Proposed Tech Spec,Changing Table 3.2.B,Sections 4.5.E.1.a & 3.2 to Reflect Mods Made to Facility to Address NUREG-0737,Item II.K.3.18ML20214Q715 |
Person / Time |
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Site: |
Pilgrim |
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Issue date: |
05/20/1987 |
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From: |
BOSTON EDISON CO. |
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To: |
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Shared Package |
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ML20214Q694 |
List: |
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References |
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RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.18, TASK-TM NUDOCS 8706050127 |
Download: ML20214Q715 (6) |
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20196D0241999-06-16016 June 1999 Proposed Tech Specs Re Reactivity Control Incorporating Operating Requirements That Are Consistent with NEDO-21231, Banked Position Withdrawal Sequence ML20206R8161999-05-11011 May 1999 Proposed Tech Specs Re HPCI & RCIC Surveillance Testing ML20206H9771999-05-0505 May 1999 Proposed Tech Specs Modifying Licensing Basis for EDG on- Site Diesel Fuel Storage Requirements & Corresponding TSs ML20207F1171999-03-0303 March 1999 Proposed Tech Specs Page 3/4.6-13 Re Reactor Vessel Matl Surveillance Program Withdrawal Schedule ML20237E0851998-08-24024 August 1998 Proposed Tech Specs Pages Supporting Proposed TS Amend Related to EDG AOT ML20236X7191998-07-31031 July 1998 Proposed Tech Specs Page 3/4.9-5 Re Suppl to EDG for Allowed Outage Time ML20249C7101998-06-26026 June 1998 Proposed Tech Specs Correcting Typos & Updating Bases ML20249B3231998-06-15015 June 1998 Revised Complete Set of TS Pages Re Relocation of Radioactive Effluent TS & Radiological Environ Monitoring Program to ODCM ML20217F9341998-03-26026 March 1998 Proposed Tech Specs Re EDG Allowed Outage Time ML20217H2791998-03-25025 March 1998 Proposed Tech Specs 3.6.A.1 & 4.6.A.1,pertaining to Primary Sys Boundary,Thermal & Pressurizations Limitations & Surveillance Requirements & Basis 3/4.6.A ML20203D4981998-02-20020 February 1998 Proposed Tech Specs Pages,Incorporating Ultimate Heat Sink Temperature of 75 F Into TS 3/4.5.B & Bases,As Required by Amend 173 ML20217K5691997-10-24024 October 1997 Proposed Tech Specs Page Adding Footnote Declaring One Containment Isolation valve,30-CK-432,operate for Limited Period Despite Not Being within IST Interval for Reverse Flow Testing ML20211G2311997-09-12012 September 1997 Proposed Tech Specs Re Radiological Environ Monitoring Program Moved to ODCM ML20210K4381997-08-0808 August 1997 Revised TS Pages 3/4.5-7,B3/4.5-6 & B3/4/.5-7 Requiring Verification That SBO-DG Is Operable Prior to Voluntarily Entering LCO ML20196J0041997-07-22022 July 1997 Proposed Revised TS Bases for Section 3.5.F,correcting Volume of Water Available in Refuel Cavity & Dryer/Separator Pool When Flooded to Elevation 114 Ft & Revising Torus Water Volume to Reflect Volume Above Min Pump NPSH ML20136J5501997-03-0707 March 1997 Proposed Tech Specs Rev for Section 3.10, Core Alterations Allowing Removal of Suspect Fuel Bundles Out of Planned Sequence ML20134C9671997-01-24024 January 1997 Proposed Tech Specs 2.0 Re Safety Limits ML20133A6511996-12-23023 December 1996 Proposed Revised Tech Specs 1.0 Re definitions,3.4 Re Standby Liquid Control sys,3.5 Re Core & Containment Cooling systems,3.7 Re Containment Systems & 3.9 Re Auxiliary Electrical Sys ML20132E5491996-12-10010 December 1996 Proposed Tech Specs Table 3.2.C.1 Re Instrumentation That Initiates Rod Blocks,Table 3.2.C.2 Re Control Rod Block Instrumentation Setpoints & Table 4.2.C Re Minimum Test & Calibr Frequency for Control Rod Blocks Actuation ML20135C1461996-11-26026 November 1996 Proposed Tech Specs,Modifying Definition 1.M, Primary Containment Integrity, to Include All Instrument Line Flow Check Valves to Make Definition Consisten W/Lco 3/4.7.A.2.a.4 ML20134K4361996-11-0707 November 1996 Proposed Tech Specs Re Installation of BWROG Enhanced Option 1A ML20117K6551996-09-0505 September 1996 Proposed Tech Specs,Providing Revised TS Pages 3/4.5-7 & 3/4.5-6 ML20116M1561996-08-12012 August 1996 Proposed Tech Specs to Plant TS Section 6.0,Administrative Controls & Sections 6.5.B.10.C & 6.14 of Util Quality Assurance Manual ML20111B4201996-05-0101 May 1996 Proposed Tech Specs,Reflecting Implementation of 10CFR50 App J,Option B ML20108C1071996-05-0101 May 1996 Proposed Tech Specs,Relocating Administrative Controls Re QA Review & Audit Requirements of Section 6 from Plant TS to Boston Edison QA Manual ML20111C2761996-05-0101 May 1996 Proposed Tech Specs 3.1.1 Re Reactor Protection Sys (Scram) instrumentation,3.2.C.1 Re Instrumentation That Initiates Rod Blocks & 3/4.4 Re Standby Liquid Control ML20108C0581996-05-0101 May 1996 Proposed Tech Specs Re Core Alteration to LCO & Surveillance Conditions Associated W/Secondary Containment ML20108A6561996-04-25025 April 1996 Proposed Tech Specs Re Part of Overall Effort to Improve Outage Performance at Plant ML20095E0771995-12-0808 December 1995 Proposed Tech Specs Bases Page B2-2,including GE11 GEXL Correlation Axial Power Profile & R-factor Ranges & Correcting Error in GE11 GEXL Correlation Mass Flux Range in GE Design ref,NEDE-31152-P, GE Fuel Bundle Designs ML20092A4421995-08-31031 August 1995 Corrected Tech Spec Page Re Reactor High Pressure Trip Setting ML20086K2661995-07-14014 July 1995 Proposed Tech Specs Re Section 2.1,Bases:Safety Limits; Section 3.3.C,scram Insertion Times & Section 4.11.C, Minimum Critical Power Ratio to Reflect Use of Advanced GE-11 Fuel Design in Cycle 11 ML20078R6221995-02-15015 February 1995 Proposed Tech Specs Substituting for Pages Contained in Proposed TS Amends Submitted Prior to 950130 Authorization ML20078N4861995-02-0909 February 1995 Proposed Tech Specs,Increasing Reactor High Water Level Isolation Trip Level Setting ML20077M6921995-01-0909 January 1995 Proposed Reformatted Tech Specs & Bases ML20078K8151994-11-22022 November 1994 Proposed TS Pages 3/4.2-20 & 3/4.2-35 Re Tables 3.2.C-1 & 4.2.C ML20078K7961994-11-22022 November 1994 Proposed Tech Specs Re Changes to MSIV Leakage Requirement ML20077A8991994-11-22022 November 1994 Proposed Tech Specs Re Suppression Chamber Water Level ML20077B1861994-11-22022 November 1994 Proposed Tech Specs 3.5.F,4.5.F.1 & 3.9.B.1 & 2 Re EDG Allowed out-of-svc Time ML20072T0521994-09-0606 September 1994 Proposed Tech Specs Modification to Append a of Operating License DPR-35 Re Maintenance of Filled Discharge Pipe ML20072S0501994-09-0606 September 1994 Proposed Tech Specs Re Instrumentation That Initiates Primary Containment Isolation & Initiates or Controls Core & Containment Systems ML20072S0081994-09-0606 September 1994 Proposed Tech Specs Re Primary Containment,Oxygen Concentration & Vacuum Relief ML20072S0861994-09-0606 September 1994 Proposed Tech Specs Re Standby Gas Treatment & Control Room High Efficiency Air Filtration Sys Requirements ML20069M3311994-06-0909 June 1994 Proposed Tech Specs,Increasing Allowed out-of-service Time from 7 Days to 14 Days for Ads,Hpci & RCIC Sys,Including Section 4.5.H, Maint of Filled Discharged Pipe ML20067B7111994-02-0909 February 1994 Proposed Tech Specs Revising Wording for Page 3 of License DPR-35,clarifying Words to Aid Operators & Removing Obsolete Mechanical Snubber Acceptance Criterion BECO-93-156, Proposed Tech Specs Requesting Changes Supporting 24 Month Fuel Cycle (Submittal 3)1993-12-10010 December 1993 Proposed Tech Specs Requesting Changes Supporting 24 Month Fuel Cycle (Submittal 3) ML20059A9361993-10-19019 October 1993 Proposed Tech Specs for Removal of Scram & Group 1 Isolation Valve Closure Functions Associated W/Msl Radiation Monitors BECO-93-132, Proposed Tech Specs Removing Low Condenser Vacuum Scram in Order to Reduce Spurious Scrams,Unnecessary Plant Transients & Turbine First Stage Pressure Setpoint1993-10-19019 October 1993 Proposed Tech Specs Removing Low Condenser Vacuum Scram in Order to Reduce Spurious Scrams,Unnecessary Plant Transients & Turbine First Stage Pressure Setpoint ML20046D0441993-08-0909 August 1993 Proposed Tech Specs,Proposing 24 Month Fuel Cycle ML20044G1331993-05-20020 May 1993 Proposed Tech Specs Reducing MSIV Low Turbine Inlet Pressure Setpoint from Greater than or Equal to 880 Lb Psig to Greater than or Equal to 810 Psig & Reducing Min Pressure in Definition of Run Mode from 880 Psig to 785 Psig BECO-93-016, Proposed TS 3.5.C,D & E Re k-infinity Factor,Spent Fuel Pool Storage Capacity & Max Loads Allowed to Travel Over Fuel Assemblies,Respectively1993-02-11011 February 1993 Proposed TS 3.5.C,D & E Re k-infinity Factor,Spent Fuel Pool Storage Capacity & Max Loads Allowed to Travel Over Fuel Assemblies,Respectively 1999-06-16
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20196D0241999-06-16016 June 1999 Proposed Tech Specs Re Reactivity Control Incorporating Operating Requirements That Are Consistent with NEDO-21231, Banked Position Withdrawal Sequence ML20206R8161999-05-11011 May 1999 Proposed Tech Specs Re HPCI & RCIC Surveillance Testing ML20206H9771999-05-0505 May 1999 Proposed Tech Specs Modifying Licensing Basis for EDG on- Site Diesel Fuel Storage Requirements & Corresponding TSs ML20205A1451999-03-23023 March 1999 Core Shroud Insp Plan ML20207F1171999-03-0303 March 1999 Proposed Tech Specs Page 3/4.6-13 Re Reactor Vessel Matl Surveillance Program Withdrawal Schedule ML20151S3851998-08-31031 August 1998 Long-Term Program:Semi-Annual Rept ML20237E0851998-08-24024 August 1998 Proposed Tech Specs Pages Supporting Proposed TS Amend Related to EDG AOT ML20236X7191998-07-31031 July 1998 Proposed Tech Specs Page 3/4.9-5 Re Suppl to EDG for Allowed Outage Time ML20249C7101998-06-26026 June 1998 Proposed Tech Specs Correcting Typos & Updating Bases ML20249B3231998-06-15015 June 1998 Revised Complete Set of TS Pages Re Relocation of Radioactive Effluent TS & Radiological Environ Monitoring Program to ODCM ML20217F9341998-03-26026 March 1998 Proposed Tech Specs Re EDG Allowed Outage Time ML20217H2791998-03-25025 March 1998 Proposed Tech Specs 3.6.A.1 & 4.6.A.1,pertaining to Primary Sys Boundary,Thermal & Pressurizations Limitations & Surveillance Requirements & Basis 3/4.6.A ML20203D4981998-02-20020 February 1998 Proposed Tech Specs Pages,Incorporating Ultimate Heat Sink Temperature of 75 F Into TS 3/4.5.B & Bases,As Required by Amend 173 ML20217K5691997-10-24024 October 1997 Proposed Tech Specs Page Adding Footnote Declaring One Containment Isolation valve,30-CK-432,operate for Limited Period Despite Not Being within IST Interval for Reverse Flow Testing ML20211N6871997-09-16016 September 1997 Rev 9 to Procedure 8.I.1.1, Inservice Pump & Valve Testing Program ML20211G2381997-09-15015 September 1997 Rev 8 to PNPS-ODCM, Pilgrim Nuclear Power Station Odcm ML20211G2311997-09-12012 September 1997 Proposed Tech Specs Re Radiological Environ Monitoring Program Moved to ODCM ML20216C0631997-08-29029 August 1997 Semi-Annual Long Term Program Schedule ML20210K4381997-08-0808 August 1997 Revised TS Pages 3/4.5-7,B3/4.5-6 & B3/4/.5-7 Requiring Verification That SBO-DG Is Operable Prior to Voluntarily Entering LCO ML20196J0041997-07-22022 July 1997 Proposed Revised TS Bases for Section 3.5.F,correcting Volume of Water Available in Refuel Cavity & Dryer/Separator Pool When Flooded to Elevation 114 Ft & Revising Torus Water Volume to Reflect Volume Above Min Pump NPSH ML20210K3551997-07-0101 July 1997 Rev 16 to Procedure 7.8.1, Water Quality Limits ML20136J5501997-03-0707 March 1997 Proposed Tech Specs Rev for Section 3.10, Core Alterations Allowing Removal of Suspect Fuel Bundles Out of Planned Sequence ML20134C9671997-01-24024 January 1997 Proposed Tech Specs 2.0 Re Safety Limits ML20133A6511996-12-23023 December 1996 Proposed Revised Tech Specs 1.0 Re definitions,3.4 Re Standby Liquid Control sys,3.5 Re Core & Containment Cooling systems,3.7 Re Containment Systems & 3.9 Re Auxiliary Electrical Sys ML20132E5491996-12-10010 December 1996 Proposed Tech Specs Table 3.2.C.1 Re Instrumentation That Initiates Rod Blocks,Table 3.2.C.2 Re Control Rod Block Instrumentation Setpoints & Table 4.2.C Re Minimum Test & Calibr Frequency for Control Rod Blocks Actuation ML20135C1461996-11-26026 November 1996 Proposed Tech Specs,Modifying Definition 1.M, Primary Containment Integrity, to Include All Instrument Line Flow Check Valves to Make Definition Consisten W/Lco 3/4.7.A.2.a.4 ML20134K4361996-11-0707 November 1996 Proposed Tech Specs Re Installation of BWROG Enhanced Option 1A ML20117K6551996-09-0505 September 1996 Proposed Tech Specs,Providing Revised TS Pages 3/4.5-7 & 3/4.5-6 ML20116M1561996-08-12012 August 1996 Proposed Tech Specs to Plant TS Section 6.0,Administrative Controls & Sections 6.5.B.10.C & 6.14 of Util Quality Assurance Manual ML20117K6611996-07-17017 July 1996 Rev 15 to PNPS Procedure 1.2.2 Administrative OPS Requirements ML20108C0581996-05-0101 May 1996 Proposed Tech Specs Re Core Alteration to LCO & Surveillance Conditions Associated W/Secondary Containment ML20111B4201996-05-0101 May 1996 Proposed Tech Specs,Reflecting Implementation of 10CFR50 App J,Option B ML20111C2761996-05-0101 May 1996 Proposed Tech Specs 3.1.1 Re Reactor Protection Sys (Scram) instrumentation,3.2.C.1 Re Instrumentation That Initiates Rod Blocks & 3/4.4 Re Standby Liquid Control ML20108C1071996-05-0101 May 1996 Proposed Tech Specs,Relocating Administrative Controls Re QA Review & Audit Requirements of Section 6 from Plant TS to Boston Edison QA Manual ML20108A6561996-04-25025 April 1996 Proposed Tech Specs Re Part of Overall Effort to Improve Outage Performance at Plant ML20095E0771995-12-0808 December 1995 Proposed Tech Specs Bases Page B2-2,including GE11 GEXL Correlation Axial Power Profile & R-factor Ranges & Correcting Error in GE11 GEXL Correlation Mass Flux Range in GE Design ref,NEDE-31152-P, GE Fuel Bundle Designs ML20100J2521995-11-22022 November 1995 Rev 7 to Pilgrim Nuclear Power Station Odcm ML20092B5861995-09-0101 September 1995 Rev 0 to Third Ten-Yr Interval ISI Plan for Pilgrim Nuclear Power Station ML20092C4331995-09-0101 September 1995 Startup Test Rept for Pilgrim Nuclear Power Station Cycle 11 ML20092A4421995-08-31031 August 1995 Corrected Tech Spec Page Re Reactor High Pressure Trip Setting ML20086K2661995-07-14014 July 1995 Proposed Tech Specs Re Section 2.1,Bases:Safety Limits; Section 3.3.C,scram Insertion Times & Section 4.11.C, Minimum Critical Power Ratio to Reflect Use of Advanced GE-11 Fuel Design in Cycle 11 ML20078R6221995-02-15015 February 1995 Proposed Tech Specs Substituting for Pages Contained in Proposed TS Amends Submitted Prior to 950130 Authorization ML20078N4861995-02-0909 February 1995 Proposed Tech Specs,Increasing Reactor High Water Level Isolation Trip Level Setting ML20077Q1181995-01-13013 January 1995 Owner'S Specification for Reactor Shroud Repair ML20077M6921995-01-0909 January 1995 Proposed Reformatted Tech Specs & Bases ML20077A8991994-11-22022 November 1994 Proposed Tech Specs Re Suppression Chamber Water Level ML20077B1861994-11-22022 November 1994 Proposed Tech Specs 3.5.F,4.5.F.1 & 3.9.B.1 & 2 Re EDG Allowed out-of-svc Time ML20078K7961994-11-22022 November 1994 Proposed Tech Specs Re Changes to MSIV Leakage Requirement ML20078K8151994-11-22022 November 1994 Proposed TS Pages 3/4.2-20 & 3/4.2-35 Re Tables 3.2.C-1 & 4.2.C ML20078N8421994-11-18018 November 1994 Rev 32 to Procedure 8.7.3, Secondary Containment Leak Rate Test 1999-06-16
[Table view] |
Text
PNPS TABLE 3.2.8 (Cont'd)
INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE A!10 CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trio System (1) Trio Function Trio Level Settina Remarks 1 Core Spray Pump Start In conjunction with. loss of Timer 0<t<l sec. power initiates sequential 1 LPCI Pump Start Timer 4<t<6 sec. starting of.CSCS pumps 1 LPCI Pump Start Timer 9<t<11 sec.
1 Auto Blowdown Timer 190, 1120 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 Pressure 11 i 2 min. Permits starting CS 4
Bypass Timer and LPCI pumps and actuating ADS SRV's ir RPV
. water level is low and drywell pressure is not high.
2 RHR (LPCI) Pump Discharge 150 1 10 psig Defers ADS actuation pending Pressure Interlock confirmation of Low Pressure core cooling system operation.
J 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 Relay voltage resets Diesel Generator to an at less than 50% unloaded emergency bus.
- 2. Permits starting of CSCS 4 kV motors.
i ~ B706050127 870520 3 i DR ADOCK 05 1
49
LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT
~
3.5.0 Reactor Core Isolation Cooling 4.5.0 Reactor Core Isolation on Cooling (RCIC) Subsystem (Cont'd) LRXIf)Ahsystem (Cont'd)
- d. Flow Rate at Once/3 months 1000 psig
- e. Flow Rate at Once/ operating 150 psig cycle
- 2. From and after the date that the The RCIC pump shall deliver at RCICS is made or found to be least 400 gpm for a system head inoperable for any reason. corresponding to a reactor continued reactor power operation is permissible only pressure of 1000 to 150 psig.
during the succeeding seven days 2.
provided that during such seven When it is determined that the RCIC subsystem is inoperable, the HPCIS days the HPCIS is operable.
shall be demonstrated to be
- 3. If the requirements of 3.5.D operable immediately and weekly cannot be met, an orderly thereafter.
N shutdown shall be initiated and the reactor pressure shall be reduced to or below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.5.E Automatic Deoressurization System 4.5.E Automatic Deoressurization System (ADS)
(ADS)
- 1. The Automatic Depressurization 1. During each operating cycle the Sub. system shall be operable following tests shall be performed when'cVer there is irradiated on the ADS:
fuel in the reactor vessel and the reactor pressure is greater- a. A simulated automatic actuation than 104 psig and prior to a test shall be performed prior to startup from a Cold Condition, startup after each refueling except as specified in 3.5.E.2 outage. The ADS manual inhibit below.
switch will be included in this test.
- b. With the reactor at pressure, each relief valve shall be manually opened until a corresponding change in reactor pressure or main turbine bypass valve positions indicate that steam is flowing from the valve.
- c. Perform a test from the alternate shutdown panel to verify that the relief valve solenoids actuate. Test shall be performed after each refueling outage prior to startup.
Amendment No. 109
.)
=- .. . -- -_ .--- . . . _ , . . ._ -. -.
.S
- 3.2 . BASES (Cont'd)-
For most parameters monitored, as listed in Table 3.2.F. there are two (2) channels of instrumentation. By comparing readings between these two (2) channels, a near continuous surveillance of instrument performance is available. Meaningful deviation in comparative rea' dings of these instruments will initiate an early recalibration, thereby maintaining the quality of the instrument readings.
The Safety --Safety / Relief Valve position indication instrumentation provides the operator with information on selected plant parameters to monitor and assess these variables during and following an accident.
, . In response to NUREG-0737, modifications were made to the ADS logic to extend automatic ADS operation to a class of transients that involve slowly uncovering the core without depressurizing the vessel or pressurizing the drywell. These transients were analyzed assuming no.
high pressure injection systems (feedwater, HPCI or RCIC) are available. Only ADS is considered'available to depressurize the vessel,
, permitting operation of LPCI. The transients generally involve pipe t
breaks outside containment. Automatic ADS would not occur on low water
] level because high drywell pressure would not be present and ADS logic
) has a high drywell pressure permissive. The modification added a timer i
to the ADS logic which bypasses the high drywell pressure permissive, and a manual inhibit switch which allows the operator to inhibit automatic ADS initiation for events where automatic initiation is not desirable.
An analysis was performed to determine an upper time limit on the bypass timer. The goal was to ensure ADS is automatically initiated in time to prevent peak clad temperature (PCT) from exceeding 1500*F for a limiting
!- break, which was determir.ed to be a Reactor ' Hater Cleanup line break.
- The analysis concluded that there are 18 minutes between the low water
- level initiation of the timer and the heatup of the cladding to the limit. Since the logic includes a 2 minute delay already, the bypass timer upper limit can not be more than 16 minutes, which provides a conservative margin for PCT and allows sufficient time for operator intervention if required. A minimum time delay is incorporated to allow
! RPV water level to recover, resetting the timer and preventing depressurization. The choice of a timer setting of 11 minutes places
! the setting in the middle and provides maximum tolerance from either
(
Reference:
GE Report " Bypass Timer Calculation for the limit.
ADS /ECCS Modification for Pilgrim Station" December 16, 1986).
The. recirculation pump trip / alternate rod insertion systems are
- i. consistent with the "Monticello RPT/ARI" design described in NED0-25016 (Reference 1) as referenced by the NRC as an acceptable design
- (Reference 2) for RPT. Reference 1 provides both system descriptions
- . and performance analyses. The pump trip is provided to minimize reactor l pressure in the highly unlikely event of a plant transient coincident i~ with the failure of all control rods te scram. The rapid flow reduction
- increases core voiding providing a negative reactivity feedback. High pressure sensors and low water level sensors initiate the trip. The Amendment No. 73 i
4
~ . , - . , , ,- - , , . _ . , _ + _ , , ~ _ , . - . _ - , _ , _ _ . . - _ _ . . . . - _ _ _ , _ _ _ . . , _ . . . _ . ~ . _ _ . _ _ - _ _ _ , . .
3.2 BASES (C@nt'd) recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam iiich cethues to be generated in this unlikely postulated
-snt. Requiring the trip to ba operable only when in the RUN mode is tnerefore conservative. The low water level trip function includes a time delay of nine (9) seconds one (1) second to avoid increasing the consequences of a postulated LOCA. This delay has an insignificant effect on ATHS consequences.
Alternate rod insertion utilizes the same initiation logic and functions as RPT and provides a diverse means of initiating a reactor scram. ARI uses sensor diverse from the reactor protection system to depressurize the scram pilot air header, which in turn causes all control rods to be inserted.
References l 1. NLDO-25016, " Evaluation of Anticipated fransients Without Scram for tle Monticello Nuclear Generating Plant," September 1976.
2, NUREG-0460, Volume 3, December 1978.
Amendment No. 73a
___ _ l
Drvwell Temoerature The drywell temperature limitations of Specification 3.2.H.1 ensure that safety related equipment will not be subjected to excess temperature.
Exposure to excessive temperatures may degrade equipment and can cause loss of its operability.
The temperature elements for monitoring drywell temperature specified in Table 3.2.H were chosen on the basis of their reliability, location, and their redundancy (duel - element RTD's). These temperature elements are the primary elements used for the PCILRT.
The temperature limits specified in 3.2.H.1 are based on the BECo report entitled Drywell Temperature Report, dated January 28, 1982. The limits derived from this report take into consideration the long-term effects of ambient temperature on equipment design limits and material degradation of components required for accident mitigation or plant shutdown. The evaluation process addressed the actual assessment of potential damage and the determination of equipment status from the standpoint of both qualification integrity (for safety-related equipment) and reliability to perform it's intended function.
If the drywell temperature exceeds the limits specified in 3.2.H.1 an engineering evaluation must be initiated in order to determine whether any safety related component has been adversely affected.
The limiting drywell temperature value of 215'F (Section 3.2.H.2) was selected as to guarantee that ECCS trips occur on/or before present Technical Specification values.
The time interval of 30 minttes between successive drywell temperature instrument readings (Section 3.2.H.1) was selected so as to guarantee that ECCS trips occur on/before present Technical Specification values in the event of a drywell temperature excursion in excess of 215*F.
The instrument check interval of once per shift provides adequate assurance of equipment operability based upon engineering judgement.
Amendment No. 73b
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