Letter Sequence Other |
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MONTHYEARML20113E3641985-04-10010 April 1985 Proposed Tech Spec,Changing Power Operation Limitations Based on Number of Operating Reactor Coolant Pumps Project stage: Other ML20113E3571985-04-10010 April 1985 Application for Amends to Licenses DPR-24 & DPR-27, Consisting of Tech Spec Change Request 103,re Power Operation Limitations Based on Number of Operating Reactor Coolant Pumps.Fee Paid Project stage: Request ML20135F6601985-09-0606 September 1985 Forwards Request for Addl Info Re 850410 Application for Amends to Licenses DPR-24 & DPR-27,changing Reactor Coolant Pump Limiting Conditions for Operation.Response Requested within 45 Days of Ltr Receipt Project stage: RAI ML20134B1011985-11-0101 November 1985 Responds to 850906 Request for Addl Info Re Tech Spec Change 103 Concerning Reactor Coolant Pump Operation.Upon Loss of Both Reactor Coolant Pumps (RCP) Reactor Will Be Shut Down If One RCP Not Restarted within 1 H Project stage: Request ML20141E5481986-02-14014 February 1986 Proposed Tech Specs Allowing Single Reactor Coolant Pump Operation W/Power Above 3.5% Project stage: Other ML20141E5161986-02-14014 February 1986 Revised Application for Amends to Licenses DPR-24 & DPR-27, Changing Tech Specs to Allow Single Reactor Coolant Pump Operation W/Power Above 3.5% Project stage: Request 1985-04-10
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML20217C0521999-10-0505 October 1999 Proposed Tech Specs 15.3.1.A,15.3.3.A & 15.3.3.C,eliminating Inconsistencies & Conflict Between TSs ML20210M6171999-08-0404 August 1999 Proposed Tech Specs 15.3.5,removing Word Plasma from Discussion of Type of Display Monitor for sub-cooling Info in CR & TS 15.3.7,removing Discussion That Allowed Delay in Declaring EDG Inoperable Due to OOS Fuel Oil Transfer Sys ML20209C1831999-07-0101 July 1999 Proposed Tech Specs Page Re Amend to Licenses DPR-24 & DPR-27,removing Ifba Enrichment Curve Methodology from TS ML20205R6741999-04-12012 April 1999 Proposed Tech Specs Updating References to Reflect Relocation of Referenced Info in UFSAR ML20203A2261999-02-0202 February 1999 Proposed Tech Specs Bases Page 15.4.4,correcting References to Pbnp FSAR Re Reactor Containment Design ML20202F6521999-01-29029 January 1999 Proposed Tech Specs 15.6 & 15.7,reflecting Administrative Control Changes ML20154G5601998-10-0707 October 1998 Proposed Tech Specs Ensuring That 4 Kv Bus Undervoltage Input to Reactor Trip Is Controlled IAW Design & Licensing Basis for Facility ML20154H4231998-10-0505 October 1998 Proposed Tech Specs Section 15.4.1,removing Requirement to Check Environ Monitors on Monthly Basis ML20154C1941998-09-28028 September 1998 Proposed Tech Specs Table 15.4.1-1 Re Min Frequencies for Checks,Calibrations & Tests of Instrument Channels ML20153G7361998-09-23023 September 1998 Proposed Tech Specs Removing Explicit Testing Requirements of TS Section 15.4.13, Shock Suppressors (Snubbers) ML20236W7861998-07-30030 July 1998 Proposed Tech Specs Change 206,revising to Incorporate Changes to TS to More Clearly Define Requirements for Service Water Sys Operability ML20249C3811998-06-22022 June 1998 Proposed Corrected Tech Specs Pages Re Radiological Effluents ML20248C6651998-05-28028 May 1998 Proposed Tech Specs Providing Specific Numerical Setting for Rt,Rcp Trip & AFW Initiation on Loss of Power to 4 Kv Buses ML20248G1671998-05-27027 May 1998 Proposed Edited Tech Specs Pages Adding CR & Condenser Air Ejector Radiation Monitor SR & Correcting Errors in Notes of Table 15.4.1-1 ML20203K9571998-02-26026 February 1998 Proposed Tech Specs,Revising Sections 15.3.6, Containment Sys bases,15.3.12 CREFS Including bases,15.4.4 Containment Tests bases,15.4.11 CREFS Including Bases, 15.6.8 Plant Operating Procedures & 15.6.12 ML20203K2981997-12-15015 December 1997 Proposed Tech Specs Re Administrative Controls ML20247P2681997-09-23023 September 1997 Proposed Tech Specs Implementing Boron Concentration Changes Re Planned Conversion of Unit 2 to 18-month Fuel Cycles ML20210M5151997-08-14014 August 1997 Proposed Tech Specs,Removing Requirement in Plant TS to Perform Pbnp Unit 2 Containment Integrated Leak Rate 60-months from Previous Test ML20210J0331997-08-0707 August 1997 Proposed Tech Specs,Replacing Wording & Double Underlining of Revised Wording ML20140C9111997-06-0303 June 1997 Proposed Tech Specs,Modifying TS Section 15.3.3, Eccs,Acs, Air Recirculation Fan Coolers & Containment Spray, to Incorporate AOT Similar to Ones Contained in NUREG-1431,Rev 1 ML20141C5931997-05-0909 May 1997 Proposed Tech Specs Section 15.3.3, Eccs,Acs,Air Recirculation Fan Coolers & Containment Spray to Incorporate Allowed Outage Times Similar to Those Contained in NUREG-1431,Rev 1 ML20140D7091997-04-14014 April 1997 Proposed Tech Specs,Eliminating Provisions for Operation of Units at Below 3.5% Rated Power W/Only One RCP ML20140D7251997-04-14014 April 1997 Proposed Tech Specs,Changing Title of Corporate Officer Responsible for Nuclear Operations from Vice President- Nuclear Power, to Chief Nuclear Officer Per TS Section 15.6.2.1.c ML20137N6451997-04-0202 April 1997 Proposed Tech Specs 14.2.4, Steam Generator Tube Rupture ML20137C8331997-03-20020 March 1997 Proposed Tech Specs 15.2.2, Safety Limit,Reactor Coolant Sys Pressure ML20138M5091997-02-13013 February 1997 Proposed Tech Specs Section 15.3.3. Re Eccs,Acs,Air Recirculation Fan Coolers & Containment Spray ML20134L9791997-02-12012 February 1997 Proposed Tech Specs Re Relocation of Turbine Overspeed Protection ML20134A3661997-01-24024 January 1997 Proposed Tech Specs 15.5.4 Re Fuel Storage ML20134A3121997-01-21021 January 1997 Proposed Tech Specs 15.6.11, Radiation Protection Program ML20133N4021997-01-16016 January 1997 Proposed Tech Specs 15.3.2 Re Chemical & Volume Control Sys, 15.3.3 Re Emergency Core Cooling Sys,Auxiliary Cooling Systems,Air Recirculation Fan Coolers & Containment Spray & 15.3.8 Re Refueling ML20133N6321997-01-16016 January 1997 Proposed Tech Specs 15.1-6,15.2.2 Re Safety Limit,Reactor Coolant Sys Pressure,Ts 15.3.1-9,TS 15.3-10,TS 15.3.4-2 & TS 15.3.4-3 ML20133L1071997-01-13013 January 1997 Proposed Tech Specs 15.3.15 Re Overpressure Mitigating Sys & 15.3.1 Re Reactor Coolant Sys ML20133E5531997-01-0606 January 1997 Proposed Tech Specs 15.4.1, Operational Safety Review, Changing Ref Note 20 from TS 15.3.10.B to TS 15.3.10.E to Match TS Section Previously Containing Hot Channel Factor Limit ML20132F3471996-12-19019 December 1996 Proposed Tech Specs Improving Sections TS 15.3.10, Control Rod & Power Distribution Limits & TS 15.4.1, Operational Safety Review ML20132C3741996-12-12012 December 1996 Proposed Tech Specs Section 15.3.3 Re Eccs,Auxiliary Cooling Sys,Air Recirculation Fan Coolers & Containment Spray ML20135A8921996-12-0202 December 1996 Proposed Tech Specs 15.3.14 & 15.4.15 Re Fire Protection Sys ML20135A8431996-12-0202 December 1996 Proposed Tech Specs 15.3.10 Re Control Rod & Power Distribution Limits & 15.4.1 ML20129F2101996-09-30030 September 1996 Proposed Tech Specs Modify TS Section 15.3.3, Eccs,Auxiliary Cooling Sys,Air Recirculation Fan Coolers, & Containment Spray,To Incorporate Allowed Outage Times ML20129C0771996-09-19019 September 1996 Proposed Tech Specs Revising Section 15.3.15, Overpressure Mitigating Sys & Section 15.3.1, Rcs ML20117D2651996-08-22022 August 1996 Proposed Tech Specs Re Licensed Power Level for Plant ML20116N1341996-08-15015 August 1996 Proposed Tech Specs,Consisting of Suppl to Change Request 170,modifying TS 15.3.10, Control Rod & Power Distribution Limits & TS 15.4.1, Operational Safety Review ML20116G4891996-08-0505 August 1996 Proposed Tech Specs Modifying 15.2.3 & 15.5.3 of Change Request 188 & 15.2.1,15.2.3 & 15.3.1.G of Change Request 189 Supporting SEs ML20116D7471996-07-29029 July 1996 Proposed Tech Specs Re Health Physics Manager Qualifications for Plant ML20112F7341996-06-0404 June 1996 Proposed Tech Specs 15.2.3, Limiting Safety Sys Settings & Protective Instrumentation & 15.5.3, Design Features - Reactor ML20117K1991996-06-0404 June 1996 Proposed Tech Specs 15.2.1, Safety Limit,Reactor Core, 15.2.3, Limiting Safety Sys Settings,Protective Instrumentation & 15.3.1.G, Operational Limitations to Maintain Safety Margin for Unit 2 W/Replacement SGs ML20112E2441996-05-29029 May 1996 Proposed Tech Specs 15.1 Re definitions,15.3.6 Re Containment sys,15.4.4 Re Containment Tests & 15.6 Re Administrative Controls ML20108D1531996-04-29029 April 1996 Proposed Tech Specs Re Removal of Fire Protection Requirements ML20108A0011996-04-24024 April 1996 Proposed Tech Specs Section 15.7 RETS Re Removing Items Identified in GLs 89-01 & 95-10 as Being Procedural Details & Relocating Items to Appropriate Documents ML20149D9691996-03-20020 March 1996 Proposed Tech Specs,Providing Corrected Page to Bases for TS 15.3.1 ML20100L3891996-02-27027 February 1996 Proposed Tech Specs,Consisting of Change Request 186, Modifying TS Section 15.4.4, Containment Tests, Spec I.C.1 to State That Type a Tests Shall Be Conducted Per 10CFR50, App J Modified by Approved Exemptions 1999-08-04
[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML20217C0521999-10-0505 October 1999 Proposed Tech Specs 15.3.1.A,15.3.3.A & 15.3.3.C,eliminating Inconsistencies & Conflict Between TSs ML20210M6171999-08-0404 August 1999 Proposed Tech Specs 15.3.5,removing Word Plasma from Discussion of Type of Display Monitor for sub-cooling Info in CR & TS 15.3.7,removing Discussion That Allowed Delay in Declaring EDG Inoperable Due to OOS Fuel Oil Transfer Sys ML20210B5121999-07-15015 July 1999 Pbnp Simulator Four-Yr Rept ML20209C1831999-07-0101 July 1999 Proposed Tech Specs Page Re Amend to Licenses DPR-24 & DPR-27,removing Ifba Enrichment Curve Methodology from TS ML20205R6741999-04-12012 April 1999 Proposed Tech Specs Updating References to Reflect Relocation of Referenced Info in UFSAR ML20203A2261999-02-0202 February 1999 Proposed Tech Specs Bases Page 15.4.4,correcting References to Pbnp FSAR Re Reactor Containment Design ML20202F6521999-01-29029 January 1999 Proposed Tech Specs 15.6 & 15.7,reflecting Administrative Control Changes ML20154G5601998-10-0707 October 1998 Proposed Tech Specs Ensuring That 4 Kv Bus Undervoltage Input to Reactor Trip Is Controlled IAW Design & Licensing Basis for Facility ML20154H4231998-10-0505 October 1998 Proposed Tech Specs Section 15.4.1,removing Requirement to Check Environ Monitors on Monthly Basis ML20154C0561998-09-30030 September 1998 Rev 5 to Pbnp Units 1 & 2 IST Program Third 10-Yr Interval ML20154C1941998-09-28028 September 1998 Proposed Tech Specs Table 15.4.1-1 Re Min Frequencies for Checks,Calibrations & Tests of Instrument Channels ML20153G7361998-09-23023 September 1998 Proposed Tech Specs Removing Explicit Testing Requirements of TS Section 15.4.13, Shock Suppressors (Snubbers) ML20206E2731998-08-26026 August 1998 Rev 11 to Pbnps Odcm ML20206E2681998-08-26026 August 1998 Rev 12 to EM Environ Manual for Wepc ML20236W7861998-07-30030 July 1998 Proposed Tech Specs Change 206,revising to Incorporate Changes to TS to More Clearly Define Requirements for Service Water Sys Operability ML20249C3811998-06-22022 June 1998 Proposed Corrected Tech Specs Pages Re Radiological Effluents ML20248C6651998-05-28028 May 1998 Proposed Tech Specs Providing Specific Numerical Setting for Rt,Rcp Trip & AFW Initiation on Loss of Power to 4 Kv Buses ML20248G1671998-05-27027 May 1998 Proposed Edited Tech Specs Pages Adding CR & Condenser Air Ejector Radiation Monitor SR & Correcting Errors in Notes of Table 15.4.1-1 ML20203K9571998-02-26026 February 1998 Proposed Tech Specs,Revising Sections 15.3.6, Containment Sys bases,15.3.12 CREFS Including bases,15.4.4 Containment Tests bases,15.4.11 CREFS Including Bases, 15.6.8 Plant Operating Procedures & 15.6.12 ML20203K2981997-12-15015 December 1997 Proposed Tech Specs Re Administrative Controls ML20203K3081997-12-0808 December 1997 Rev 0 to Draft Radiological Effluent & Matls Control & Accountability Program ML20203K3231997-12-0202 December 1997 Rev 11 to Draft Odcm ML20203K3121997-12-0101 December 1997 Rev 12 to Draft Environ Manual Wisconsin Electric ML20203K3201997-11-25025 November 1997 Rev 0 to Draft Radiological Effluent Control Manual Wisconsin Electric ML20247P2681997-09-23023 September 1997 Proposed Tech Specs Implementing Boron Concentration Changes Re Planned Conversion of Unit 2 to 18-month Fuel Cycles ML20210M5151997-08-14014 August 1997 Proposed Tech Specs,Removing Requirement in Plant TS to Perform Pbnp Unit 2 Containment Integrated Leak Rate 60-months from Previous Test ML20210J0331997-08-0707 August 1997 Proposed Tech Specs,Replacing Wording & Double Underlining of Revised Wording ML20140C9111997-06-0303 June 1997 Proposed Tech Specs,Modifying TS Section 15.3.3, Eccs,Acs, Air Recirculation Fan Coolers & Containment Spray, to Incorporate AOT Similar to Ones Contained in NUREG-1431,Rev 1 ML20141C5931997-05-0909 May 1997 Proposed Tech Specs Section 15.3.3, Eccs,Acs,Air Recirculation Fan Coolers & Containment Spray to Incorporate Allowed Outage Times Similar to Those Contained in NUREG-1431,Rev 1 ML20140H0321997-05-0101 May 1997 Rev 5 to Training Programs Trpr 33.0, Licensed Operator Requalification Training Program ML20140H0171997-05-0101 May 1997 Rev 12 to Training Courses Trcr 86.0, Administrative ML20140G9811997-04-29029 April 1997 Assessment of Corrective Action Program Pbnp ML20140D7251997-04-14014 April 1997 Proposed Tech Specs,Changing Title of Corporate Officer Responsible for Nuclear Operations from Vice President- Nuclear Power, to Chief Nuclear Officer Per TS Section 15.6.2.1.c ML20140D7091997-04-14014 April 1997 Proposed Tech Specs,Eliminating Provisions for Operation of Units at Below 3.5% Rated Power W/Only One RCP ML20140G9241997-04-11011 April 1997 Rev 6 to ISTs IT 536, Containment Sump B Suction Line Leak Test (Refueling Shutdown) ML20140G9071997-04-0707 April 1997 Rev 8 to ISTs IT 525B, Leakage Reduction & Preventive Maint Program Test of 2SI-896A&B,SI Pump Suction Valves (Refueling) ML20140G8741997-04-0707 April 1997 Rev 35 to ISTs IT 04, Low Head Safety Injection Pumps & Valves (Quarterly) ML20217P2851997-04-0303 April 1997 Rev 10 to Point Beach Nuclear Plant,Units 1 & 2 Odcm ML20137N6451997-04-0202 April 1997 Proposed Tech Specs 14.2.4, Steam Generator Tube Rupture ML20140G9031997-03-21021 March 1997 Rev 8 to ISTs IT 325, CVCS Valves (Cold Shutdown) ML20140G9741997-03-21021 March 1997 Rev 0 to Operations Refueling Tests Ort 10A, Recovery from Integrated Lrt W/Core Off-Loaded ML20137C8331997-03-20020 March 1997 Proposed Tech Specs 15.2.2, Safety Limit,Reactor Coolant Sys Pressure ML20140G9621997-03-19019 March 1997 Rev 0 to Operations Refueling Tests Ort 9A, Preparation for Integrated Lrt W/Core Off-Loaded ML20140G9131997-03-17017 March 1997 Rev 4 to ISTs IT 535B, Leakage Reduction & Preventive Maint Program Test of Train B HHSI & RHR Sys (Refueling) ML20140G9521997-03-17017 March 1997 Rev 16 to Operations Refueling Tests Ort 6, Containment Spray Sequence Test ML20140G8821997-03-0404 March 1997 Rev 9 to ISTs IT 115, Instrument Air Valves (Quarterly) ML20140H3401997-02-28028 February 1997 WO Work Plan WO 9612073, Removal/Replacement of Breakers 1Y-06-11 ML20140H3011997-02-28028 February 1997 WO Work Plan WO 9612056, Removal/Replacement of Breakers 1Y-06-05 ML20140H3331997-02-28028 February 1997 WO Work Plan WO 9612072, Removal/Replacement of Breakers 1Y-06-01 ML20140H3241997-02-27027 February 1997 WO Work Plan WO 9612057, Removal/Replacement of Breakers 1Y-06-03 1999-08-04
[Table view] |
Text
- - - . - - -- -
i 3
j as actuated by either high current, low supply voltage or low electrical fre-i quency, or by a manual control switch. The significant feature of the breaker j trip is the frequency setpoint, 55.0 HZ, which assures a trip signal before the
[ pump inertia is reduced to an unacceptable value. The high pressurizer water
[
- level reactor trip protects the pressurizer safety valves against water relief. ;
! The specified setpoint allows adequate operating instrument error (2) and -
! transient overshoot in level before the reactor trips.
l .
i The low-low : steam generator water level reactor trip protects against loss of [
feedwater flow accidents. The specified setpoint assures that there will be j sufficient water inventory in the steam generators at the time of trip to allow l for starting delays for the auxiliary feedwater system.I9)
{ t 1
Numerous reactor trips are blocked at low power where they are not required [
for protection and would otherwise interfere with normal plant operations, f
- The prescribed setpoint above which these trips are unblocked assures their j availability in the power range where needed. Specifications 15.2.3.2.A(1)
] and 15.2.3.2.C have 1% tolerance to allow for a 2% deadband of the P10 '
l bistable which is used to set the limit of both items. The difference
- betweenthenominalandmaximumallowedvalue(orminimumallowedvalue)is j to account for "as measured" rack drift effects.
l
] Sustained operation with only one pump will not be permitted above 3.5 percent power. If a pump is lost while operating between 3.5 percent and 50 percent j power, an orderly and immediate reduction in power level to below 3.5 percent is i j allowed. The power-to-flow ratio will be maintained equal to or less than unity, .
J which ensures that the minimum DNB ratio increases at lower flow because the i
1 maximum enthalpy rise does not increase above the maximum enthalpy rise which !
- occurs during full power and full flow operation. ;
I References t
i 4
(1) FSAR 14.1.1 (4) FSAR 14.3.1 (7) FSAR 3.2.1 j (2) FSAR, Page 14-3 (5) FSAR 14.1.2 (8) FSAR 14.1.9 (3) FSAR 14.2.6 (6) FSAR 7.2, 7.3 (9) FSAR 14.1.11 Wh P
15.2.3-7 I
I
, a
s 15.3 LIMITING CONDITIONS FOR OPERATION l
15.3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.
i Objective To specify those limiting conditions for operation of the Reactor Coolant System l which must be met to ensure safe reactor operation.
Spacification A. OPERATIONAL COMP 0NENTS
- 1. Coolant Pumps *
- a. When the reactor is critical, except for tests, at least one reactor coolant pump shall be in operation.
(1) Reactor power shall not be maintained above 3.5% of rated
' power unless both reactor coolant pumps are in operation.
(2) If either reactor coolant pump ceases operating, immediate power reduction shall be initiated under administrative control as necessary to reduce power to less than 3.5% of rated power.
l (3) If both reactor coolant pumps cease operating and power is i
greater than 3.5% of rated power, reactor shutdcwn shall j commence imediately and the reactor trip breakers opened within one hour,
- b. When the reactor is subcritical and the average reactor coolant temperature is greater than 350*F except for tests, at least one reactor coolant pump shall be in operation. -
(1) Both reactor coolant pumps may be deenergized provided:
- a. No operations are permitted that would cause dilution of the reactor coolant system boron concentration,
- b. Core outlet temperature is maintained at least 10'F below saturation temperature, and
- c. The reactor trip breakers are open.
- c. At least one reactor coolant pump or residual heat removal system shall be in operation when a reduction is made in the ~b'6FoIn con-centration of the reactor coolant.
- 2. Steam Generator *
- a. One steam generator shall be operable whenever the average reactor
- coolant temperature is above 350*F.
'3 . Components Required for Redundant Decay Heat Removal Capability *
- a. Reactor coolant temperature less than 350*F and greater than 140*F.
(1) At least two of the decay heat removal methods listed shall be operable.
(a) Reactor Coolant Loop A, its associated steam generator and either reactor coolant pump (b) Reactor Coolant Loop B its associated steam genera *or and either reactor coolant pump applicauie only when one or more fuel assemblies are in the reactor vessel.
I 15.3.1-1
t Specification 15.3.1.A.1 requires that at least one reactor coolant pump must be operating whenever the average reactor coolant temperature is above 350 F unless the listed restrictions are established. This is required so that the FSAR zero power transients (rod withdrawal from subcritical and rod ejection) are addressed j from conservative conditions. With the reactor subcritical, with required shut-down margin, and with the trip breakers open, a single rod ejection will not result in criticality being reached. With the reactor subcritical and the average reactor coolant temperature above 350*F, a single reactor coolant pump provides sufficient decay heat removal capability. Heat transfer analysesII) show that reactor heat equivalent to 3.5% of the rated power can be removed with natural circulation only.
Items 15.3.1.A.1.a.(2) permits an orderly reduction in power if a reactor coolant pump is lost during operation between 3.5% and 50% of rated power.
Above 50% power, an automatic reactor trip will occur if either pump is lost.
The power-to-flow ratio will be maintained equal to or less than 1.0, which ensures that the mininum DNB ratio increases at lower flow since the maximum enthalpy rise does not increase above its normal full-ficw maximum value.(2)
Specification 15.3.1.A.3 provides limiting conditions for operation to ensure that redundancy in decay heat removal methods is provided. A single reactor coolant loop with its associated steam generator and a reactor coolant pump or a single residual heat removal loop provides sufficient heat removal capacity for removing the reactor core decay heat; however, single failure considerations require that at least two decay heat removal methods be avail-able. Operability of a steam generator for decay heat renoval includas two sources of water, water level indication in the steam generator, a vent path to atmosphere, and the Reactor Coolant System filled and vented so thermal convection cooling of the core is possible. If the steam generators are not available for decay heat removal, this Specification requires both residual heat removal loops to be operable unless the reactor system is in the refueling shutdown condition with the refueling cavity flooded and no operations in progress which could cause an increase in reactor decay heat load or a decrease in boron concentration, in this condition, the reactor vessel is essentially l
a fuel storage pool and removing a RHR loop from service provides conservative conditions ',hould operability problems develop in the other RHR loop. Also, one residual heat removal loop may be temporarily out of service due to 15.3.1-3b
I e
restricts leakage so that, in the event of a pipe break or isolation valve failure, makeup water for the leakage can be provided by a single coolant charging pump. If a RCGVS vent path from either the pressurizer or reactor vessel head is inoperable, Specification 15.3.1.A.7.c requires the remotely operable valves in that inoperable path to be shut with power renoved. If a vent path from the common header to the pressurizer relief tank or contain-ment atmosphere is inoperable, the isolation valve in that path must b! shut but reactor operations may continue. If both vent paths to or both vent paths from the common header are inoperable, the RCGVS is inoperable and the steps in specification 15.3.1.A.7.d must be taken.
(1) FSAR Section 14.1.11.
(2) FSAR Section 7.2.3.
15.3.1-3d