Text

Technical Specification Bases
	ML071140203
Person / Time
Site:	Indian Point
Issue date:	03/29/2007
From:	; Entergy Corp
To:	; Office of Nuclear Reactor Regulation
References
	IP-SMM-AD-103, Rev 0
	Download: ML071140203 (51)
	v • d • e

O March 29, 2007 DISTRIBUTION CONTROL LIST Document Name: IP3 ITS / BASES CC# NAME DEPARTMENT LOCATION 1

OPS PROC GP SUPERVISOR OPS PROCEDURE GROUP IP2 5 CONTROL ROOM OPERATIONS IP3 11 RES DEPARTMENT MGR RES 45-4-A 19 CONROY, PAT LICENSING / ROOM 205 GSB-2D 20 CHEMISTRY SUPERVISOR CHEMISTRY DEPARTMENT 45-4-A 21 TSC (IP3) 53' ELEVATION IP2 22 SHIFT MANAGER OPERATIONS IP3 23 LIS LICENSING & INFO SERVICE OFFSITE 25 SIMULATOR TRAINING 48-2-A 28 RESIDENT INSPETOR US NRC 88' ELEVATION IP2 32 EOF E-PLAN EOF 47 CHAPMAN, N BECHTWL OFFSITE 50 SCOZZARO, ANTHONY WESTINGHOUSE ELECTRIC OFFSITE 61 SIMULATOR TRAINING 48-2-A 69 CONROY, PAT LICENSING / ROOM 205 GSB/2D 99 KRAUS, KEVINV ST. EMERGENCY MGMT. OFFICE OFFSITE 106 SIM INSTRUCT AREA TRAINING #48 164 CONTROL ROOM OPERATIONS IP3 273 FAISON, CHARLENE NUCLEAR LICENSING WPO-12 319 GRANT, LEAH LRQ TRAINING #48 354 GRANT, LEAH LRQ TRAINING #48 357 GRANT, LEAH ILO TRAINING 48-2-A 424 GRANT, LEAH TRAINING #48 474 OUELLETTE P ENG, PLAN & MGMT INC OFFSITE 489 CLOUGHNESSY, PAT PLANT SUPPORT TEAM GSB/3B 492 WORK CONTROL OPERATION 72' ELEVATION IP2 493 OPERATIONS FIN TEAM 33 TURBIN DECK 45-1-A 494 AOEF / A. GROSSJEAN E-PLAN WPO/12-D 496 GRANT, LEAH LRQ TRAINING #48 497 GRANT, LEAH LRQ TRAINING #48:

500 GRANT, LEAH LRQ TRAINING #48 501 GRANT, LEAH LRQ TRAINING #48 512 GRANT, LEAH LRQ TRAINING #48 513 GRANT, LEAH LRQ TRAINING #48 518 DOC CONTROL DESK NRC OFFSITE 4oof

IPEC SITE QUALITY RELATED IP-SMM-AD-103 Revision 0

'Entergy MANAGEMENT ADMINISTRATIVE PROCEDURE MANUAL INFORMATIONAL USE Page. 13 of 21 ATTACHMENT 10.1 SMM CONTROLLED DOCUMENT TRANSMITTAL FORM SITE MANAGEMENT MANUAL CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES

_Page 1 of 1 7:-Enterov CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES TO: DISTRIBUTION DATE: 4/6107 PHONE NUMBER: 271-7057 FROM: IPEC DOCUMENT CONTROL The Document(s) identified below are forwarded for use. In accordance with IP-SMM-AD-103, please review to verify receipt, incorporate the document(s) into your controlled document file, properly disposition superseded, void, or inactive document(s). Sign and return the receipt acknowledgement below within fifteen (15) working days.

AFFECTED DOCUMENT: IP3 ITS/BASES/TRM DOC# REV# TITLE INSTRUCTIONS

- - - - FOLLOW THE ATTACHED INSTRUCTIONS*********************.*******

- - - - PLEASE NOTE EFFECTIVE DATE***********

RECEIPT OF THE ABOVE LISTED DOCUMENT(S) IS HEREBY ACKNOWLEDGED. I CERTIFY THAT ALL SUPERSEDED, VOID, OR INACTIVE COPIES OF.THE ABOVE LISTED DOCUMENT(S) IN MY POSSESSION HAVE BEEN REMOVED FROM USE AND ALL UPDATES HAVE BEEN PERFORMED IN ACCORDANCE WITH EFFECTIVE DATE(S) (IF APPLICABLE) AS SHOWN ON THE DOCUMENT(S).

NAME (PRINT) SIGNATURE DATE CC#

INDIAN POINT 3 TECHNICAL SPECIFICATION BASES INSTRUCTIONS FOR UPDATE: 22-04/11/07 Pages are to be inserted into your controlled copy of the IP3 Technical Specifications Bases following the instructions listed below. The TAB notation indicates which section the pages are located.

List of Effective Sections, Rev. 21 List of EffE (6 pages) (6 pages)

TAB - Table of Contents TOC, Rev. 2 TOC, Rev. 3 (4 pages) (4 pages)

TAB 3.4 - Reactor Coolant System B 3.4.4 (Rev. 0) B 3.4.4 (Rev. 1)

(4 pages) (4 pages)

B 3.4.5 (Rev. 0). B 3.4.5 (Rev. 1)

(6 pages) (6 pages)

B 3.4.6 (Rev. 1) B 3.4.6 (Rev. 2)

-(6 pages) (6 pages)

B 3.4.7 (Rev. 0) B 3.4.7 (Rev. 1)

(7 pages) (7 pages)

B 3.4.13 (Rev. 3) B 3.4.13 (Rev. 4)

_(6 pages) (7 pages)

B 3.4.17 (Rev. 0)

(8 pages)

Page 1 of 1

TECHNICAL SPECIFICATION BASES LIST OF EFFECTIVE SECTIONS BASES NUMBER EFFECTIVE BASES NUMBER EFFECTIVE SECTION REV OF PAGES DATE SECTION REV OF PAGES DATE TOC 3 4 04111/2007 B 3.5.3 1 4 08/10/2005 B 3.5.4 0 9 03/19/2001 B2.1.1 1 -_ 4 06/03/2005 _______ iB*3i6!C.'ONTAINMENTZ,2X&j B 2.1.2 1 .4 06/03/2005 B 3.6.1 0 5 03/19/2001

- B, 3.0 LCO-AND- SR APPLICABILITY ......÷

.... B 3.6.2 1 9 06/03/2005 B3.0 13 ] 18 11/07/2006 B 3.6.3 0 17 03/19/2001

~ B 3.1REACTIVITY CONTROL- - B 3.6.4 0 3 03/19/2001 B 3.1.1 1 6 06/03/2005 B 3.6.5 1 5 06/20/2003 B 3.1.2 0 7 03/19/2001 B 3.6.6 2 13 06/03/2005 B 3.1.3 1 7 10/27/2004 B 3.6.7 1 6 06/03/2005 B 3.1.4 0 13 03/19/2001 B 3.6.8 1 1 08/10/2005 B 3.1.5 0 5 03/19/2001 B 3.6.9 1 8 .06/03/2005 B 3.1.6 0 6 03/19/2001 B 3.6.10 2 12 09/16//2005 B 3.1.7 0 8 03/19/2001 ______ PANT5.SYSTEMS' 3:7,Pi B 3.1.8 0 7 03/19/2001 B 3.7.1 2 6 06/03/2005

ýB 2*.O`24,WER DISTRIBTIONIIMITS B 3.7.2 1 10 06/03/2005 B 3.2.1 0 7 03/19/2001 B 3.7.3 1 7 05/18/2001 B 3.2.2 1 7 06/03/2005 B 3.7.4 1 4 08/10/2005 B 3.2.3 0 9 03/19/2001 B 3.7.5 3 9 08/10/2005 B 3.2.4 0 7 03/19/2001 B 3.7.6 2 4 06/03/2005 R`-B33:1JNSTRUMENTATION ' B 3.7.7 1 4 12/17/2004 B 3.3.1 2 58 06/03/2005 B 3.7.8 1 7 06/03/2005 B 3.3.2 4 45 04/11/2005 B 3.7.9 2 9 06/03/2005 B 3.3.3 3 18 08/10/2005 B 3.7.10 1 3 06/03/2005 B 3.3.4 1 6 08/10/2005 B 3.7.11 5 7 08/10/2005 B 3.3.5 1 6 10/27/2004 B 3.7.12 1 4 04/11/2005 B 3.3.6 1 8 04/11/2005 B 3.7.13 3 7 06/03/2005 B 3.3.7 1 6 04/11/2005 B 3.7.14 1 3 04/11/2005 B 3.3.8 2 4 06/03/2005 B 3.7.15 0 5 03/19/2001 B34

-' tREACTOR COOLANT.-SYSTEM B 3.7.16 0 6 03/19/2001 B 3.4.1 1 6 06/03/2005 B 3.7.17 1 4 06/03/2005 B 3.4.2 0 3 03/19/2001 B 8- ELRICAL;POW'ER B 3.4.3 2 9 06/03/2005 B 3.8.1 3 30 11/04/2005 B 3.4.4 1 4 04/11/2007 B 3.8.2 0 7 03/19/2001 B 3.4.5 1 6 04/11/2007 B 3.8.3 0 13 03/19/2001 B 3.4.6 2 6 04/11/2007 B 3.8.4 1 11 01/22/2002 B 3.8.5 0 4 03/19/2001 B 3.4.7 1 7 04/11/2007 B 3.8.6 0 8 03/19/2001 B 3.4.8 0 4 03/19/2001 B 3.8.7 1 8 06/20/2003 B 3.4.9 3 5 06/03/2005 B 3.8.8 1 4 06/20/2003 B 3.4.10 0 5 03/19/2001 B 3.8.9 2 14 06/20/2003 B 3.4.11 1 7 08/10/2005 B 3.8.10 0 4 03/19/2001 B 3.4.12 2 19 08/10/2005 39,RE FUE LING1OPE RATI ON S<

3;~

B 3.4.13 4 7 04/11/2007 B 3.9.1 1 4 07/06/2006 B 3.4.14 0 10 03/19/2001 B 3.9.2 0 4 03/19/2001 B 3.4.15 3 7 08/10/2005 B 3.9.3 2 7 06/03/2005 B 3.4.16 2 6 08/10/2005 B 3.9.4 0 4 03/19/2001 B 3.4.17 0 8 04/11/2007 B 3.9.5 0 4 03/19/2001

_______--5 '9;7B ECC' C B 3.9.6 2 3 04/11/2005 B3.5.1 1 10 10/27/2004 B 3.5.2 2 13 09/16/2005 INDIAN POINT 3 Page 1 of 6 Revision 22

TECHNICAL SPECIFICATION BASES REVISION HISTORY REVISION HISTORY FOR BASES AFFECTED EFFECTIVE SECTIONS REV DATE DESCRIPTION Initial issue of Bases derived from NUREG-1431, in ALL 0 03/19/01 conjunction with Technical Specification Amendment 205 for conversion of 'Current Technical Specifications' to

'Improved Technical Specifications'.

UPDATE PACKAGE01;-031901?

________BASES T. Changes regarding containment sump flow monitor per B 3.4.13 1 03/19/01 NSE 01-3-018 LWD Rev 0.

B 3.4.15 Change issued concurrent with Rev 0.

________BASE _______UPDATE PACKAGE 02-,,01'861 Table of T Contents 1 05/18/01 Title of Section B 3.7.3 revised per Tech Spec Amend 207 B 3.7.3 1 05/18/01 Implementation of Tech Spec Amend 207 SBASES UPDATE PACKAGE 03-111901 Correction to statement regarding applicability of Function B 3.3.2 1 11/19/01 5, to be consistent with the Technical Specification.

Changes to reflect reclassification of certain SG narrow B 3.3.3 1 11/19/01 range level instruments as QA Category M per NSE 97 439, Rev 1.

Changes to reflect installation of a new control room alarm B 3.4.13 2 11/19/01 for 'VC Sump Pump Running'. Changes per NSE 01 B 3.4.15 018, Rev 1 and DCP 01-3-023 LWD.

Clarification of allowable flowrate for CRVS in 'incident B 3.7.11 1 11/19/01 mode with outside air makeup.'

_BASES UPDATE PACK GE 1;22 B 3.3.2 2 01/22/02 Clarify starting logic of 32 ABFP per EVL-01-3-078 MULTI,

_Rev 0.

B 3.8.1 1 01/22/02 Provide additional guidance for SR 3.8.1.1 and Condition Statements A.1 and B.1 per EVL-01-3-078 MULTI, Rev 0.

B 3.8.4 1 01/22/02 Revision of battery design description per plant I modification and to reflect Tech Spec Amendment 209.

B 3.8.9 1 01/22/02 Provide additional information regarding MCC in

_Table B 3.8.9-1 per EVL-01-3-078 MULTI, Rev 0.

BASES PATE PA1CKAGE.0509300 B 3.0 1 09/30/02 Changes to reflect Tech Spec Amendment 212 regarding delay period for a missed surveillance. Changes adopt TSTF 358, Rev 6.

B 3.3.1 1 09/30/02 Changes regarding description of turbine runback feature per EVAL-99-3-063 NIS.

B 3.3.3 2 09/30/02 Changes to reflect Tech Spec Amendment 211 regarding CETs and other PAM instruments.

B 3.7.9 1 09/30/02 Changes regarding SWN 1 and -2 valves per EVAL-00-3-095 SWS, Rev 0.

INDIAN POINT 3 Page 2 of 6 Revision 22

TECHNICAL SPECIFICATION BASES REVISION HISTORY AFFECTED JEFFECTIVE SECTIONS REV DATE DESCRIPTION

________ ____ BASES-,UP '-TE~AKG~610OX:

B 3.3.2 3 12/04/02 Changes to reflect Tech Spec Amendment 213 regarding B 3.6.6 1 1.4% power uprate.

B 3.7.1 1 B 3.7.6 1

_______ ___ 1§I~~ASE `JPDtEPCKAGE 7-0ý3170 B 3.3.8 1 03/17/2003 Changes to reflect Tech SpecAmendment 215 regarding B 3.7.13 1 implementation of Alternate Source Term analysis B 3.9.3 1 methodology to the Fuel Handling Accident.

BASE UPDATE B 3.4.9 1 03/28/2003 Changes to reflect Tech Spec Amendment 216 regarding 4 I_relaxation of pressurizer level limits in MODE 3.

BASE$ UPDATE PACKAGE Q97-0 B 3.4.9 2 06/20/2003 Changes to reflect commitment for a dedicated operator per Tech Spec Amendment 216.

B 3.6.5 1 06/20/2003 Implements Corrective Action 11 from CR-IP3-2002-02095; 4 FCUs should be in operation to assure representative measurement of containment air temperature.

B 3.7.11 2 06/20/2003 Correction to Background description regarding system response to Firestat detector actuation per ACT 02-62887.

B 3.7.13 2 06/20/2003 Revision to Background description of FSB air tempering units to reflect design change per DCP 95-3-142.

B 3.8.7 1 06/20/2003 Changes to reflect replacement of Inverter 34 per DCP-B 3.8.8 1 06/20/2003 01-022.

B 3.8.9 2 06/20/2003 7~ BAES UPDAT P'AC-K'AGE iO6*1,01O204, I B 3.1.3 1 10/27/2004 Clarification of the surveillance requirements for TS 3.1.3 per 50.59 screen.

B 3.3.5 1 10/27/2004 Clarify the requirements for performing a Trip Actuating Device Operational Test (TADOT) on the 480V degraded grid and undervoltage relays per 50.59 screen.

B 3.4.3 1 10/27/2004 Extension of the RCS pressure/temperature limits and corresponding OPS limits from 16.17 to 20 EFPY (TS B 3.4.12 1 Amendment 220).

B 3.5.1 1 10/27/2004 Changes to reflect Tech Spec Amendment 222 regarding extension of completion time for Accumulators.

~~ BASES8UPMATTEPACKAGEii-A1,21OO.l, B 3.7.7 1 12/17/2004 Addition of valves CT-1300 and CT-1302 to Surveillance SR 3.7.7.2 to verify that all city water header supply isolation valves are open. Reflects Tech Spec Amendment 218.

- ____ BASES -UPDATE -PAC*KA-G. N2O145$

B 3.7.11 3 01/24/2005 Temporary allowance for use of KI/SCBA for unfiltered inleakage above limit.

INDIAN POINT 3 Page 3 of 6 Revision 22

TECHNICAL SPECIFICATION BASES REVISION HISTORY AFFECTED EFFECTIVE SECTIONS REV DATE DESCRIPTION

______________Ij~jBASES :UPDALT.E PAC,,,KAGEý $13025,05§ ~ ~ Ai B 3.7.5 1 02/25/2005 Clarification on Surveillance Requirement 3.7.5.3 as it relates to plant condition/frequency of performance of Auxiliary Feedwater Pump full flow testing.

.. YBASES UPDAT.EPACAGE 430705K B 3.9.6 1 03/07/2005 Changes to reflect that the decay time prior to fuel movement is a minimum of 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months per Tech Spec Amendment 215.

<BASES UPDATEPE KAE4-GEI41 105k B 3.3.2 4- 04/11/2005 Changes to reflect AST as per Tech Spec Amendment B 3.3.6 1 224.

B. 3.3.7 1 B 3.7.11 4 NOTE: In addition to the AST changes to B. 3.7.11, the B 3.7.12 1 temporary allowance for use of KI/SCBA for unfiltered B 3.7.14 1 inleakage above limit is being removed. Tracer Gas B 3.9.6 2 testing is complete.

BEUPDATE PCKAG 1606,0305' B 2.1 .1 1 06/03/2005 Changes to reflect SPU as per Tech Spec Amendment B 2.1.2 1 225.

B 3.1.1 1 B 3.2.2 1 B 3.3.1 2 B 3.3.8 2 B 3.4.1. 1 B 3.4.3 2 B 3.4.6 1 B 3.4.9 3 B 3.4.13 3 B 3.4.16 1 B 3.5.2 1 B 3.6.2 1 B 3.6.6 2 B 3.6.7 1 B 3.6.9 1 B 3.6.10 1 B 3.7.1 2 B 3.7.2 1 B 3.7.5 2 B 3.7.6 2 B 3.7.8 1 B 3.7.9 2 B 3.7.10 1 B 3.7.13 3 B 3.7.17 1 B 3.9.3 2 INDIAN POINT 3 Page 4 of 6 Revision 22

TECHNICAL SPECIFICATION BASES REVISION HISTORY AFFECTED EFFECTIVE SECTIONS REV DATE DESCRIPTION

ý.,ý ýýýBASESfUPDA5TE 'PAC-KAJGEA1~.8I05 ýi TOC 2 08/10/2005 B 3.3.3, B 3.6.8 - Removal of Hydrogen Recombiners from the bases as per Technical Specification Amendment B 3.0 2 228. B 3.3.3 is also affected by Amendment 226.

B 3.3.3 3 B 3.7.11 - Add reference that if the primary coolant source of containment is in question, refer to ITS 5.5.2.

B 3.3.4 1 All other bases changes for this revision are associated B 3.4.11 1 with Technical Specification Amendment 226 regarding increase flexibility in Mode Restraints.

B 3.4.12 2 B 3.4.15 3 B 3.4.16 2 B 3.5.3 1 B 3.6.8 1 B 3.7.4 1 B 3.7.5 3 B 3.7.11 5 B 3.8.1 2

_______ ___ BASE'S,UPDAT E`-PACKAGE 1-09i16O5'-

B 3.5.2 2 09/16/2005 Reflect implementation of ER-04-2-029 as part of Stretch

Power Uprate (SPU) - HHSI Modification.

B 3.6.10 2 Update LCO and Condition B to clarify required actions consistent with FSAR.

....... ...... .... BASES.UPDAT PACKAGE 19; 1U040 .

B 3.8.1 3 11/04/2005 Include operability criteria for 138 kV and 13.8 kV offsite BASEScircuits.

B 3.9.1 ! 1 07/06/2006 Clarification on effective method for ensuring shutdown

_ I margin.

INDIAN POINT 3 Page 5 of 6 Revision 22

TECHNICAL SPECIFICATION BASES REVISION HISTORY BASES~ UPATTE P CKAGR-Z! 1,I0,72=06'.

B 3.0 3 11/07/2006 Reflect allowing a delay time for entering a supported system TS when the inoperability is due solely to an inoperable snubber, if risk is assessed and managed.

Limiting Condition of Operation 3.0.8 is added to provide this allowance and define the requirements and limitations

_of its use. (Amendment 229)

TOC 03 04OA1E1/ACKA -E'22_-,0 -20,72 TOC 3. 04/11/2007 Implement TS Amendment 233 related to steam generator tube integrity.

B 3.4.4 1 B 3.4.5 1 B 3.4.6 2 B 3.4.7 1 B 3.4.13 4 B 3.4.17 0 INDIAN POINT 3 Page 6 of 6 Revision 22

Facility Operating License No DPR-64 Appendix A - Technical Specification Bases TABLE OF CONTENTS B.2.0 SAFETY LIMITS (SLs)

B.2.1.1 Reactor Core SLs B.2.1.2 Reactor Coolant System Pressure SL B.3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY B.3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B.3.1 REACTIVITY CONTROL SYSTEMS.

B.3.1.1 SHUTDOWN MARGIN B.3.1.2 Core Reactivity B.3.1.3 Moderator Temperature Coefficient (MTC)

B.3.1.4 Rod Group Alignment Limits B.3.1.5 Shutdown Bank Insertion Limits B.3.1.6 Control Bank Insertion Limits B.3.1.7 Rod Position Indication B.3.1.8 PHYSICS TESTS Exceptions-MODE 2 B.3.2 POWER DISTRIBUTION LIMITS B.3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

B.3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F'AH)

B.3.2.3 AXIAL FLUX DIFFERENCE (AFD)

B.3.2.4 QUADRANT POWER TILT RATIO (QPTR)

B.3.3 INSTRUMENTATION B.3.3.1 Reactor Protection System (RPS) Instrumentation B.3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation B.3.3.3 Post Accident Monitoring (PAM) Instrumentation B.3.3.4 Remote Shutdown B.3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation B.3.3.6 Containment Purge System and Pressure Relief Line Isolation Instrumentation B.3.3.7 Control Room Ventilation (CRVS) Actuation Instrumentation B.3.3.8 Fuel Storage Building Emergency Ventilation System (FSBEVS)

Actuation Instrumentation (continued)

INDIAN POINT 3 Bi Revision 3

Facility Operating License No DPR-64 Appendix A - Technical Specification Bases TABLE OF CONTENTS (continued)

B.3.4 REACTOR COOLANT SYSTEM (RCS)

B.3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits B.3.4.2 RCS Minimum Temperature for Criticality B.3.4.3 RCS Pressure and Temperature (P/T) Limits B.3.4.4 RCS Loops-MODES 1 and 2 B.3.4.5 RCS Loops-MODE 3 B.3.4.6 RCS Loops-MODE 4 B.3.4.7 RCS Loops-MODE 5, Loops Filled B.3.4.8 RCS Loops-MODE. 5, Loops Not Filled B.3.4.9 Pressurizer B.3.4.10 Pressurizer Safety Valves B.3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

B.3.4.12 Low Temperature Overpressure Protection (LTOP)

B.3.4.13 RCS Operational LEAKAGE B.3.4-.14 RCS Pressure Isolation Valve (PIV) Leakage B.3.4.15 RCS-Leakage Detectnr--Instrumentation---W B.3.4.16 .RCS Specific Activity B.3.4.17 Steam Generator (SG) Tube Integrity B.3.5 EMERGENCY CORE COOLING SYSTEMS(ECCS)

B.3.5.1 Accumulators B.3.5.2 ECCS-Operating B.3.5.3 ECCS- Shutdown B.3.5.4 Refueling Water Storage Tank (RWST)

B.3.6 CONTAINMENT SYSTEMS B.3.6.1 Containment B.3.6.2 Containment Air Locks B.3.6.3 Containment Isolation Valves B.3.6.4 Containment Pressure B.3.6.5 Containment Air Temperature B.3.6.6 Containment Spray System and Containment Fan Cooler System B.3.6.7 Spray Additive System B.3.6.8 Not Used B.3.6.9 Isolation Valve Seal Water (IVSW) System B.3.6.10 Weld Channel and Penetration Pressurization System (WC & PPS)

(continued)

INDIAN POINT 3 B ii Revision 3

Facility Operating License No DPR-64 Appendix A - Technical Specification Bases TABLE OF CONTENTS (continued)

B.3.7 PLANT SYSTEMS B.3.7.1 Main Steam Safety Valves (MSSVs)

B.3.7.2 Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs)

B.3.7.3 Main Boiler Feedpump Discharge Valves (MBFPDVs), Main Feedwater Regulation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs) and Main Feedwater (MF) Low Flow Bypass Valves B.3.7.4 Atmospheric Dump Valves (ADVs)

B.3.7.5 Auxiliary Feedwater (AFW) System B.3.7.6 Condensate Storage Tank (CST)

.B.3.7.7 City Water (CW)

B.3.7.8 Component Cooling Water (CCW) System B.3.7.9 Service Water (SW) System B.3.7.10 Ultimate Heat Sink (UHS)

B.3.7.11 Control Room Ventilation System (CRVS)

B.3.7.12 Control Room Air Conditioning System (CRACS)

B.3.7.13 Fuel Storage Building Emergency Ventilation System (FSBEVS)

B.3.7.14 Spent Fuel Pit Water Level B.3.7.15 Spent Fuel Pit Boron Concentration B.3.7.16 Spent Fuel Assembly Storage B.3.7.17 Secondary Specific Activity B.3.8 ELECTRICAL POWER SYSTEMS B.3.8.1 AC Sources-Operating B.3.8.2 AC Sources-Shutdown B.3.8.3 Diesel Fuel Oil and Starting Air B.3.8.4 DC Sources-Operating B.3.8.5 DC Sources-Shutdown B.3.8.6 Battery Cell Parameters B.3.8.7 Inverters-Operating B.3.8.8 Inverters-Shutdown B.3.8.9 Distribution Systems-Operating B.3.8.10 Distribution Systems-Shutdown (continued)

INDIAN POINT 3 B iii Revision 3

Facility Operating License No DPR-64 Appendix A - Technical Specification Bases TABLE OF CONTENTS (continued)

B.3.9 REFUELING OPERATIONS B.3.9.1 Boron Concentration B.3.9.2 Nuclear Instrumentation B.3.9.3 Containment Penetrations B.3.9.4 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level B.3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level B.3.9.6 Refueling Cavity Water Level 0

INDIAN POINT 3 B iv Revision 3

RCS Loops-MODES 1 and 2 B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.4 RCS Loops-MODES 1 and 2 BASES BACKGROUND The primary function of the RCS is removal of the heat generated in the fuel due to the fission process, and transfer of this heat, via the steam generators (SGs), to the secondary plant.

The secondary functions of the RCS include:

a. Moderating the neutron energy level to the thermal state, to increase the probability of fission;

b. Improving the neutron economy by acting as a reflector;

c. Carrying the soluble neutron poison, boric acid;

d. Providing a second barrier against fission product release to the environment; and

e. Removing the heat generated in the fuel due to fission product decay following a unit shutdown.

The reactor coolant is circulated through four loops connected in parallel to the reactor vessel, each containing an SG, a reactor coolant pump (RCP), and appropriate flow and temperature instrumentation for both control and protection. The reactor vessel contains the fuel. The SGs provide the heat sink to the isolated secondary coolant. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and prevent fuel damage. This forced circulation of the reactor coolant ensures mixing of the coolant for proper boration and chemistry control.

Calculations have shown that reactor heat equivalent to 10% rated power can be removed via the steam generators with natural circulation without violating DNBR limits. This analysis assumed conservative flow resistances including steam-generator tube plugging and a locked rotor in each loop (Ref.1).

(continued)

INDIAN POINT 3 B 3.4.4 - 1 Revision 1

RCS Loops-MODES 1 and 2 B 3.4.4 BASES APPLICABLE SAFETY ANALYSES Safety analyses contain various assumptions for the design bases accident initial conditions including RCS pressure, RCS temperature, reactor power level, core parameters, and safety system setpoints. The important aspect for this LCO is the reactor coolant, forced flow rate, which is represented by the number of RCS loops, in service.

Both transient and steady state analyses have been performed to establish the effect of flow on the departure from nucleate boiling (DNB). The transient and accident analyses for the plant have been performed assuming four RCS loops are in operation. The majority of the plant safety analyses are based on initial conditions at high core power or zero power. The accident analyses that are most important to RCP operation are the four pump coastdown, single pump locked rotor, single pump (broken. shaft or coastdown), and rod withdrawal events (Ref. 1).

-- Steady state DNB analysis has been performed for the four RCS loop operation. For four RCS loop operation, the steady state DNB analysis, which generates the pressure and temperature Safety Limit (SL) (i.e., the departure from nucleate boiling ratio (DNBR) limit) assumes a maximum power level of 109% RTP. This is the design overpower condition for four. RCS'.loop operation. The value for the accident analysis setpoint of the nuclear overpower (high flux) trip is 108% and is based on an analysis assumption that bounds possible instrumentation errors. The. DNBR limit defines a locus of pressure and temperature points that result in a minimum DNBR greater than or equal to the critical heat flux correlation limit.

The plant is designed to operate with all RCS loops in operation to maintain DNBR above the SL, during all normal operations and anticipated transients* By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is.provided between the fuel cladding and the reactor coolant.

RCS Loops-MODES 1 and 2'satisfy Criterion 2 of 10 CFR 50.36.

(continued)

INDIAN POINT 3 B 3.4.4 - 2 Revision 1

RCS Loops-MODES 1 and 2 B 3.4.4 BASES LCO The purpose of this LCO is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat by the SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power.

An OPERABLE RCS loop consists of an OPERABLE RCP in operation providing forced flow for heat transport and an OPERABLE SG.

APPLICABILITY In MODES 1 and 2, the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are-reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

Operation in other MODES is covered by:

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation- Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits; (continued)

INDIAN POINT 3 B 3.4.4 - 3 Revision 1

RCS Loops- MODES 1 and 2 B 3.4.4 BASES ACTIONS A.1 (continued)

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SURVEILLANCE REQUIREMENTS SR 3.4.4.1 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that each RCS loop is in operation. Verification can be based on flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to DNB. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.

REFERENCES 1. FSAR, Section 14.

INDIAN POINT 3 B 3.4.4 - 4 Revision 1

RCS Loops- MODE 3 B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 RCS Loops -MODE 3 BASES BACKGROUND In MODE 3, the primary function of the reactor coolant is removal of decay heat and transfer of this heat,. via the steam generator (SG),

to the secondary plant fluid. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

The reactor coolant is circulated through four RCS loops, connected in parallel to the reactor vessel, each containing an SG, and a reactor coolant pump (RCP). Appropriate flow, pressure, and temperature instrumentation are available for control, protection, and indication. The reactor vessel contains the fuel. The SGs provide the heat sink. The RCPs circulate the water through the reactor vessel andSGs ata sufficient rate to ensure proper heat transfer and prevent fuel damage.

In MODE 3, RCPs are used to provide forced circulation for heat removal during heatup and cooldown. The MODE 3 decay heat removal requirements are low enough that a single RCS loop with one RCP running is sufficient to remove core decay heat. However, two RCS loops are required to be OPERABLE to ensure redundant capability for decay heat removal.

Calculations have shown that reactor decay heat equivalent to 10%

rated power can be removed via the steam generators with natural circulation. This analysis assumed conservative flow resistances including steam generator tube plugging and a lock rotor in each loop (Ref. 1).

APPLICABLE SAFETY ANALYSES Whenever the reactor trip breakers (RTBs) are in theclosed position and the control rod drive mechanisms (CRDMs) are energized, an inadvertent rod withdrawal from subcritical, resulting in a power excursion, is possible. Such a transient could be caused by a (continued)

INDIAN POINT 3 B 3.4.5 - 1 Revision 1

RCS Loops- MODE 3 B 3.4.5 BASES APPLICABLE SAFETY ANALYSES (continued)

Malfunction of the rod control system. In addition, the possibility of a power excursion due to the ejection of an inserted control rod is possible with the breakers closed or open. Such a transiept could be caused by the mechanical failure of a CRDM.

Therefore, in MODE 3 with RTBs in the closed position and Rod Control System capable of rod withdrawal, uncontrolled control rod withdrawal from subcritical is postulated and requires at least two RCS loops to be OPERABLE and in operation to ensure that the accident analyses limits are met. For those conditions when the Rod Control System is not capable of rod withdrawal, two RCS loops are required to be OPERABLE, but only one RCS loop is required to be in operation to be consistent with MODE 3 accident analyses.

Failure to provide decay heat removal may result in challenges to a fission product barrier. The RCS loops are part of the primary success path that functions or actuates to prevent or mitigate a Design Basis Accident or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier.

RCS Loops-MODE 3 satisfy Criterion 3 of 10 CFR 50.36.

LCO The purpose of this LCO is to require that at least two RCS loops be OPERABLE. In MODE 3 with the RTBs in the closed position and Rod Control System capable of rod withdrawal, two RCS loops must be in operation. Two RCS loops are required to be in operation in MODE.3 with RTBs closed and Rod Control System capable of rod withdrawal due to the postulation of a power excursion because of an uncontrolled rod withdrawal. The required number of RCS loops in operation ensures that the Safety Limit criteria will be met for. all of the postulated accidents.

With the RTBs in the open position, or the CRDMs de-energized, the Rod Control System is not capable of rod withdrawal; (continued)

INDIAN POINT 3 B 3.4.5 - 2 Revision 1

RCS Loops- MODE 3 B 3.4.5 BASES LCO therefore, only one RCS loop in operation is necessary to ensure (continued) removal of decay heat from the core and homogenous boron concentration throughout the RCS. An additional RCS loop is required to be OPERABLE to ensure redundant decay heat removal capability.

The Note permits all RCPs to be not be in operation for < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit performance of required tests or maintenance that can only be performed with all reactor coolant pumps not in operation. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period specified is acceptable because operating experience has shown that boron stratification is not a problem during this short period with no forced flow.

Utilization of the Note is permitted provided the following conditions are met, along with any other conditions imposed by test or maintenance procedures:

a. No operations are permitted that would dilute the RCS boron concentration, thereby maintaining the margin to criticality.

Boron reduction is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and

b. Core outlet temperature is maintained at least 10 OF below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

An OPERABLE RCS loop consists of one OPERABLE RCP and one OPERABLE SG, which has the minimum water level specified in SR 3.4.5.2. An I RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.

APPLICABILITY In MODE 3, this LCO ensures forced circulation-of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. The most stringent condition of the LCO, that is, two RCS loops OPERABLE and two RCS loops in operation, applies to MODE 3 with RTBs in the closed position. The least stringent (continued)

INDIAN POINT 3 B 3.4.5 - 3 Revision 1

RCS Loops- MODE 3 B 3.4.5 BASES APPLICABILITY condition, that is, two RCS loops OPERABLE and one RCS loop in (continued) operation, applies to MODE 3 with the RTBs open.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops- MODES 1 and 2";

LCO 3.4.6, "RCS Loops -MODE 4";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS A.1 If one required RCS loop is inoperable, redundancy for forced circulation heat removal is lost. The Required Action is restoration of the required RCS loop to OPERABLE status within the Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This time allowance is a justified period to be without the redundant, nonoperating loop because a single loop in operation has a heat transfer capability greater than that needed to remove the decay heat produced in the reactor core and because of the low probability of a failure in the remaining loop occurring during this period.

B.1 If restoration is not possible within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , the unit must be brought to MODE 4. In MODE 4, the unit may be placed on the Residual Heat Removal System. The additional Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is compatible with required operations to achieve cooldown and depressurization from the existing plant conditions in an orderly manner and without challenging plant systems.

(conti nued)

INDIAN POINT 3 B 3.4.5 - 4 Revision 1

RCS Loops- MODE 3 B 3.4.5 BASES ACTIONS C.1 and C.2 (continued)

If the required RCS loop is not in operation, and the RTBs are closed and Rod Control System is capable of rod withdrawal, the Required Action is either to restore the required RCS loop to' operation or to de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets. When the RTBs are in the closed position and Rod Control System are capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod withdrawal. This mandates having the heat transfer capacity of two RCS loops in operation. If only one loop is in operation, the RTBs must be opened. The Completion Times of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the required RCS loop to operation or de-energize all CRDMs is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period.

0.1, D.2, and D.3 If two required RCS loops are inoperable or no RCS loop is in operation, except-as during conditions permitted by the Note in the LCO section, all CRDMs must be de-energized by opening the RTBs or de-energizing.the MG sets. All operations involving a reduction of RCS boron concentration must be suspended, and action to restore one of the RCS loops.to OPERABLE status and operation must be initiated.

Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflectsthe importance of maintaining operation for forced circulation heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

(continued)

INDIAN POINT 3 B 3.4.5 _i 5 Revision 1

RCS Loops-MODE 3 B 3.4.5 BASES SURVEILLANCE REQUIREMENTS SR 3.4.5.1 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loops are in operation. Verification can be based on flow rate, temperature, or pump status monitoring, which ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS loop performance.

SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the.actual secondary side water level is > 71% wide range for each required loop. Depending on plant conditions, either wide range or narrow range SG level instruments may be used to verify this SR is met. Operators may be required to adjust the indicated level -to compensate for the effects of SG temperature. If the SG secondary side actual water level is < 71%

wide range, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink for removal of the decay heat. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency. is considered adequate in view of other indications available in the control room to alert the operator to a loss of SG level.

SR 3.4.5.3 Verification that the required RCPs are OPERABLE ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required RCPs.

REFERENCES 1. FSAR 14.1.6.

INDIAN POINT 3 B 3.4.5 - 6 Revision 1

RCS Loops- MODE 4 B 3.4.6 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Loops-MODE 4 BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

The reactor coolant is circulated through four RCS loops connected.

in parallel to the reactor vessel, each loop containing a SG and a reactor coolant pump (RCP). Appropriate flow, pressure, and temperature instrumentation are available for control, protection, and indication. .The RCPs and RHR pumps circulate the coolant

-through--the reactor-vessel-and SGs-at -a-sufficient rate to ensure proper heat transfer and to prevent boric acid stratification.

Each RHR loop consists of-one .RHR pump and one RHR heat exchanger as well as associated piping and valves to transfer heat between the RHR heat exchanger and the core. Although either RHR heat exchanger may be credited for either RHR loop, one RHR heat exchanger must be OPERABLE for each OPERABLE RHR loop.

In MODE 4, either RCPs or RHR loops can be used to provide forced circulation. The intent of this LCO is to provide forced flow from at least one RCP or one RHR loop for decay heat removal and transport. The flow provided by one RCP loop or RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that two paths be available to provide redundancy for decay heat removal.

When the boron concentration of the RCS is reduced, the process should be uniform.to prevent sudden reactivity changes. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one (continued)

INDIAN POINT 3 B 3.4.6-1 Revision 2

RCS Loops- MODE 4 B 3.4.6 BASES BACKGROUND residual heat removal pump is running while boron concentration (continued) is being changed. The residual heat removal pump will circulate the primary system volume in approximately one half hour. Boron concentration in the pressurizer is not a concern because of the low pressurizer volumeand because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant.

Calculations have shown that reactor decay heat equivalent to 10%.

rated power can be removed via the steam generators with natural circulation. This analysis-assumed conservative flow resistances including steam generator tube plugging and a lock rotor in each loop.(Ref. 1).

APPLICABLE SAFETY ANALYSES In MODE 4, RCS circulation is considered in the determination of the time available for mitigation of the accidental boron dilution event. The RCS and RHR loops provide this circulation.

The RHR System in conjunction, with, the CCW and SWS Systems function to cool the unit from RHR entry condition (T<350 0 F) to Mode 5 (T<200F), during normal and post accident operations. The time required to cool from 350°F to, 200OF is a function of the number of CCW, SWS and RHR. trains operating. As presented in UFSAR, Section 9, two trains of pumps and heat exchangers are usually used to remove residual and sensible heat during normal plant cool-down. If one train of pumps and/or heat exchangers is not operable, safe operation is governed by Technical Specifications and safe shutdown of the plant is not affected; however, the time for cool-down is extended.

RCS Loops-MODE 4 satisfy Criterion 4 of 10 CFR 50.36.

LCO The purpose of this LCO is to require that at least two loops be OPERABLE in MODE 4 and that one of these loops be in operation. The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced circulation. An additional loop is required to be OPERABLE to provide redundancy for heat removal.

(continued)

INDIAN POINT 3 B 3.4.6- 2 Revision 2

RCS Loops- MODE 4 B 3.4.6 BASES LCO Note 1 permits all RCPs and RHR pumps to not be in (continued) operation for # 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit performance of required tests or maintenance thai can only be performed with no forced circulation. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is acceptable because operating experience has shown that boron stratification is not a problem during this short period with no forced flow.

Utilization of Note 1is permitted provided the following conditions

.are met along with any other conditions imposed by test or maintenance procedures:

a. No operations are permitted that would dilute the RCS boron concentration, therefore maintaining the margin to criticality. Boron reduction is prohibited because a uniform concentration distribution throughout the RCS cannot be

ensured-when in natural circulation; and

b. Core outlet temperature is maintained at least 10OF below saturation temperature, so that no vapor bubble may form and possibly-cause a natural circulation flow obstruction.

Note 2 requires that the reactor coolant pump starting requirements of LCO 3.4.12, Low Temperature Overpressure Protection (LTOP), must be met before the start of an RCP with any RCS cold leg temperature less than or equal to the LTOP arming temperature. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG, which has the minimum water level specified in SR 3.4.6.2.

Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required.

(continued)

INDIAN POINT 3 B 3.4.6- 3 Revision 2

RCS Loops- MODE 4 B 3.4.6 BASES APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of either RCS or RHR provides sufficient circulation for these purposes. However, two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to meet single failure considerations.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops- MODE 3";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation- High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation- Low Water Level" (MODE 6).

ACTIONS A.1 If one required RCS loop is inoperable and two RHR loops are inoperable, redundancy for heat removal is lost. Action must be initiated to restore a second RCS or RHR loop to OPERABLE status.

0 The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 If one required RHR loop is OPERABLE and in operation and there are no RCS loops OPERABLE, an inoperable RCS or RHR loop must be restored to OPERABLE status to provide a redundant means for decay heat removal.

If the parameters that are outside the limits cannot be restored, the unit must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to MODE 5 is a conservative action with regard to decay heat removal. With only one RHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the only OPERABLE RHR loop, it would be safer to initiate that loss from MODE 5 (< 200 0F) rather than MODE 4 (200 to 350 0F). The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an orderly manner and without challenging plant systems.

.(continued)

INDIAN POINT 3 B 3.4.6- 4 Revision 2

RCS Loops-MODE 4 B 3.4.6 BASES ACTIONS C.1 and C.2 (continued)

If no loop is OPERABLE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RCS or RHR loop to OPERABLE status and in operation must be initiated. Boron dilution requires forced circulation for proper mixing, and the margin to criticality must not be reduced in this type of operation. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that one RCS or RHR loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RCS and RHR loop performance.

SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the actual secondary side water level is > 71% wide range for each required loop. Depending on plant conditions, either wide range or narrow range SG level instruments may be used to verify this SR is met. Operators may be required to adjust the indicated level to compensate for the effects of SG temperature. If the SG secondary side actual water level is < 71%

wide range, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.

(continued)

INDIAN POINT 3 B 3.4.6- 5 Revision 2

RCS Loops- MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS SR 3.4.6.2 (continued)

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level.

SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pump and associated support systems.

The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

REFERENCES 1. FSAR Chapter 14.1.6.

INDIAN POINT 3 B 3.4.6 - 6 Revision 2

RCS Loops -MODE 5, Loops Filled B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 RCS Loops-MODE 5, Loops Filled BASES BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and transfer this heat either to the steam generator (SG) secondary side coolant, via natural circulation (Ref. 1), or the component cooling water via the residual heat removal (RHR) heat exchangers. While the principal means for decay heat removal is via the RHR System, the SGs, via natural circulation (Ref. 1), are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this MODE, they are capable'of being a heat sink due to their large contained volume of secondary'water. As-long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RHR loops connected to the RCS, each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication.

One RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification. The boron concentration in the pressurizer is of no concern because of the low pressurizer volume and because the pressurizer boron concentration will be higher than the rest of the reactor coolant.

Each RHR loop consists of one RHR pump and one RHR heat exchanger as well as associated piping and valves to transfer heat between the RHR heat exchanger and the core. Although either RHR heat exchanger may be credited for either RHR loop, one RHR heat exchanger must be OPERABLE for each OPERABLE RHR loop.

The number of loops in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at (continued)

INDIAN POINT 3 B 3.4.7 - 1 Revision 1

RCS Loops-MODE 5, Loops Filled B 3.4.7 BASES BACKGROUND least one RHR loop for decay heat removal and transport. The (continued) flow provided by one RHR loop is adequate for decay heat removal.

The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal.

The LCO provides for redundant paths of decay heat removal capability. The first path can be an RHR loop that must be OPERABLE and in operation. The second path can be another OPERABLE RHR loop or maintaining two SGs with secondary side water levels > 71% wide range to provide an alternate method for decay heat removal via natural circulation (Ref. 1).

When using SGs depending on natural circulation as the backup decay heat removal system in Mode 5, consideration should be given to the potential need for the following: (1) the ability to pressurize and control pressure in the RCS, (2) secondary side water level in the SG relied upon for decay heat removal, (3) availability of a supply of feedwater, and (4) availability of an auxiliary feedwater pump capable of injecting into the relied-upon SGs (Ref.1).

During natural circulation, the SGs secondary side water may boil creating the need to release steam through the atmospheric relief valves or other openings that may exist during shutdown conditions.

Therefore, consideration should be given to avoiding the potential for pressurization of the SGs secondary side. It is also important to note that during decay heat removal using natural circulation, a MODE change (MODE 5 to MODE 4) could occur due to heat up of the RCS (Ref.1).

APPLICABLE SAFETY ANALYSES In MODE 5, RCS circulation is considered in the determination of the time available for mitigation of the accidental boron dilution event. The RHR loops provide this circulation.

RCS Loops-MODE 5 (Loops Filled) satisfy Criterion 4 of 10 CFR 50.36.

(continued)

INDIAN POINT 3 B 3.4.7 -2 Revision I

RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES LCO The purpose of.this LCO is to require that at least one of the RHR loops be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs with secondary side water level > 71% wide range. One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant under the~e conditions. An additional RHR loop is required to be OPERABLE to meet single failure considerations. However, if the standby RHR loop is not OPERABLE, an acceptable alternate method is two SGs with secondary side water level > 71% wide range. Should the operating RHR loop fail, the SGs could be used to remove the decay heat via natural circulation.

Note 1 permits all RHR pumps to not be in operation # 1.hour per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of-the Note is to permit testing and maintenance that can be performed only when in MODE 5 with no forced circulation. This allowance is acceptable because operating experience has shown that boron stratification is not likely during this short period with no forced flow.

Utilization of Note 1 is permitted provided the following conditions are met, along with any other conditions imposed by maintenance or test procedures:

a. No operations are permitted that would dilute the RCS boron concentration, therefore maintaining the margin to criticality.. Boron reduction is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and

b. Core outlet temperature is maintained at least 10 OF below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during MODE 5 with no forced circulation.

(continued)

INDIAN POINT 3 B 3. 4.7 - 3 Revision 1

RCS Loops-MODE 5, Loops Filled B 3.4.7 BASES LCO Note 3 requires that the reactor coolant pump starting (continued) requirements of LCO 3.4.12, Low Temperature Overpressure Protection (LTOP), must be met before the start of a reactor coolant pump (RCP) with an RCS cold leg temperature less than the LTOP arming temperature specified in LCO 3.4.12, Low Temperature Overpressure Protection (LTOP). This restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

Note 4 provides for an orderly transition from MODE 5 to MODE 4

.during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. A SG can perform as a heat sink with forced flow or natural circulation when it has an adequate water level and is OPERABLE.

APPLICABILITY In MODE 5 with RCS loops filled,this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least two SGs is required to be

> 71% wide range.

Loops filled is based on the ability to use the SGs as a backup.

means of decay heat removal. The RCS loops are considered filled provided that pressurizer level has been maintained >10%. The loops are also considered filled following the completion of filling and venting the RCS. The ability to pressurize the RCS to >100 psig and to control pressure must be established to take credit for use of the SGs as backup decay heat removal. This is to prevent flashing and void formation at the top of the SG tubes (continued)

INDIAN POINT 3 B 3.4.,7 - 4 Revision 1

RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES APPLICABILITY which may degrade or interrupt the natural circulation flow path (continued) (Ref. 1).

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops- MODE 4";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6)..

ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side-water level <-71% wide range-redundancy for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects.-the importance of maintaining the availability of two paths for heat-removal.

B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE status and in operation must be initiated. To prevent boron dilution, forced circulation is required.to provide proper mixing and preserve the margin to criticality in this type of operation. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

(continued)

INDIAN POINT 3 B 3.4.7 - 5 Revision 1

RCS Loops-MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS SR 3.4.7.1 This SR requires verification every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that the required loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient considering other indications and alarms available to the operator in the control room to monitor RHR loop performance.

SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring the secondary side water level > 71% wide range ensures an alternate decay heat removal method, via natural circulation, in the event that the second RHR loop is not OPERABLE. Depending on plant conditions, either wide range or narrow range SG level instruments may be used to verify this SR is met. Operators may be required to adjust the indicated level to compensate for the effects of SG temperature.

If both RHR loops are OPERABLE, this Surveillance is not needed.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications available in the control room to alert the operator to the loss of SG level.

SR 3.4.7.3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the RHR pump. If secondary side water level is > 71% wide range in at least two SGs, this Surveillance is not needed. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience.

(continued)

INDIAN POINT 3 B 3.4.7 - 6 Revision 1

RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation."

INDIAN POINT 3 B 3.4.7 - 7 Revisionl1

RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of

-- LEAKAGE.- 10 CFR 50, Appendi-x-A,-GDC-30 (Ref. 1),-requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure.

boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

(continued)

INDIAN POINT 3 B.3.4.13 - 1 Revision 4

RCS Operational LEAKAGE B 3.4.13:

BASES APPLICABLE SAFETY ANALYSES Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for events resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is one gallon per minute or increases to one gallon per minute as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to.less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting.from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The FSAR (Ref. 2) analysis for. SGTR. assumes the contaminated secondary fluid is released via safety valves and atmospheric dump valves. The 1 gpm primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes 0.3 gpm primary to secondary LEAKAGE is through the affected SG and 1 gpm through all SGs as an initial condition. The dose consequences resulting from the SLB accident are well within.the limits defined in 10 CFR 50.67 and the staff approved licensing basis (i.e., a small fraction of these limits).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36.

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, (continued)

INDIAN POINT 3 B.3.4.13 - 2 Revision 4

RCS Operational LEAKAGE B 3.4.13 BASES LCO resulting in higher LEAKAGE. Violation of this LCO could (continued) result in continued degradation of the RCPB LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount and is consistent with the capability of the equipment required by LCO 3.4.15, RCS Leakage Detection Instrumentation. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE Up to 10 gpm-of-identified LEAKAGE. is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified.LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE, the leakage into closed systems or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered.LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE Through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 3). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day."

The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

(continued)

INDIAN POINT 3 B.3.4.13 - 3 Revision 4

RCS Operational LEAKAGE B 3.4.13 BASES APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest.

when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

Leakage past PIVs or other leakage into closed systems is that leakage that can be accounted for and contained by a system not directly connected to the atmosphere. Leakage past PIVs or other leakage into closed systems is not included in the limits for either identified or unidentified LEAKAGE but PIV leakage must be within the limits specified for PIVs in LCO 3.4.14, "RCS Pressure Isolation Valves (PIV)." Leakage past PIVs or other leakage into closed systems is quantified before being exempted from the limits for identified LEAKAGE.

ACTIONS A.1 Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time tozverify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.

B.1 and B.2 If any pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limit, or if unidentified or identified LEAKAGE, cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure.

boundary LEAKAGE. The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

(continued)

INDIAN POINT 3 B.3.4.13 - 4 Revision 4

RCS Operational LEAKAGE B 3.4.13 BASES ACTIONS B.1 and B.2 (continued)

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full bower conditions in an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.

SURVEILLANCE REQUIREMENTS SR 3.4.13.1 Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of' the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be met with the reactor at steady state operating conditions and near operating pressure. The surveillance is modified by two Notes. Note 1 states that this SR I is not required to be performed in MODES 3 and 4 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and a Note requires the Surveillance to be met when steady state is established. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

(continued)

INDIAN POINT 3 B.3.4.13 - 5 Revision 4

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS SR 3.4.13.1 (continued)

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the systems that monitor the containment atmosphere radioactivity and the operation of the containment sump pump. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation." It should be noted that LEAKAGE past seals and gaskets, measured leakage past PIVs, and other leakage into closed systems is not pressure boundary LEAKAGE.

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. A Note under the Frequency column states that this SR is required to be performed during steady state operation.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met.

If this SR is not met, compliance with LCO 3.4.17 "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 4. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after.

establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

(continued)

INDIAN POINT 3 B.3.4.13 - 6 Revision 4

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS SR 3.4.13.2 (continued)

The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 4).

REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.

2. FSAR, Section 14.

3. NEI 97-06, "Steam Generator Program Guidelines."

4. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

INDIAN POINT 3 B.3.4.13 - 7 Revision 4

SG Tube Integrity B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.17 Steam.Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG. The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops - MODES 1 and 2,"

-[C0-3.4i5, "RCS L66ps---MODE 3,"-ECO-3-.C46,-"RCS Loops -- MODE 4,"

and LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject.to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear.

These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 5.5.8, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.5.8, tube integrity is maintained when the SG performance criteria are met.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.5.8.

INDIAN POINT 3 B 3.4.17 - 1 Revision 0

SG Tube Integrity B 3.4.17 BASES BACKGROUND (continued)

Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

APPLICABLE SAFETY The steam generator tube rupture (SGTR) accident is thelimiting ANALYSES design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released to the atmosphere via SG safety valves or atmospheric relief valves.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the applicable limits of 10 CFR 50.67 (Ref. 2) and Regulatory Guide 1.183 (Ref. 3).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

INDIAN POINT 3 B 3.4. 17 - 2 Revision 0

SG Tube Integrity B 3.4.17 BASES LCO (continued)

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld as the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.8, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to met the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated'transients included in the design specification.

Tube burst is defined as,-"The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of thedegradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero."

The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradations, axial thermal loads are INDIAN POINT 3 B 3.4. 17 - 3 Revision 0

SG Tube Integrity B 3.4.17 BASES LCO (continued) 0 classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5)..

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed'O.3 gpm per SG and I gpm through.all SGs.

The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

INDIAN POINT 3 B 3.4. 17 - 4 Revision 0

SG Tube Integrity B 3.4.17 BASES APPLICABILITY (continued)

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required-Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.I and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling out age or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

INDIAN POINT 3 B 3.4.17 - 5 Revision 0

SG Tube Integrity B 3.4.17 BASES ACTIONS (continued)

A.1 and A.2 (continued)

If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

INDIAN POINT 3 B 3.4. 17 - 6 Revision 0

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.17.1 (continued)

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws

_.satisfying the tube repair criteria. Inspection scope (i.e.,

which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations.

The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6)i The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.8 contains prescriptive requirements concerning inspection intervals-to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 5.5.8 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

INDIAN POINT 3 B 3.4:.17 - 7 Revision 0

SG Tube Integrity B 3.4.17 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.17.2 (continued)

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50. 67.

3. Regulatory Guide 1.183, "Alternate Radiological Source Terms for Evaluating Design Basis Accidents in Nuclear Power Reactors", July 2000.

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6. EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

INDIAN POINT 3 B 3.4.17 - 8 Revision 0

ML071140203

Text

Navigation menu

Search