJP j ;0nce/ cycle ;0nceicycle l RCIC end HPIC operabilty 3 4.12.5 na__ _ _ _ jna_ _ 'ne :(18 months) :(24 months) gGJP___ Verify ADS relief valve j , After Each RF0 & After Each RF0 &. l solenoids actuate 3 4.12.8 'na 'na na pner to startup prior to startup :QJP l
! j Each RF0 Each RF0 l l l i #
Oiesel Generator Startup 3i4.12.7 na :ne na !'(18 months) l124 months) ;0JP
?
3.13 laservice Code Testing in accordance ;In accordance l Inservice Code Testing of l with Section XI of Iwith Section XI i Pumps & Valves 3 4.13 na .no ne :ASME ofASME Ina j i 4
+
i l l Page 20
instrument Summary by Technical Specification Section Attachment E Proposed Current Surveillance Calculation kumher i Technical Exnimg Techmcal Govermng Requirement i er Quahtstive , instrument or Component Specification Associated Component Specihcation Proposed Technical Surveitlance Calibration Justification Function Reference or Instrument Number (Setpointt Va'un Specification Value Requirement Intervel Provided (DJPI Logic System Functional Testmg l I I > 1 MSL isolation Valves. MSL l i l Drsin Valves, Reactor Water . l l Sample Valves T4 2.A__ , na_ _ __ _ . ,n a . _ . ,n a _ _,0ncel18 months. Oncel24 months QJP__ __ l i I ' j RHR lsolation Valve Control, ! l - i [ Shutdovvn Cooling Valves, j j i Head Spray. Discharge to ; ! Radeste T42.A ,n a _ _ _ _ _ _ _ n_a , _ _ _ _ na__ __ _ _ ;0ncel18 months ' Oncet24 months QJ_P _ _ Reacter Water Cleanup
, l ,
isolation T42.A_ _ _ n a_ _ _ _ _ _ na _ __ _.
,na . _ ,0_nce!19 months Oncei24 months QJP Drywellisolation, TIP l j l {
Withdraml, Atmospheric ! ! 1 Contrel valves. Sump Drein l l valves T 4.2.A 'na na 'na Onca!18 month _sOncei_24 month _s]QJP _ ___ _ __ _ SBCT System, Reactor j j l l l E u Iding isolation T4 2.4 na _ _ _ . na _ _ _ na .Oncel1B months Once!?4 months QJP Core Spray Subsystem T 4.2.B _ _ _
,n a _ _ _ _ _ __ _ _ _ na_._ _ .na i0ncellB months ,0nceT24 months QJP _ _
l Low Pressure Coolant _l 1 1 i l l injection System T4.2.B na na na Once!18 months Once!24 months QJP
]
Containment Spray j j ! j Subsystem T4.2.B na na 'na Oncel18 months Oncel24 months QJP l HPCI Subsystem T 4.2.B na na na Oncef18 months Once!24 months QJP , HPCI Subsystem Auto l ) ; i isolation T4.2.B 'na na 'ns Oncel18 months Once:24 months 'QJP i ADS Subsystem , ns ' T42.B _ ,n a ___ jna ,___. _ ,0ncell8 months _ OncC4 m_onths QJP _ _ _ _ _ _ _ _ RCIC Subsystem Auto ' i t i j isolation T4.2.B . __ na ___ na _ __ . na _ _ _ _ __ Oncal18 months Once!24 months _ QJP__ _ _ _ _ _ 1 Diesel Generator initiation T4.2.B ._ na ._ _ na _ _ _ _ n a _ _ _ _ _ __ __,0ncel10 months Oncef24 months jQJP_ _ __ _ Area Cooling for safeguard ; System ( T42_.B _ _ ,na na ' na Once!18 monihs Oncel24 months QJP System Logic check T4.2.C na 'ns na Oncm18 months ,0nce!24_ months ,DJP I I Reactor Building isolation T 4.2.D _ _ _na _ j na _ _ ,na . Oncel1B months 0ncef24 months QJP i [ i 1 l SEGT System Actuation T42.D na 'ne ns Oncel18 months Once'24 months QJP
! I Recirculation Pump Trip T4.2.G __ _ n e na_ _ _ [na _ _ _ _ _ ___ Once[1_8_ months Oncel24_ months ,QJ_P , ___
Mechanica1 Vacuum Pump _! t i ! i Testing 3 4 B_H na ns na Oncel18 months Once'24 months QJP i l l Page 21
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