Proposed Tech Spec Change 82-3,consisting of Sections 3.6.D,safety & Relief Valves & 4.6.D,basisML20041C210 |
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Pilgrim |
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Issue date: |
02/19/1982 |
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BOSTON EDISON CO. |
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Shared Package |
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ML20041C209 |
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NUDOCS 8202260343 |
Download: ML20041C210 (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
.
Proposed Technical Specification Proposed Chanae Reference is made to Pilgrim Nuclear Power Station, Unit #1 Technical Specifications, Section G.6.0, " Safety and Relief Valves" and Section 4.6.D contained on Page 126, the " Basis" for these sections contained on Page 145, and Table 3.2.F on pages 58 a (new) and 59.
Currently, Section 3.6.D consists of two paragraphs, denoted 1 and 3.
Boston Edison proposes the addition of two new paragraphs, denoted 3 and 4. These paragraphs shall state:
- 3. If the temperature of any safety relief discharge pipe exceeds 212 F during normal reactor power operation, a Report shall be issued per T.S. Section 6.9.B.1 which shall address the actions that have been taken or a schedule of actions to be taken.
- 4. The limiting conditions of operation for the instrunientation that monitors tail pipe temperature are given in Table 3.2.F.
The paragraph currently designated 3.6.D.3 will be re-designated 3.6.D.2.
Section 4.6.D currently consists of two paragraphs, denoted 1 and 2.
Boston Edison proposes the addition of two new paragraphs denoted 3 and 4. These paragraphs shall state:
- 3. Whenever the safety relief valves are required to be operable, the discharge pipe temperature of each safety relief valve shall be logged daily.
- 4. Instrumentation shall be talibrated and checked as indicated in Table 4.2.F.
We proposed to add a paragraph to the " Basis" (pg.145) in explanation of these additions, it shall state:
Pilgrim's experience with 2 stage safety / relief valves has demonstrated that minimum leakage exists when the tailpipe temperature is 215 Farenheit. Therefore, a reporting requirement triggered by a .
temperature of 212 F is conservative, and assures timely reporting before leakage reaches significant proportions.
8202260343 820219 PDR ADOCK 05000293 P PDR o : . . _ _ . _. _ _ _ _ __
t The Table 3.2.F, Note (5) will have an asterisk
word back-up to reference it to the addition of Note (6). -
Note (6) shall state:
(6) At a minimum, the above listed SRV tail pipe temperature thermo-couples will be operable for each valve when the valves are required to be operable. If a thermocouple becomes inoperable, it shall be returned to an operable condition within 31 days or the reactor shall be placed in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Add page 58a to Table 3.2.F (Continued) to incorporate Note (6).
Reason for Change We propose these changes at the behest of USNRC, as stated in a meeting conducted between Boston Edison and NRC on December 18, 1981 and restated on page 3 of a meeting summary issued January 5,1982.
Safety Considerations These proposals are made at the request of NRC. They do not present unreviewed safety questions. They have been reviewed and approved by the Operations Review Committee, and reviewed by the Nuclear Safety Review and Audit Committee.
Schedule of Change These changes will be put into effect upon our receipt of the Commission's approval.
Fee Determination i
Pursuant to 10 CFR Section 170.12, Boston Edison Company proposes this change as a Class III.
l l
l l
I
l TIMITE G CONDITION FOR OPERATION SURVEII.I.ANCE REQUIREMENT 31 3.6.C Coolant chemistry (Cont'd) 4.6 l
l power operation is per=issible .
only during the succeeding seven days.
- 3. If the conditions in'1.or 2 above cannot be met, an orderly shutdown
- shall be initiated and the reactor
+
shall be in a Cold Shutdown Condi- ..,
, tien within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. Safety and Relief Valves D. Safety and Relief Valves L. During reactor power operating 1. At least one safety valve and two conditions and prior to reactor relief / safety valves shall be startup from a Cold Condition, or checked or replaced with bench -
whenever reactor coolant pressure checked valves once per operating is greater than 104 psig and tem- cycle. All valves will be tested perature greater than 340 F, both every two cycles.
safety valves and the safety modes of all relief valves shall be op- The set point of the safety valves erable.
shall be as specified in Specifi-2.
cation 2.2.
If Specification 3.6.D.1 is not met, an orderly shutdown shall be in- 2. At least one of the relief / safety itiated and the reactor coolant valves shall be disassenbled and pressure shall be below 104 psig inspected each refueling outage, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- 3. 3. Whenever the safety relief valves If the temperature of any safety are required to be operable, the relief discharge pipe exceeds 212 0F during normal reactor power operation, discharge pipe temperature of each safety relief valve shall be a Report shall be issued per T.S. logged daily. ~
Section 6.9.B.1 which shall address -
the actions that have been taken or a 4. Instrumentation shall be calibrated schedule of actions to be taken, and checked as indicated in Table 4.2.F.
- 4. The limiting conditions of operation for the instrumentation that monitors tail pipe temperature are given in Table 3.2.F.
il 4 .,
126
.p -
c!
i; -
Surveillance Instrumentation Table 3.2.F. (Continued) 9 Minimum # of i
Operable Instrument Instrument Type Indication Channels # Instrument and Range Notes 1/ Valve See Note (6) Tail Pipe Temperature Thermocouple (6)
Indication i
4 4
1 4
1 1
l 58a
l o
' ~
ll I
l NOTES FOR TABLE 3.2.F_
g .
(1)
From and after the date that one of thest parameters is reduced t.
(' to one indication, continued operation is permissible during the i
succeeding thirty days unless such instrumentation is sooner made .
4 operable. '
- Trom and after the date that one of these parameters is not indi-(2) cated in the control room, continued operation is permissible
- h instrumentation is
, during the succeeding seven days unless suc sooner made operable.
(3) If the requirements of notes (1) and (2) cannot be set, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(4) These surveillance instruments are considered to be redundant to each other. ,
At a minimum, the primary or back-up* parameters shall Withbe operable pricaryfer
. (5) both each and valve when the valves are required to be operable. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. backup to operable status within 31 days or be in a shutdown mode within The following instruments sre associated with the safety / relief & safety valves:
s
- Secondary Primary Valve Tail Pipe Temperature Thernocouple Acoustic Monitor ,
TE6271-B 203-3A ZT-203-3A TE6272-B 203-3B ZT-20 3-3B i
\t TE6273-B 203-3C ZT-203-3C TE6276-B 203-3D ZT-203-3D TE6274-B 203-4A ZT-203-4A TE6275-B 203-4B ZT-20 3-4B
/
At a minimum, the above listed SRV tail pipe temperature, one of the
. (6) dual thermocouples, will be operable for each valve when the valves If a thermocouple becomes inoperable, it are required to be operable.
shall be returned to an operable condition within 31 days or the reactor snall be placed in a shutdown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I' A=endment No. 27 59
'g w --
7 - , . -
g .w. , *
/: ,,. ,
..- y 3.6.D and b.6.D ,
h Safety and Relief Valves As discussed in Subsection 4.4.6 of the Final Safety Analysis, Report, design of the nuclear system pressure relief system is intended to protect the nuclear system from,overpressurization in the event of the safety valve sia.ing transient.JJLn in, direct scram is assumed because.ASME Boiler and' Pressure Vessel Code,Section III, requires that protection systems directly related to the valve sizing transient must not be credited with action in determin4ng valve relieving capacity. Atotalof4 relief /safetyvalvesand2safetyvalvesis provided by the design. -
/ ,
Experience in safety valve operation shows that a testing of at least 50% of the safety valves per refueling outage is adequate to detect fad ures or deterioration. The tolerance value of +1% is in accor-
. . . . ith Section III of the ASME Boiler and Pressure Vessel Code.
An anQ.is ca.., been performed which shows that with all safety valves set 1% higher, the reactor coolant pressure safety limit of 1375 psig is not exceeded.
l Therelief/safetyvalveshavetwofunctions;i.e.,powerreliefor
~
self-actuated by high pressure. Power relief is a solenoid actuaced
- f.
\.
function (Autooatic Pressure Relief) in which external instrumen-tation signals of coinci. lent hi h 6drywell pressure and low-los water
- level initiate the valves to open. This function is discussed in Specification 3 5.D.. In addition,. the valves can be operated manually . . , ,
Pilgrim's ' experience with 2 stage safety / relief valves has demonstrated that minimum leakage exists when the tailpipe temperature is 215 Farenheit.
Therefore, a reporting requirement triggered by a temperature of 212 " is conservative, and assures timely reporting before leakage reaches sip'_f ~ ~
eant proportions.
1
!l 145 C, .
e
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