Text

Technical Specification Bases Change
	ML072910292
Person / Time
Site:	Indian Point
Issue date:	10/01/2007
From:	; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	IP-SMM-AD-103, Rev 0
	Download: ML072910292 (73)
	v • d • e

September 27, 2007 DISTRIBUTION CONTROL LIST Document Name: 1P3 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 WALEPOLE, ROBERT LICENSING / ROOM 205 GSB-2D 20 CHEMISTRY SUPERVISOR CHEMISTRY DEPARTMENT 45-4-A 21 TSC (TP3) 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 WALEPOLE, ROBERT 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 492 UNIT 3 FSS 53' ONE STOP SHOP 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 4ooi'

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

]-*En-er"gy 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 En tergy CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES TO: DISTRIBUTION DATE: 10/1/07 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please review to verify receipt, incorporate the document(s) into your controlled document file, properly disposition superseded, void, or inactive document(s).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

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INDIAN POINT 3 TECHNICAL SPECIFICATION BASES INSTRUCTIONS FOR UPDATE: 24-10/05/2007 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.

Remove PaensertPRageý,,

TAB - List Of Effective Sections List of Effective Sections, Rev. 23 List of Effective Sections, Rev. 24 (6 pages) .(6 pages)

TAB 3.3 - Instrumentation B 3.3.1 (Rev. 2) B 3.3.1 (Rev. 3)

(58 pages) (57 pages)

TAB 3.8 - Electrical Power B 3.8.2 (Rev. 0) B 3.8.2 (Rev. 1)

(7 pages) (7 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 04/11/2007 B 3.5.3 1 4 08/10/2005 B2.0 SAFETY LIMITS:, B 3.5.4 0 9 03/19/2001 B 2.1.1 1 4 06/03/2005 B 3.6 C.ONTAINMENT B 2.1.2 1 4 06/03/2005 B 3.6.1 0 5 03/19/2001 B 3.0 LCO AND SR APPLICABILITYI . B 3.6.2 1 9 06/03/2005 B3.0 ] 31 18 11107/2006 B 3.6.3 0 17* 03/19/2001 B 3.1 REACTIVITY 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 B 3.7 PLANT SYSTEMS.

B 3.1.8 0 7 03/19/2001 B 3.7.1 2 6 06/03/2005 B 3.2 POWER DISTRIBUTION LIMITS B 3.7.2 2 10 05/16/2007 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 B 3.3 INSTRUMENTATION B 3.7.7 1 4 12/17/2004 B 3.3.1 3 57 10/05/2007 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 B 3.4 REACTOR 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-34.2- -- 0. 3 319/200 B78 :ELECTRICAL POWER 7 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 1 7 10/05/2007 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 3.4.7 1 7 04/11/2007 B 3.8.5 0 4 03/19/2001 B 3.4.8 0 4 03/19/2001 B 3.8.6 0 8 03/19/2001 B 3.4.9 3 5 06/03/2005 B 3.8.7 1 8 06/20/2003 B 3.4.10 0 5 03/19/2001 B 3.8.8 1 4 06/20/2003 B 3.4.11 1 7 08/10/2005 B 3.8.9 2 14 06/20/2003 B 3.4.12 2 19 08/10/2005 B 3.8.10 0 4 03/19/2001 B 3.4.13 4 7 04/11/2007 B 3.9 REFUELING OPERATIONS' B 3.4.14 0 10 03/19/2001 B 3.9.1 '1 4 07/06/2006 B 3.4.15 3 7. 08/10/2005 B 3.9.2 0 4 03/19/2001 B 3.4.16 2. 6 08/10/2005 B 3.9.3 2 7 06/03/2005 B 3.4.17 0 8 04/11/2007 B 3.9.4 0 4 03/19/2001 B 3.5 ECCS B 3.9.5 0 4 03/19/2001 B 3.5.1 1 B 3.9.6 2 3 04/11/2005 10 10/27/2004 B 3.5.2 2 13 09/16/2005 INDIAN POINT 3 Pagel of 6 Revision 24

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'.

BASES UPDATE PACKAGE0'1-'03.T01T, 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.

___ ____BASES. UPDATE-PACKAGE'02-.051S801 Table of 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 BASES 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.'

..BASESIU PDATEACAGG'0-0i02202* :

B 3.3.2 2 01/22/02 Clarify starting logic of 32 ABFP per EVL-01-3-078 MULTI,

.. .Rev--.

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 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 UDATERAKC GEe052 %93002. ."

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 24

TECHNICAL SPECIFICATION BASES REVISION HISTORY AFFECTED EFFECTIVE SECTIONS REV DATE DESCRIPTION

________ ___ BASES' UPDT PACKAGE0400 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 ..., ,.::.BASES"'UP DAT E, A K -GE$ _

B 3.3.8 1 03/17/2003 Changes to reflect Tech Spec Amendment 215 regarding B 3.7.13 1 implementation of Alternate Source Term analysis B 3.9.3 1 methodology to the Fuel Handling Accident.

BASES UPDATE -RACKGQ80383 B 3.4.9 1 03/28/2003 Changes to reflectTech Spec Amendment 216 regarding relaxation of pressurizer level limits inMODE 3.

K BASES UPDATEPAC1GE 09-62003 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-P3-2002-.

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 022.

B 3.8.9 2 06/20/2003 BAESUPDATE PACAG ;0,'1 0270 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-OperationalTest-(-T-ADOT-)-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.

,_,_".._ . *.!. BASES UPDATE PACKAGEIýI 1'21 004-B 3.7.7 1 12/17/2004 Addition of valves CT-1300 and CT-1.302 to Surveillance SR 3.7.7.2 to verify that all city water header supply isolation valves are open. Reflects Tech Spec Amendment 218.

____________UPDATEiPACKA E12101240 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 24

TECHNICAL SPECIFICATION BASES REVISION HISTORY AFFECTED EFFECTIVE SECTIONS REV DATE DESCRIPTION BASES UPDATE PACKAGE 13-022505 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.

_BASES UPDATE PACKAGE 14-030705 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 UPDATE PACKAGE 15-041105 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.

BASES.UPDATE PACKAGE 16-060305 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 24

TECHNICAL SPECIFICATION BASES REVISION HISTORY AFFECTED EFFECTIVE SECTIONS REV DATE DESCRIPTION BASES UPDATE PACKAGE 17-081005 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 BASES UPDATE PACKAGE 18-091605 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 UPDATE PACKAGE 19-110405 B 3.8.1 3 11/04/2005 Include operability criteria for 138 kV and 13.8 kV offsite circuits.

BASES UPDATE PACKAGE 20-070606 B 3.9.1 1 07/06/2006 Clarification on effective method for ensuring shutdown margin.

INDIAN POINT 3 Page 5 of 6 Revision 24

TECHNICAL SPECIFICATION BASES REVISION HISTORY RASPS UPDATE PACKAGF 21-1107200A 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)

SBASES UPDATE PACKAGE 22-04112007 TOC 3 04/11/2007 Implement TS Amendment 233 related to steam generator B 3.4.4 1 tube integrity.

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 BASES UPDATE PACKAGE 23-05162007 B 3.7.2 2 05/16/2007 Removal of extraneous information regarding testing" frequency.

BASES UPDATE PACKAGE 24-10052007 B 3.3.1 3 10/05/2007 B 3.3.1 - The TS and bases currently allow a normal B 3.8.2 shutdown without the SR testing by reducing power below

-the modes--of,appl,;cability-.or-SR-3-,,3.1.8.- Clarify 'that- - .

testing is not required if such testing was done within the prior 92 days, even if a mode of applicability was still met.

B 3.8.2 - Clarify LCO with regard to the required power sources for modes 5 and 6.

INDIAN POINT 3 Page 6 of 6 Revision 24

RPS Instrumentation "B3:3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protection System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuelFdesign limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences (AOOs) and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents.'

The protection and monitoring systems have been designed to assureý safe operation of the reactor. This is achieved by specifying limiting safety system settings. (LSSS) in terms of parameters directly monitored by-the RPS, as well, as specifying LCOs on other reactor system parameters and equipment performance.

'The LSSS, defined in this specification as the Allowable.Value, in conjunction with the LCOs,.establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During AOOs, which are those events expected to occur one or. more times. during the unit life, the acceptable limits are:

1. The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained, above the Safety Limit (SL) value to prevent' departure from nucleate boiling .(DNB);

2. Fuel centerline melt. shall not occur; and

3. The RCS pressure SL of 2735 psig shall not be exceeded.,

Operation within the SLs of Specification 2.0, "Safety Limits..

(SLs)," also maintains the above values and assures that offsite dose will be within the 10 CFR 50 and 10 CFR.50.67 criteria during' AOOs.

(continued)

INDIAN POINT 3 B,3.3.1-1 :Revision 3

RPS Instrumentation B 3.3.1 BASES BACKGROUND Accidents are events that are analyzed even though they are not (continued) expected to occur during the unit life. The acceptable limit during accidents is that offsite dose shall be maintained within an acceptable fraction of 10 CFR 50.67 limits. Different accident categories are allowed a different fraction of these limits, based.

on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

The RPS instrumentation is segmented into four distinct but interconnected modules as described in FSAR, Chapter 7 (Ref. 1), and as identified below:

1. Field transmitters or process sensors: provide a measurable electronic signal based upon the physical characteristics of the parameter being measured;

2. Signal Process Control and Protection System including Analog Protection System, Nuclear Instrumentation System (NIS), field.

contacts, and protection channels: provides signal conditioning, bistable setpoint comparison, process algorithm actuation, compatible electrical signal output to protection system devices, and control board/control room/miscellaneous indications;

3. RPS automatic initiation relay logic, including input, logic, and output: initiates proper unit shutdown in accordance with the defined logic, which is based on the bistable outputs from

. ..... the signalpprocess control and _protection system; and

4. Reactortrip switchgear, including-reactor trip breakers (RTBs) and bypass breakers: provides the means to interrupt power to the control rod drive mechanisms (CRDMs) and allows the rod cluster control assemblies (RCCAs), or "rods," to fall into the core and shut down the reactor. The bypass breakers allow testing' of the RTBs at power.

(continued)

INDIAN POINT 3 B 3.3.1-2 Revision 3

RPS Instrumentation B 3.3.1 BASES BACKGROUND Field Transmitters or Sensors (continued)

To meet the design demands for redundancy and reliability, more than

-,,one, and often as many. as four, field transmitters or sensors are used to measure unit parameters. To account for the calibration tolerances and instrument, drift, which are assumed to occur between calibrations, statistical allowances are provided in the Allowable Values. The OPERABILITY of each transmitter or sensor can be evaluated when its "as found" calibration data are compared against its documented Allowable Value.

Signal Process Control and Protection System Generally, three or four channels of process control equipment'are used for the signal processing of unit parameters measured by the field instruments. The process control equipment-provides signal conditioning, comparable output signals for instruments located on the main control board, and comparison of measured input signals with setpoints established to ensure that actuation will occur within the limits assumed in the' accident analyses (Ref. 3). If the measured value of a uni~t parameter. exceeds the predetermined

.setpoint, an output from a bistable is forwarded to the RPS relay logic. Channel separation is maintained up to and through the actuation logic. However, not all. unit parameters require four channels of sensor measurement and signal processing. Some unit parameters provide input only to the RPS relay logic, while others provide input to the RPS relay logic, the main control board, the

unit computer, and one or more control systems.

I Gen-ral-ly-,if a parameter is used only for input to the protection circuits, three channels with a two-out-of-three 'logic are sufficient to provide the required reliability and redundancy., If

.,one channel fails in a direction that would not result in a partial Function trip,.,the Function is still OPERABLE with a two-out-of-two logic. If one channel, fails, such that a partial Function trip occurs, a trip will not occur and the Function -isstill OPERABLE with a one-out-of-two logic.

Generally, if a parameter is used for input to the RPS relay logic and a control function, four channels with a two-out-of-four logic

.(continued)

INDIAN POINT 3 B 3.3.1-3 Revision 3.

RPS Instrumentation B,3.3.1 BASES BACKGROUND are sufficient to provide the required reliability and redundancy..

(continued) The circuit must be able to withstand both an input failure, to the control system, which may then require the protection function-actuation, and a single failure in the other channels providing the..

protection function actuation. Again, a single failure will. neither cause nor prevent the protection function actuation. These requirements are described in IEEE-279-1968'(Ref. 4). The actual number of channels required for each unit parameter is specified in Reference I and discussedlater in these Technical Specification, Bases.

Two logic channels are required-to~ensure no single randomfailure of a logic channel will disable the RPS. The logic channels are designed such that testing required while the reactor is at power may be accomplished without causing trip.

Trip Setpoints and Allowable Values The following describes the relationship between the safety limit,.

analytical limit, allowable value and channel component calibration acceptance criteria:

a. A Safety Limit (SL) is a limit on the combination of THERMAL POWER, RCS highestloop average temperature, and RCS'pressure needed to protect the integrity of physical barriers, that guard against the uncontrolled release'of radioactivity (i.e.,'fuel, fuel cladding, RCS pressure boundary and containment). The safety limits are identified in Technical Specification 2.0,

............... Safety. Limits (SLs) . _.... _ .. . .

b. An Analytical Limit (AL) is the trip actuation point used as an input to the accident analyses presented in FSAR, Chapter 14 (Ref. 3). Analytical limits are developed from event analyses models which consider parameters such as process delays, .rod insertion times, reactivity changes, instrument response times, etc. An analytical limit for a trip actuation point is established at a point that will ensure that a Safety Limit (SL) is not exceeded'.'

(continued)

INDIAN POINT 3 B 3.3.1-4 Revision 3

RPS Instrumentation B 3.3.1 BASES BACKGROUND c. An Allowable Value (AV) is the limiting actuation point for (continued). the entire channel of a trip function that will ensure, within the required level of confidence, that sufficient allocation exists between this actual trip function actuation point and the analytical limit. The Allowable Value is more conservative than the Analytical..Limit to account for instrument uncertainties that. either are not present or are not measured during periodic testing. Channel uncertainties that either are not present.or, are not measured during periodic testing may include design basis accident temperature and radiation effects (Ref. 5) or process dependent effects. The channel allowable value for each RPS function is controlled by Technical Specifications and is listed in Table 3.3.1-1, Reactor Protection System Instrumentation.

d. Calibration acceptance criteria are established by plant administrative programs for the components, of a channel (i.e.,

required sensor, alarm, interlock, display, and trip function).

The calibration acceptance criteria are established to ensure, within the required level of-confidence, that the Allowable Value for the entire channel will not be exceeded during the calibration interval.

A description of the methodology used to calculate the channel allowable values and calibration acceptance criteria is provided in References 6 and 8.

Setpoints in accordance with the Allowable Value ensure that SLs are not, violated during AOOs (and that the consequences of DBAs will be acceptable, _prd- ngth-e--n iit s operate-d-fr-w-ithTiin-the-LCOsC t the onset of the AO0 or DBA and the equipment functions as designed).

Each channel of the relay logic protection system can be tested on line to verify that the signal or setpoint accuracy is within the specified allowance requirements of calculations performed in accordance with Reference 6 that arebased on analytical limits consistent with Reference 3. Once a designated channel is taken out of service for testing, a simulated signal is injected in place of the field instrument signal.

(continued)

INDIAN POINT 3 B 3.3.1-5 Revision 3

RPS Instrumentation B 3.3.1 BASES BACKGROUND The process equipment for the channel in test is then tested, (continued) verified, and calibrated. SRs for the channels are specified in the SRs section. The Allowable Values listed in Table 3.3.1-1 and the Trip Setpoints calculated to ensure that Allowable Values are not exceeded*during the calibration interval are based on the methodology described in Reference 6, which incorporates all of the known uncertainties applicable for each channel. All field sensors and signal processing equipment for these channels are assumed to operate within the allowances of these uncertainty magnitudes.

Relay Loqic Protection System Relay logic is used for the decision logic processing of outputs from the signal processing equipment bistables. To meet the redundancy requirements, two trains of relay logic, each performing the same functions, are provided. If one train is taken out of.

service for maintenance *or test purposes, the second train will provide reactor trip and/or ESF actuation for the unit. .If both trains are taken out of service or placed in test, a reactor trip will result. Each train is packaged in its own cabinet for physical and electrical separation to satisfy separation and independence requirements., The system.has been designed to trip in the event of a loss of power, directing the unit to a safe shutdown condition.

The relay logic performs the decision logic for actuating a reactor trip or ESF actuation, generates the electrical output signal that will initiate the required trip or actuation, and provides the status, permissive, and annunciator output signals to the control room. .

The bistable outputs from the signal processing equipment are sensed by the relay logic equipment and combined into logic matrices that represent combinations indicative of various unit upset and accident transients. If a required logic matrix combination is completed, the system will initiate a reactor trip or send actuation signals via master and slave relays to those components whose aggregate Function best serves to alleviate the condition and restore the unit to a safe condition. Examples are given in the Applicable Safety Analyses, LCO, and Applicability sections of this Bases.

(continued)

INDIAN POINT 3 B 3.3.1-6 Revision 3

RPS Instrumentation B 3.3.1 BASES BACKGROUND Reactor Trip Breakers (continued)

The.RTBs are in the electrical power supply line from the control rod .drive motor, generator set. power supply tothe CRDMs. Opening of the RTBs interrupts power to the CRDMs, which allows the shutdown rods and control rods to fall into the core by gravity. Each RTB is equipped with a.bypass breaker to allow testing of the.RTB while the unit is at power. During normal operation the output from the reactor protection.system.is a voltage signal that energizes the undervoltage. coils inn,,.the.RTBs andbypass breakers, if in use. When the required logic matrix combination is completed, the reactor protection system output Voltage signal is removed, the undervoltage coils are de-energized, the breaker trip lever is actuated by the de-energized undervoitage coil,.and the RTBs and bypass breakers are

tripped open., This allowsthe shutdown rods and control rods to fall into the core. In addition to the de-energization of the undervoltage coils, each breaker'is also equipped with a shunt trip device that is energized to tripthe:breaker open upon receipt of a reactor trip signal from the reactorprotection system. 'Either the undervoltage coil or the shunt trip mechanism is sufficient by itself, thus providing-a-diverse trip mechanism.

There are two reactor trip breakers in series~so that opening either will interrupt power to, the control rod drive mechanisms (CRDMs) and' allowthe rod cluster control assemblies (RCCAs), or "rods,"' to fal.l into the core and shut down the reactor. Each reactor trip breaker.

has a parallel reactor trip.bypass breaker that is normally open.

This feature allows testing.of the reactor trip breakers at power.

A trip signal from RPS logic train A.will trip reactor trip breaker A and reactor trip bypass breaker B; and,"a trip signal from logic train'B will trip reactor trip breaker B and reactor trip bypass breaker A.. During normal operation, both reactor trip breakers are closed and both reactor trip.bypass breakers are open. An interlock

.trips both reactor.trip bypass breakers if an attempt ismade to close a reactor.trip~bypass breaker when theother reactor trip

.bypass breaker is already closed.

A trip breaker train.consists of both the reactor trip breaker and reactor trip bypass breaker associated with a single RPS logic train if the breaker is racked in, closed,'and capable of supplying power (continued)

INDIAN POINT 3 B 3..3.1-7 Revision-3

RPS Instrumentation

!B 3.3.1 BASES BACKGROUND to the CRD System. Thus, the train consists of the main breaker;,

(continued) or, the main breaker and bypass breaker associated with this same" RPS logic train if both the breaker and bypass are racked in, closed, and capable of supplying power to the CRD System.

The RPS decision logic:Functions are described in the functional,-

diagrams included in Reference 2. In addition to the reactor protection and ESFAS trips,.,the various "permissive interlocks" that are associated with unit conditions are also.described.

When any one RPS train is.taken out of-service for testing, the other train is capable of providing unit monitoring and protection until the testing has been completed.

APPLICABLE SAFETY ANALYSES, LCO,,and APPLICABILITY The RPS functions to maintain the Safety Limits (SLs) during.all Abnormal Operating Occurrences (AOOs) and mitigates the consequences of DBAs in all MODES in which the Rod Control system is capable of rod withdrawal and one or more rods not fully inserted..

Each of the analyzed accidents and transients can be detected by one or more RPS Functions. The accident analysis described in Reference 3 takes credit for most'RPS trip Functions. RPS trip Functions not specifically credited in the accidentanalysis are qualitatively credited in the safety analysis and the NRC staff approved licensing basis. These RPS trip Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. They mayalso serve as backups to RPS trip Functions that were credited in the accident analysis.

The LCO requires all instrumentation performing an RPS Function, listed in Table 3.3.1-1in.the accompanying LCO, to be OPERABLE.

Failure of any instrument renders the affected channel(s) inoperable and reduces the reliability of the affected Functions. .

(continued)

INDIAN POINT 3 B 3'3.1-8 Revision 3

RPS Instrumentation B 3.3.1 BASES.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued).

The.LCO generally requires OPERABILITY of four or three channels in each instrumentation Function, two channels of Manual Reactor Trip, and two trains in each AutomaticTrip Logic Function. Generally, four'OPERABLE instrumentation channels in a two-out-of-four configuration are required when one RPS'channel is also used as a control system input.-Isolation-amplifiers.prevent a control system.

" failure from affecting the protection system (Ref. 1). This.

configuration.accounts for the possibility of the shared channel failing in such a mariner that it creates a transient that requires.

- RPS action-. Inthis case, the RPS will still provide protection, even with random failure of one of the other three protection channels. Three OPERABLE instrumentation channels in a two-out-of-three configuration are generally required when there is no potential for control system and protection system interaction that could simul-taneously create a need for RPS trip and disable one RPS channel. The. two-out-of-three and two-out-of-four

... configurations allow one channel to be tripped during maintenance or testing without causing a reactor trip.' Specific exceptionsto the

.above general philosophy exist and are discussed below.

Reactor Protection System Functions The safety analyses and OPERABILITY requirements applicable to each RPS Function are discussed below:

1. Manual'Reactor Trip -

The Manual Reactor Trip-ensures that the control room operator can initiate a reactor trip at any time by using either of two reactor trip push buttons in the control room. A Manual .

Reactor Trip accompl-ishes the same results as any one of the,.

automatic trip Functions., It is used by the reactor operator; to shut down thereactor whenever any parameter is rapidly trending toward its Trip Setpoint.

The LCO requires two Manual Reactor Trip channels to be OPERABLE. Each channel is controlled by a manual reactor trip push button. Each channel activates the reactor trip breaker (continued)

INDIAN POINT 3 B 3.3.1-9-' Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) in both trains. Two independent channels are required to be OPERABLE so that no single random failure will disable the Manual Reactor Trip Function.

In MODE 1 or 2, manual initiation of a reactor trip must be OPERABLE. These are the MODES in which the shutdown rods and/or control rods are partially or fully withdrawn from the core. In MODE 3, 4, or 5, the manual initiation Function must also be OPERABLE if one or more shutdown rods or control rods are withdrawn or the Rod Control System is capable of withdrawing the shutdown rods or the control rods. In this condition, inadvertent control rod withdrawal is possible. In MODE 3, 4, or 5, manual initiation of a reactor trip does not have to be OPERABLE if the Rod Control System is not capable of withdrawing the shutdown rods or control rods and if all rods are fully inserted. If the rods cannot be withdrawn from the core, or all of the rods are inserted there is no need to be able to trip the reactor. In MODE 6, neither the shutdown rods nor the control rods are permitted to be withdrawn and the CRDMs are disconnected from the control rods and shutdown rods.

Therefore, the manual initiation Function is not required.

2. Power Range Neutron Flux The NIS power range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NISpower NI -range.detectors-provide--input -to -the-Rod-Control .....

System'and Turbine Control System. Four channels of NIS are required because the actuation logic must be able to withstand an input failure to the control system which may then require the protection function-actuation and a single failure in the other three channels providing the protection function actuation. Note that this Function also provides a signal to prevent automatic and manual rod withdrawal prior to initiating a reactor trip.

(continued)

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RPS Instrumentationi.

B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABiLITY (continued)'

Limiting further rod withdrawal may terminate the transient and eliminate the need*to trip the reactor.

a. Power Range Neutron Flux-High The Power Range Neutron Flux-High trip Function ensures that-protection is provided, from all power levels, against a positive reactivity excursion leading to DNB .

during power,operations. These can be caused by rod withdrawal or reductions in RCS temperature.

The LCO requires all four of the Power Range Neutron-Flux

-High channels to be OPERABLE.. These channels are considered OPERABLE during required Surveillance .tests that require insertion of a test signal if the channel remains. untripped and capable of tripping due to an increasing neutron flux signal. During MODE 2 Physics Tests, only 3 channels are required because the output from one detector is used for test instrumentation.

In MODE I or 2, when a positive reactivity excursion could occur, the Power Range Neutron Flux-High trip must be OPERABLE. This Function will terminate the reactivity:,

excursion and shut down the reactor prior to reaching a power level that could damage the fuel. In MODE 3, 4, 5, or 6, the-NIS power rangedetectors cannot detect neutron levels in this range. In.these MODES, the Power Range Neutron Flux-High does not have to be OPERABLE because the reactor is shut down and reactivity excursions into

. the power~range are extremely unlikely. Other RPS Functions and administrative controls provide-protection against reactivity additions when in MODE. 3, 4, 5, or 6.

(continued)

INDIAN POINT 3 B 3.3.1-11 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The Power Range Neutron Flux-High Allowable Value and Trip Setpoint are in accordance with Consolidated Edison Company of New York, Inc. Indian Point Nuclear Generating Station Unit No. 3 Plant Manual Volume VI:

Precautions,'Limitations, and Setpoints, March 1975 (Ref.

8).

b. Power Range Neutron Flux-Low The LCO requirement for the Power Range Neutron Flux- Low trip Function ensures that protection is provided against a positive reactivity excursion from low power or subcritical conditions.

The LCO requires all four of the Power Range Neutron Flux

-Low channels to be OPERABLE. During MODE 2 Physics Tests, only 3 channels are required because the output from one detector is used for test instrumentation.

In MODE 1, below the Power Range Neutron Flux (P-10 setpoint), and in MODE 2, the Power Range Neutron. Flux-Low trip.must be OPERABLE. This Function may be manually blocked by the operator when two out of four power range channels are greater than approximately 10% RTP (P-10.

setpoint). This Function is automatically unblocked when three out of four power range channels are below the P-10 setpoint. Above the P-IO setpoint, positive reactivity additions are mitigated by the Power Range Neutron Flux-High trip Function.

In MODE 3, 4, 5, or 6, the Power Range Neutron Flux-Low trip Function does not have to be OPERABLE because the reactor is'shutdown and the NIS power range detectors cannot detect neutron levels in this range.. Other RPS (continued)

INDIAN POINT 3 B 3.3.1-12 Revision 3

RPS Instrumentation B 3.3.1 0 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) trip Functions and administrative controls provide protection against positive reactivity additions or power excursions in MODE.3, 4, 5, or 6.

The Power Range Neutron Flux-Low Allowable Value and Trip

'Setpoi.nt are in accordance with.Consolidated Edison Company of New York, Inc. Indian Point Nuclear Generating Station Uni,t No. 3 Plant Manual Volume VI:

Precautions, Limitations, and Setpoints, March 1975 (Ref.

8).

Intermediate Range Neutron Flux A3 The Intermediate Range Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal'accident from a subcritical condition during startup. This trip,.Function provides redundant protection to the Power Range Neutron Flux-Low Setpoint trip. Function.

Therefore, only one of the two channels of Intermediate Range Neutron Flux is Required to be OPERABLE in the Applicable MODES. Either of. the two channels can be used to satisfy this requirement.; The NIS intermediate range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS intermediate range detectors do not provide any input to control systems. Note that this Function also provides a signal to prevent, automatic and manual rod withdrawal prior to initiating a reactor trip. Limiting further rod withdrawal may terminate the transient'and

.elimina~te th ne to The LCO requires one channel of Intermediate Range Neutron Flux to. be OPERABLE. One OPERABLE channel is sufficient to provide redundant protection to.the Power Range Neutron. Flux-Low Setpoint trip. Function.

(continued)

INDIAN POINT 3 B 3.3.1-13 Revision 3

RPS Instrumentation B 3.3.1, BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Table 3.3.1-1 identifies the Technical' Specification Allowable Value for this trip function as not applicable (NA) because LCO 3.3.1, Function.2.b, Power. Range Neutron Flux-Low, is used to bound the analysis for an uncontrolled control rod assembly.

withdrawal from a-subcritical condition. The'surveillance acceptance criterion used for this function is

28% RTP.

This value~was established based on Indian Point Nuclear Generating Station Unit No. 3 Plant Manual Volume VI:

Precautions, Limitations,'and Setpoints, March 1975, (Ref. 8).

Because'this trip Function is'important only during startup, there is generally no need to disable channels for testing.

while the Function is required to be OPERABLE. Therefore, a third channel is unnecessary.

The Intermediate'RangeNeutron Flux trip must be OPERABLE in MODE 1 below the P-1O setpoint, and in MODE 2 above the P-6 setpoint, when there is a potential for an uncontrolled RCCA bank rod withdrawal accident during reactor startup.. Above the P-1O setpoint, the Power Range Neutron Flux-High Setpoint trip provides core protection for a rod withdrawal accident. In MODE 2, below the P-6 setpoint, the source Range Neutron Flux Trip provides backup core protection for reactivity accidents.'

In MODE 3, 4, or 5, the Intermediate Range Neutron Flux trip does not have to be OPERABLE because the control rods must be fully inserted and only the.shutdown rods may be withdrawn.

The reactor cannot 'be started up'in this condition. The core also has the required SDM to mitigate the consequences.of a positive reactivity addition 'accident. In MODE 6, all rods are fully inserted and thecore has a required increased SDM.

Also, t.he NIS intermediate ra'nge detectors cannot detect

.neutron levels present in this MODE.

(continued)

INDIAN POINT 3 B 3.3.1-14 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and.APPLICABILITY .(continued)

4. Source Range Neutron Flux The LCO requirement for the Source Range Neutron Flux trip.

Function ensures that protection is provided against an uncontrolled RCCAbank rod withdrawal accident from a subcritical condition during, startup. This trip Function:

provides redundant protection to the Power Range Neutron Flux-Low trip Function. Therefore, only one of the two channels of Source Range Neutron Flux is.required to be OPERABLE in the Applicable MODES. Either of the two channels can be used to satisfy this requirement. In MODES 3, 4, and 5, administrative controls also prevent the uncontrolled&

withdrawal of rods. The NISsource range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS source range detectors do not provide any inputs to control systems. The source range' trip is the only RPS automatic protection function required in MODES 3, 4, and 5 when rods are capable of withdrawal or one or more rods are not, fully inserted.

The LCO requires one channel of Source Range Neutron Flux to be OPERABLE. One OPERABLE channel is sufficient to provide redundant protection to the Power Range Neutron Flux-Low Setpoint trip Function.

Table 3.3.1-1 identifies the Technical Specification Allowable Value for this trip function as not applicable (NA) because LCO 3.3.1, Function 2.b, Power Range Neutron Flux-Low, is used to bound the analysis for an uncontrolled control rod assembly withdrawal from a subcritical condition. The surveillance acceptance criterion used for this function is *6.0 E+5 counts per second.

The Source Range Neutron Flux Function provides protection for control rod withdrawal from subcritical.' The Function also provides visual neutron flux indication in the control room.

(continued)

INDIAN POINT 3 B 3.3.1-15 Revision 3 -

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

In MODE 2 when below the P-6 setpoint and in MODES 3, 4, and 5, when there is a potential for an uncontrolled RCCA bank withdrawal accident, the Source Range Neutron. Flux trip must be OPERABLE. Abovethe-P-6 setpoint, the Intermediate Range Neutron Flux trip and the Power Range Neutron Flux-Low trip will provide core-protection for reactivity accidents. Above the P-6 setpoint, the NIS-source range detectors are de-energized.

In MODEs'.3, 4, and 5 with all rods'fully inserted and the Rod Control System not capable of rod withdrawal, and in MODE 6, the outputs of this function to the RPS logic are not required to be OPERABLE. The requirements for the NIS source range detectors in MODE 6 are addressed in LCO 3.9.2, "Nuclear Instrumentation."

5. Overtemperature AT The Overtemperature AT trip Function is provided to ensure that the design limit DNBR is met. This trip Function also limits the range over which the Overpower AT trip Function must provide protection. The inputs to the Overtemperature AT trip include pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop AT assuming full reactor coolant flow. Protection from violating the DNBR limit is assured for those transients that are slow

....... respect.to delays from -the -core-to-themeasurement -

system.'

The Overtemperature AT trip Function uses each loop's AT as a measure'of reactor power and is compared with a setpoint that is automatically Varied with the following parameters:

reactor coolant average temperature-the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; pressurizer pressure-the Trip Setpoint is varied to correct for changes in system pressure; and (continued)

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V.VI IU I S

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

axial power distribution-f(AI), the Trip Setpoint is varied to account for imbalances in the axial power distribution as detected by the NIS upper and lower power range detectors. If axial peaks are greater than the Technical:Specification limit, as indicated by. the difference.between the.upper and lower NIS power range detectors,ý.the Trip Setpoint is reduced in accordance with Note 1 of Table 3.3.1-1.

Dynamic compensation is included for system piping delays from the core to the temperature measurement system.

The Overtemperature AT trip Function is calculated for each loop as described in Note 1 of Table 3.3.1-1. Trip occurs if Overtemperature AT is-indicated in two loops. The.

pressure and temperature signals are used for other control functions. Therefore, the actuation logic is designed to withstand an input failure to. the control system, which may then require the protection function actuation,.and a single failure in the other channels providing the protection function actuation.

The LCO requires all four channels of the Overtemperature AT trip Function to be OPERABLE. Note:that the Overtemperature AT Function receives input from channels shared with other RPS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions..

In MODE 1 or 2, the Overtemperature AT trip must be OPERABLE to prevent DNB.. In MODE 3, 4, 5,.or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient-heat production to be concerned.about DNB.

6. Overpower AT The. Overpower AT trip Function ensures that protection is provided.to ensure the integrity of the fuel (i.e., no. fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions. This trip Function also limits (continued)

INDIAN POINT 3 B 3.3.1-17 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) the required range of the Overtemperature AT trip Function and provides a backup to the Power Range Neutron Flux-High Setpoint trip. The Overpower AT trip Function ensures that the allowable heat generation rate (kW/ft) of the-fuel is not exceeded. It uses the AT of each loop as a measure of reactor power with a setpoint that is.automatically varied with the following parameters:

reactor coolant average temperature-the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; and rate of change 'of reactor coolant average temperature-including a constant determined by.dynamic considerations that provides compensation for the delays between the core and the temperature measurement system.

The Overpower AT trip Function is calculated for each loop as per Note 2 of Table 3.3.1-1. Trip occurs if Overpower AT is indicated in two loops. The temperature signals are used for other control functions. Therefore, the actuation logic~is designed to withstand an input failure to the control system, which may then require the protection function actuation and a single failure in the remaining channels providing the

-- proteclti-on-f un-ction'-a~ctuatl oh".

The LCO requires four channels of the Overpower AT trip Function to be OPERABLE_. Note that the Overpower AT trip Function receives input-from channels shared with other RPS.

Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overpower AT trip Function must be OPERABLE. These are. the only times that enough heat is generated in the fuel to be'concerned about the heat generation rates and overheating of the fuel. In MODE 3, 4, (continued)

INDIAN POINT'3 B 3.3.1-18 Revision 3

RPS Instrumentation B 3.3.1:

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY .(continued) 5, or 6, this.trip Function does not haveto be OPERABLE because the reactor is not operating and there is. insufficient heat production to be concerned about fuel overheating and fuel damage.

7. Pressurizer Pressure The.same sensors provide input to the Pressurizer Pressure-'

High and -Low trips and the Overtemperature AT trip. -The Pressurizer Pressure channels are also used to provideinputto the Pressurizer Pressure Control System. Therefore, the actuation logic is designed to withstand an input failure to.:_

the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. Note that .the plant design and this LCO require 4 channels for the Pressurizer Pressure-Low trips but requires only 3 channels of Pressurizer Pressure-High. This difference recognizes the role of pressurizer code safety valves in response to a high,-

pressure condition.

a. Pressurizer Pressure-Low The Pressurizer Pressure-Low trip Function ensures that protection is provided against violating the DNBR limit due to low pressure.

................ .......... .... ......... T f -L* YU B nn T O - s rfz Tr s e- -i The LCO requires-four chann'els of PressurizerPressure-Low to be OPERABLE.

In MODE 1, when DNB is a major concern, the Pressurizer Pressure-Low trip. must be OPERABLE. This trip Function.

is automatically enabled on increasing power.by the P-7 interlock (NIS power range: P-1O or turbine first stagege pressure greater. than approximately 10% of full power equivalent). On decreasing power, this trip Function is".

automatically blocked below:P-7. Below the P-7.setpoint, no conceivable power distributions can occur that would cause DNB concerns..

continued)

INDIAN POINT 3 B 3.3.1-19 Revision 3

RPS Instrumentation B 3.3.1

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

b. Pressurizer Pressure-High The Pressurizer Pressure-High trip Function ensures that protection is°'provided against overpressurizing *the RCS.

This trip-FUnction operates in conjunction with the pressurizer relief and *safety valves to prevent RCS overpressure Conditions.

The LCO requires three channels of the Pressurizer Pressure-High' to be OPERABLE.

The Pressurizer Pressure-High Allowable Value is selected to be below the pressurizer safety valve actuation pressure and above the poweroperated relief valve (PORV) setting.. This *setting minimizes challenges to safety valves while avoiding unnecessary reactor trip for those pressure increases that can be controlled by the PORVs.

In MODE 1 or 2; the Pressurizer Pressure-High trip must be OPERABLE to help prevent RCS overpressurization and minimize challenges to the safety valves. In MODE 3, 4, 5, or.6, thePressurizer Pressure-High trip Function does not have to be OPERABLE because transients that could cause an overpressure condition will be slow to occur. Therefore, the operator will have sufficient time to evaluate unit conditions and take corrective actions.

-Additionally, low temperature overpressure protection systems provide overpressure protection when RCS temperature is less than the LTOP arming temperature specified in LCO 3.4.12, Low Temperature Overpressure Protection (LTOP).

(continued)

INDIAN POINT 3 .B 3.3.1-20 IRevision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO,tand APPLICABILITY (continued)

8. Pressurizer Water Level-High The,.Pressurizer Water Level -High trip Function provides a backup signal forthe Pressurizer Pressure-High trip and also provides protection.against water relief through the pressurizer safety Valves. These valves are designed to pass steam in order.,to achieve their design energy removal: rate. A reactor trip is actuated prior to the pressurizer becoming water solid.. The-LCO requires three channels of Pressurizer Water, Level-High to be OPERABLE.' The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is.not required to address control/protection interaction concerns because the level channels do not actuate the safety valves, and the.high pressure reactor trip is set below the safety valve setting.

Therefore, with the slow rate of charging available, pressure overshoot due to level channel. failure cannot cause the safety valve to lift before reactor high pressure trip.

In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock.

On decreasing power, this trip.Function is automatically blocked below P-7. Below the P-7 setpoint, transientsthat could raise the pressurizer water level will be slow and the o-operator will have sufficient time to evaluate unit conditions and take corrective actions.

(continued)

INDIAN POINT 3 B 3.3.1-21 Revision 3

RPS Instrumentation B 3.3.1

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

9. Reactor Coolant Flow-Low

a. Reactor Coolant Flow-Low'(Single Loop)

The' Reactor Coolant Flow-Low (Single Loop) trip. Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow. Above the P-8 setpoint, a loss of flow in any RCS loop will actuate a reactor trip. Each RCS loop has three flow detectors to monitor flow. The flow signals are:not used for any control system input.

The LCO requires three Reactor Coolant Flow- Low channels per RCS loop to be OPERABLE.in MODE 1 above P-8; Each reactor coolant loop is considered to be a separate function. Therefore, separate condition entry is allowed for each loop.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 9.b) because of the lower power level and the greater margin to the design limit DNBR.

b. Reactor Coolant Flow-Low (Two Loops) .

The Reactor Coolant *Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. Each loop has three flow detectors to monitor flow.

The flow signals are not used for any control system input.

h I TA. nft1 INDIANf POtfNl a l *Ifnf....I-0: J.3.3--44 (continued)

Kev Is jon, a

RPS Instrumentation B 3.3.1

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The LCO requires three Reactor Coolant Flow-Low channels per loop. to be OPERABLE. Each reactor coolant loop is considered to be a separate function. Therefore, separate condition entry is allowed for each loop.

In MODE 1 above, the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor tripson low flow are automatically blocked since no

-conceivable power distributions could occur that would cause a DNB concern at this low power level. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops isautomatically enabled. Above the P-8 setpoint, a loss of flow in any one loop (Function9.a) will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.

10. Reactor Coolant Pump (RCP) Breaker Position Both RCP Breaker Position trip Functions operate to anticipate the Reactor Coolant Flow- Low trips to avoid RCS heatup that would occur before.the low flow trip actuates.

a. Reactor Coolant Pump Breaker Position (Single LOOP)

The RCP Breaker Position (Single Loop) trip Function ensures that protection is provided against violating the

. ~ * ...........

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position of each RCP breaker is monitored.: If one RCP m breaker is open above the P-8 setpoint, a reactor trip is initiated. This trip Function will. generate a reactor trip before the Reactor Coolant Flow-Low (Single Loop)

Trip Setpoint is reached.

The LCO requires one RCP Breaker Position channel per: RCP to be OPERABLE.. One OPERABLE channel is sufficient for this trip Function because the RCS Flow-Low trip alone provides, suffilcient protection of unit'SLs -for loss of (continued)

INDIAN POINT 3 B 3.3.1-23. Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) flow events. The RCP Breaker Position trip serves only to anticipate the low flow trip, minimizing the thermal transient associated with loss of a pump. Each reactor coolant loop.is considered to be a separate function.

Therefore, separate condition entry is allowed for each loop.

This Function measures only the discrete position (open or closed)'of'the RCP breaker, using a position switch..

Therefore, the Function' has no adjustable trip setpoint with which to associate an LSSS.

In MODE ' above the P-8 setpoint, when a loss of flow in any RCS loop could result in DNB conditions in the core, the RCP Breaker Position (Single Loop) trip must be OPERABLE. In MODE 1 below the P-8 setpoint, a loss of.

flow i~n two or more loops (Function 1O.b) is required to actuate a'reactor trip because of the lower power level and the greater margin to the design limit DNBR.

b. Reactor Coolant Pump Breaker Position (Two Loops)

The RCP Breaker Position (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in two. or more RCS loops. The position of each RCP breaker is monitored.

Above the P-7 setpoint a loss of flow in two or more loops will initiate a reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant Flow-Low (Two Loops) Trip S'etpoint is reached.

The LCO requires one RCP Breaker Position channel per RCP to be OPERABLE. One OPERABLE channel is sufficient for this Function because the.RCS Flow-Low trip alone provides. sufficient protection of unit SLs for loss of (continued)

INDIAN POINT 3 B 3.3.1-24* Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) flow events. The RCP Breaker Position trip serves only to anticipate the low flow trip, minimizing the thermal transient associated with loss of an RCP. Each reactor coolant loop i.s considered to be a separate function.

Therefore, separate condition entry is allowed for each loop.

This Function measures only the discrete position (open or closed) of the RCP breaker, using a position switch.

-Therefore,.the Function has no adjustable trip. setpoint with. which.to associate.an LSSS.

In MODE 1 above the P-7 setpoint and below the P-8 setpoint,.the.RCP Breaker Position (Two Loops). trip must be OPERABLE. Below the. P-7 setpoint, all reactor trips on loss of flow are automatically blocked since no conceivable power distributions could occur that would cause a DNB concern at this low power level. Above the P-7 setpoint, the reactor trip on loss of flow' in two RCS loops is automatically enabled. Above the P-8 setpoint,,

a loss of. flow in any one loop (Function-10.a) will actuate a reactor trip because of the higher power level..

and the reduced margin to the design limit DNBR.

11. Undervoltage Reactor Coolant Pumps (6.9 kV Bus)

The Undervoltage RCPs direct'reactor trip Function ensures that protection is provided against violating the DNBR limit due to a lsoffl'ow in two or more RCS loops. The voltage to each 6.9 kV bus used to power an RCP is monitored. Above the P-7 setpoint, a loss of Voltage detected on two or more RCP buses' will initiate' a'direct reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant Flow-Low (Two Loops) Trip Setpoint is reached. Time delays are incorporated into the Undervoltage RCPs channels associated.

with the direct reactor trip and are provided to prevent reactor trips due to-momentary electrical power transients.

(continued)

INDIAN POINT 3 B 3.3.1-2.5 Revision. 3

RPS Instrumentation B 3.3.1 BASES 0 APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The LCO requires one Undervoltage RCPs channel per, bus to be OPERABLE. The Allowable Value for this trip function is shown as NA because there is no Analytical Limit for RCP Undervoltage. The6"RCPs will continue to operate and deliver required: RCS flow duringan Undervoltage Condition. The reactor'trip on RCP Undervoltage is a time-zero initiating event assumed in the safety analysis (Reference 3). The UV relay is adjusted for a nominal trip setpoint of 75% of the 6900 Vac'bus voltage and the surveillance acceptance criterion used for this function'is Ž 70%.

In MODE 1 above the P-7 setpoint, the Undervoltage RCP trip must be'OPERABLE. Below the P-7 setpoint, all reactor trips on loss of flow are automatically blocked since no conceivable power distributions could occur that would cause a DNB concern at this'low power level. Above the P-7 setpoint, the reactor trip on loss of flow in two or more RCS loops is automatically.

enabled.

12. Underfrequency Reactor Coolant Pumps The Underfrequency RCPs reactor trip Function ensures that protection is provided against violating the DNBR limit due to a loss of flow in.two or more RCS loops from a major network frequency disturbance. An underfrequency condition will slow down the pumps, thereby reducing their coastdown timefollowing a Dumptip.JheoroDercoastdownti.mejs.sreau-red.

-- so-that.-:

reactor heat can be removed immediately after reactor trip.

The frequency of each RCP bus is monitored. A loss of frequency detected on two or more RCP buses trips all four RCPs, a condition that will initiate a reactor trip. This trip Function will generate a reactor trip before the Reactor Coolant Flow-Low (Two Loops) Trip Setpoint is reached.'

The LCO requires one Underfrequency RCP channel per bus to be OPERABLE.

(continued)

INDIAN POINT 3 B 3.3.1-26 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, andAPPLICABILITY. (continued)

In Mode I above the P-7 Setpoint, the Underfrequency RCP's trip must be OPERABLE. Below the P-7 Setpoint, all reactor trips on loss off flow are automatically blocked since no conceivable power distribution, could occur that would cause a DNB. Concern at this low power level. Above the P-7 Setpoint, the reactor trip on loss of flow in two. or more RCS loops is automatically enabled.

13. Steam Generator Water Level-Low Low The SG Water Level-Low Low trip Function ensures that

.protection is provided against a loss of heat sink and actuates the AFW System prior to uncovering the SG tubes. The SGs are the heat sink for the reactor. In order to act as a heat sink, the SGs must contain *aminimum amount of water. A narrow range low low level in any SG is indicative of a loss of. heat sink for the reactor. The "B" channel level transmitters provide input to the SG Level Control System. Therefore, the actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation, and. a singlefailure in the other channels providing the protection function actuation. This Function also performs the ESFAS function of starting the AFW pumps on low low SG level.

The LCO requires three channels of SG Water Level-Low Low per SG to be OPERABLE. Each SG is considered to be a separate fuct*-n.Threfore, separate condition entry is allowed for each SG.

In MODE 1 or 2,. when the reactor requires a heat'sink, the SG Water Level-Low Low trip must be OPERABLE. The normal source of. Water for the SGs is the Main Feedwater (MFW) System (not safety related). The MFW System is only in operation in MODEI or 2. The AFW System is the safety related~backup source of water to ensure that the SGs remain the heat sink for the reactor. During normal startups and shutdowns, the AFW System provides feedwater to maintain SG level. In MODE 3,-4, 5, (continued)

INDIAN POINT 3 .B 3.3.1-27 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) or 6, the SG Water Level-Low Low Function does not have to be OPERABLE because the MFW System is not in operation and the reactor is not critical . Decay heat removal is accomplished by the AFW System in MODE 3 and 4 andtby the Residual. Heat. Removal (RHR) System in MODE 4, 5, or 6.

14. Steam Generator Water Level-Low, Coincident With, Steam Flow/Feedwater Flow Mismatch SG Water Level-Low, in conjunction with the Steam Flow/Feedwater Flow Mismatch, ensures that protection is provided against a loss of heat sink and actuates the AFW-System. In addition to a decreasing water.level in. the SG, the difference between feedwater flow and steam flow is evaluated

. to determine if feedwater flow is significantly less than steam

.flow. With.less feedwater flow than steam flow, SG level*will decrease at a rate dependent upon the magnitude of the difference in flow rates. The required logic is developed from two SG level channels and two Steam Flow/Feedwater Flow Mismatch channels per SG. One narrow range level channel

coincident with the associated Steam Flow/Feedwater Flow Mismatch channel for the same SG (steam flow greater than feed flow) will actuate a reactor trip.

The LCO requires two channels of SG Water Level -Low coincident with Steam Flow/Feedwater Flow Mismatch.

- Each SG is considered to be a separate function._ _Therefore,..

separate conditionentry is allowed for each SG.

Table 3.3.1-1 identifies the Technical Specification Allowable Value for this trip function as not applicable (NA) because LCO 3.3.1, Function 13, Steam Generator Water Level-Low Low, is used todbound the analysis for a loss of feedwater event. The allowable values required for OPERABILITY of Function 13 is Ž 4.0%. The surveillance acceptance criteria used for Function

14 are 7.5% narrow range level and

1.33E+6 pounds per hour steam'flow/feedwater flow mismatch.

(continued)

TNnTAN POTNT - R 1 ' 1-* 9 vicinn S

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES,..LCO, and APPLICABILITY (continued)

In MODE 1 or 2, when the reactor requires a heat sink, the SG Water Level-Low coincident with Steam Flow/Feedwater Flow Mismatch trip must be OPERABLE. The normal source'of watervfor the SGs is the MFW.System (not safety related). The MFW System is only in operation in MODE 1 or 2. The AFW System is the safety related backup source of water to ensure that the SGs remain, the heat sink for the;:reactor. During normal startups and shutdowns, the AFW System provides feedwater-to maintain SG level.. In MODE 3, 4, 5, or 6:, the SG Water. Level -Low coincident withSfeam Flow/Feedwater Flow Mismatch Function does not have to be OPERABLE because the MFW System is not in operation and the reactor is not critical. Decay heat removal is accomplished by the AFW System in MODE 3 and 4 and by.the RHR.System in MODE 4, 5, or 6. The MFW System is in operation only in MODE 1 or 2 and, therefore, this trip Function need, only be OPERABLE.in these MODES.

15. Turbine Trip,- Low Auto-Stop Oil Pressure The Turbine Trip-Low Auto-Stop Oil Pressure trip Function anticipates the loss of heat removal capabilities of the secondary system following a turbine trip. This trip Function acts to minimize the pressure/temperature transient on the reactor. Any turbine trip from a power level below the P-8.

setpoint will not actuate a reactor trip. Three pressure switches monitor the control oil pressure in the Turbine Control System. A low pressure condition sensed.by two-out-of-three pressure switches will actuate a reactor trip..

These pressure switches do not provide any input to the control

- system. The unit is designed to withstand a complete loss of load and not sustain core damage or challenge the RCS pressure limitations. Core protection is provided by the Pressurizer Pressure-High trip Function and RCS integrity is ensured by the pressurizer safety-valves..

The LCO requires three channels of Turbine Trip-Low Auto- Stop Oil Pressure to be.OP.ERABLE in MODE 1 above P-8.

(continued)

INDIAN POINT 3 B 3.3.1-29 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Below the P-8.setpoint, a turbinetrip does not actuate a reactor trip. In MODE.l (below P-8 setpoint), 2, 3, .4, 5, or 6, there is no potential for a turbine trip that would require a reactor trip., and the Turbine Trip- Low Auto-Stop Oil Pressure trip Function does not need to be OPERABLE.

16. Safety Injection Input; from Engineered Safety Feature Actuation System

. The SI Input from ESFAS ensures that-if a reactor trip hasnot already been generated by the RPS, the ESFAS automatic actuation logic will initiate a reactor trip signal upon any signal that initiates SI. Thisis a.condition of acceptability for the LOCA.. However, other transients and accidents.take credit for varying levels of ESF performance and rely upon rod insertion, except for the most reactive rod that is assumed to be fully withdrawn, to ensure reactor shutdown. Therefore, a reactor trip is initiated every time an SI signal*is present.

Trip Setpoint and Allowable Values are not applicable to this Function. The SI Input is provided by relay in the ESFAS.

.Therefore, there is no measurement signal with which to associate an LSSS.

The LCO requires two *trains of SI Inputfrom ESFAS to be OPERABLE in MODE I or 2.

A reactor trip is initiated every time an. SI signal is present.

Therefore, this. trip Function must be OPERABLE in MODE 1.or 2, when the reactor is cri:tical, and must be shut down in the event of an accident. In MODE 3, 4, 5, or 6, the reactor is not critical, and this trip Function does not need to be OPERABLE.

(continued)

INDIAN POINT 3 B 3.3.1-30 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

.17. Reactor Trip System Interlocks Reactor protection interlocks are provided to ensure. reactor trips are in the correct configuration for the current unit status. They back up operator actions to ensure protection system Functions are not bypassed during unit conditions under which thesafety-analysis assumes the Functions are not.

bypassed. Therefore, the interlock Functions do not need to be OPERABLE when the associated reactor trip functions are outside the applicable MODES. These are:

a. Intermediate Range Neutron Flux, P-6 The Intermediate Range. Neutron Flux, P-6 interlock is actuated when.any NIS intermediate range channel goes approximately. one decade above the minimum channel reading. If both channels drop below the setpoint, the permissivewill-automatically be defeated. Manual defeat of the P-6 interlock can be accomplished at any time by simultaneous .actuation of both Reset pushbuttons. The LCO requirement for the.P-6 interlock ensures that the following Functions are performed:

on increasing power, the P-6 interlock allows the manual block of the NIS Source Range, .Neutron Flux reactor trip. This prevents a premature block of.

__ ....... ... the source range trip and allows the operator to ensure that the. intermediate range is OPERABLE prior to leaving the source range. The source range trip-is blocked by removing the high voltage to the detectors;.

".on ,decreasing power, the P-6 interlock automatically energizes the NIS source range .

detectors and enables the NIS Source Range Neutron Flux reactor trip; and The LCO requires two channels of Intermediate Range Neutron Flux, P-6 interlock to be.OPERABLE in MODE 2 when below the P-6 interlock setpoint.

(continued)

INDIAN POINT 3 B 3.3.1-31 Revision 3

RPS Instrumentation B 3.3.1

BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY. (continued)

Above the P-6 interlock setpoint, the NIS Source Range Neutron Flux reactor trip will.be blocked,.and this Function willno longer be necessary.

In, MODE 3, 4, 5, or 6, the P-6 interlock does not have to be OPERABLE because the NIS Source Range is providing core protection if required.

The Allowable'Value is NA for this function because there-is. no corresponding analytical limit modeled in the accident analysis. The surveillance acceptance criterion used for this Function is Ž 3.1E-11 Amps .

b. Low Power Reactor Trips Block, P-7 The Low Power Reactor Trips Block, P-7 interlock, is actuated by input from either the Power Range Neutron Flux, P-1O, or the Turbine First Stage Pressure. The LCO requirement for the P-7 interlock ensures that the

following Functions are performed:

(1), on increasing power, the P-7 interlock (i.e., 2 of 4 Power Range channels increasing above the P-1O (Function 17.d) setpoint or.1 of 2 Turbine First Stage Pressure (Function 17.e) setpoint) automatically enables reactor trips on the

. . . . ... ..- fol-1 owi-ng-Func-t-i ons ..

Pressurizer Pressure- Low;

Pressurizer Water Level-High;

Reactor Coolant Flow-Low (Two Loops);

RCPs Breaker Open (Two Loops);

Undervoltage RCPs; and U

Underfrequency RCPs (continued)

S INDIAN POINT 3 B 3.3.1-32 Revision 3

RPS Instrumentation B 3.3.1

BASES APPLICABLE SAFETY ANALYSES, LCO, andAPPLICABILITY (continued)

These reactor trips are only required when operating above the P-7. setpoint (approximately 10% power). The reactor trips provide protection against violating the DNBR limit. Belowthe P-7 setpoint, the RCS is capable of providing sufficient natural circulation without any RCP running.

(2) on decreasing power, the P-7 interlock (i.e., 3 of

. 4 Power Range channels decreasing below the P-1O (Function 17.d) setpoint and 2 of 2 Turbine First Stage Pressure channels decreasing below the Turbine First Stage Pressure (Functioh 17.e)'

setpoint) automatically blocks reactor trips on the following Functions:

Pressurizer Pressure-Low;

Pressurizer Water Level-High; a Reactor Coolant Flow-Low (Two Loops);'

RCP Breaker Position (Two Loops);

" Undervoltage RCPs; and

. ..... ...... u,,Underfrequency--R .Ps.. . . . .. .

An Allowable Value-is not applicable to the P-7 interlock because it is a logic Function. The P-10 interlock (Function 17.d) governs input from the Power Range..

instruments and the Turbine First Stage Pressure interlock (Function 17..e) governs input for turbine power.

The P-7 interlock is a logic Function with train and not channel identity. Therefore, the LCO requires-one channel per train (i.e., two trains) of Low Power Reactor Trips Block, P-7 interlock to be OPERABLE in MODE 1.

(continued)

INDIAN POINT 3 B 3.3.1-33 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

The low power trips are blocked below the P-7 setpoint and unblocked above the P-7 setpoint. In MODE 2, 3, 4, 5, or 6, this Function does not have to be OPERABLE because the interlock performs its Function when power level drops:,below 10% power, which is in MODE 1.

c. Power Range Neutron Flux, P-8 The PowerRange Neutron Flux, P-8 interlock is actuated below. 50% power as determined by NIS power range detectors. The P-8 interlock automatically enables the Reactor Coolant Flow-Low (Single Loop) and RCP Breaker Position (Single Loop) reactor trips on low flow in one or more RCS loops whenever at least 2 of 4 of the Power Range instruments increase to above the P-8 setpoint.

The LCO requirement for this trip Function ensures that

protection is provided against a loss of flow in any RCS loop that could result in DNB conditions in the core when greater than approximately 50% power. On decreasing power, the reactor trip on low flow in any loop is

automatically blocked whenever at least 3 of 4 the Power Range instruments decrease to below the P-8 setpoint.

The.LCO requires four channels of Power Range Neutron Flux, P-8 interlock to be OPERABLE in MODE 1.

In MODE 1, a loss of flow in one RCS loop could result in

______.__ DNB conditions, so the Power Range Neutron Flux, P-8 interlock must be OPERABLE. In MODE 2, 3, 4, 5, or 6,

-this Function does not have to be OPERABLE because the core is not producing sufficient power to be concerned about DNB conditions.

The Allowable Value is NA for this Function because there is no corresponding analytical limit modeled in the accident analysis. The surveillance acceptance criterion used for this Function is

35% RTP.

(continued)

INDIAN POINT 3 B 3.3.1-34 Revision. 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

d. Power.Range Neutron Flux, P-10 The Power.Range Neutron Flux, P-1O interlock-is actuated at approximately 10% power, as determined by two-out-of-four NIS'power range detectors. If power level falls, below 10% RTP on 3 of 4 channels, the nuclear instrument trips will be automatically unblocked. The LCO requirement for the P-lO interlock ensures that the

following Functions arel.performed:

on increasing power, the P-10 interlock allows the operator to manually block.the Intermediate Range Neutron Flux reactor trip; on -increasing power, the P-10 interlock allows the operator to. manually block the Power Range Neutron Flux-Low reactor trip; on. increasing power, the P-1O interlock automatically provides a backup signal to block the Source Range Neutron Flux reactor trip by de-energizing the NIS source range detectors; the P-10 interlock provides one of the two inputs to the P-7 interlock; and

...-----on-decreasing-power-,-the-P--1O-i-nter-l oGk---.. ....-

automatically enables the Power Range Neutron Flux

-Low reactor trip and the Intermediate Range "Neutron Flux reactor trip (and rod stop).

The LCO requires four channels of Power Range Neutron Flux, P-10 interlock to.be OPERABLE in MODE 1 or 2.

(continued)

INDIAN POINT 3B B 3.3.1-35 Revision 3.

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

OPERABILITY in MODE 1 ensures the Function is available to perform its decreasing power Functions in the event of a reactor shutdown. *This Function must be OPERABLE in MODE 2 to ensure that core protection is provided during a startup or shutdown by the Power Range Neutron Flux-Low and Intermediate Range Neutron Flux reactor trips.

In MODE 3, 4, 5,-or 6, .this Function does not have to be OPERABLE becauseý the reactor is not at power and the Source Range Neutron Flux reactor trip provides core protection. -

The Allowable Value is NA for this Function because there is no corresponding analytical limit modeled in the accident analysis. The surveillance acceptance criterion used for this Function is

9- RTP.

e. Turbine First Stage Pressure The Turbine First Stage Pressure interlock is actuated whenthepressure in the first stage of the high pressure turbine is greater than approximately 101 of the rated full power pressure. This is determined by one-out-of-two pressure detectors. The LCO requirement for this Function ensures that one of the inputs to the P-7 interlock is available.

The LCO requires two channels of Turbine Impulse Pressure, input to the P-7 interlock, to be OPERABLE in MODE 1.

The Turbine First Stage Pressure interlock must be, OPERABLE: when the' turbine generator is operating. The interlock Function is not required OPERABLE in MODE 2, 3, 4, 5, or 6 because the turbine generator is not operating.

The Allowable Value is NA for this Function because there is no corresponding analytical.limit modeled inthe accident analysis. The surveillance acceptance criterion used for this Function is 9.5% RTP.

.(continued)

INDIAN POINT 3 B 3.3.1-36 Revision 3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY. (continued)

18. Reactor Trip Breakers This trip Function applies to the RTBs exclusive of individual trip mechanisms. The LCO requires two OPERABLE trains of trip.

breakers. A trip breaker train consists of all trip breakers associated with a single RPS logic train that are racked in, closed, and capable of supplying power to the Rod Control System. Thus,. the train may. consist of the main breaker, bypass breaker, or main breaker and bypass breaker, depending upon the system configuration. Two OPERABLE trains ensure no single random failure can disable the RPS trip capability.

The LCO requires two OPERABLE trains of trip breakers. Two OPERABLE trains ensure no single random failure can disablethe RPS trip capability. When a reactor trip.breaker is being tested, both reactor trip breaker and the reactor trip bypass breaker associated with the.RPS logic train not in test are closed. In this. configuration, a single failure in the RPS logic train not in test could disable RPS trip capability; therefore, limits on the duration of testing are established.

These trip Functions must be OPERABLE in MODE I or 2 when the reactor is critical. In.MODE 3, 4, or 5, these RPS trip Functions must be OPERABLE when the Rod Control System is capable of rod withdrawal or one or more rods are not fully inserted.

-1T9R- ctor-T.ilBreak-r Ui-diltge a-idS t Trip Mechanisms.

The LCO requires both the Undervoltage and Shunt Trip Mechanisms to be OPERABLE for each RTB that is in service.- The trip mechanisms are not required to be OPERABLE for trip breakers thatare open, racked out, incapable of supplying power to the Rod Control System, or declared inoperable under Function 18 above. OPERABILITY of both trip mechanisms on each breaker ensures that no single trip mechanism failure will prevent opening any breaker on a valid signal.

(continued)

INDIAN POINT 3 B 3.3.1-37 Revision.3

RPS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RPS trip Functions must be OPERABLE when the Rod'Control System is capable of rod withdrawal or one..or more rods are not fully inserted.'.

20. Automatic Trip Logic, The LCO requirement for the RTBs (Functions 18 and 19) and Automatic Trip Logicý (Function'20) ensures that means are providedto interrupt the power to allow the rods to fall into the *reactor core.: Each RTB is equippedwith a bypass breaker (RTBB) to allow testing of the trip breaker while the unit is at power. Each RTB and RTBB is equipped with an undervoltage coil and a shunt trip coil to trip the breaker.open when needed. The reactor trip signals generated by the RPS Automatic Trip Logic cause the RTBs and associated bypass breakers to open and shut down the reactor.

The LCO requires two trains of RPS Automatic Trip Logic to be OPERABLE. *Having two OPERABLE channels ensures that random failure of a single logic channel-will not prevent reactor trip.

These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RPS trip Functions must be OPERABLE when the Rod Control System is

...capable' of rod withdrawal or one or more rods are not fully inserted.

The RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36.

ACTIONS A Note has been added to the ACTIONS to clarify'the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.1-I.

In the event a channel's.Trip Setpoint is found nonconservative with respect to 'the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or bistable is 'found inoperable, then all affected Functions provided by that-channel must be declared inoperable and the LCO Condition(s) entered for the protection Function(s) affected.

(continued)

INDIAN POINT 3 .B 3.3.1-38 Revision 3

RPS Instrumentation B 3.3.1 BASES ACTIONS (continued)

When the number.of inoperable-channels in a trip Function exceed those specified .in one or other related Conditions associated with a trip.Function, then the unit is outside the safety analysis.;

Therefore, LC0I3.0.3 must be immediately entered if applicable in the current MODE of operation.

A.1 Condition.-A appl.ies to all RPS protection Functions. Condition A addresses the situation where one or more required channels or trains for one or more Functions are inoperable at the same time.

The Required Action is to refer to Table 3.3.1-1 and to take the Required Actions for the protection functions affected. The Completion Times are those from the referenced Conditions and Required-Actions'.

B.1 and B.2 Condition B applies to'theManual Reactor Trip in MODE 1 or 2. This action addresses the train orientation of the relay logic for this Function. With one channel inoperable, the inoperable channel must be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . In this Condition, the remaining OPERABLE channel is adequate to perform the safety function.

The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is.reasonable considering that there are two automatic actuation 'trains and another manual initiation channel OPERABLE, and the low probability of an event occurring during this interval.

If the Manual Reactor Trip Function cannot be restored to OPERABLE status within the all'owed'48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time, the unit must be brought to a MODE. inwhich the requirement does not apply. To

-.achieve this status, the unit must be brought to at least MODE 3 within 6 additional hours (54 hour6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months s-total time). The 6 additional hours to reach MODE 3 is reasonable, based on operating experience, to. reach MODE 3 from full power operation in an orderly manner and without challenging unit systems. With the unit in MODE 3, ACTION C.

applies to any inoperable Manual Reactor Trip, Function if the Rod Control System is capable of rod withdrawal or one or more rods are not fully inserted.

(continued)

INDIAN POINT 3 B 3.3.1-39 Revision a

RPS Instrumentation B 3.3.1 BASES ACTIONS (continued)

C.1 and C.2 Condition C applies to the following reactortrip Functions in MODE 3, 4, or 5 when the Rod Control System capable of rod withdrawal or one or more rods are not fully inserted:

Manual Reactor Trip; RTBs; RTB Undervoltage and Shunt Trip Mechanisms; and

Automatic Trip Logic.

This action addresses the train orientation of the relay logic for these Functions. With one channel or-train inoperable, the inoperable channel or train must be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . If the affected Function(s) cannot be restored to OPERABLE status within the allowed 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Completion Time, the unit must be placed in a MODE in which the requirement does not apply.

To achieve this status, action must be initiated within-the same 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to ensure that all rods are fully inserted, and the Rod Control System must be placed in a condition incapable of rod withdrawal within the next hour. The additional hour provides sufficient time to accomplish the action in anlorderly manner. *With rods fully inserted and the Rod Control System incapable of rod withdrawal, these Functions are no longer required.

The Completion Time is reasonable considering that in this Condition, the remaining OPERABLE train is adequate to perform the safety *function, and given the low probability of an event occurring during this interval.

D.1 and D.2 Condition D applies to the Power Range Neutron Flux-High.Function.

The NIS power range detectors provide-input to the Rod Control System and, therefore, have a two-out-of-four trip logic. A known inoperable channel must be placed in the tripped condition. This results in a partial trip condition requiring only one-out-of-three (continued)

INDIAN POINT 3 B 3.,3. 1 -40 Revision 3,

RPS Instrumentation B 3.3.1 BASES ACTIONS D.1 and D.2.(continued)

- logic for actuation. The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowedito place the inoperable channel in the tripped condition is justified in WCAP-10271-P-A

..(Ref.. 7).

The 6-hour Completion Time is consistent with LCO 3.2.4, "QUADRANT POWER TILT RATIO .(QPTR)."

As an alternative to the above Actions, the plant must be placed in a MODE where this Function is no longer required OPERABLE. Twelve hours are allowed to place theplant in MODE 3. This is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems. If Required Actions cannot be completed within their.

allowed Completion Times, LCO 3.0.3,must be entered.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypass condition for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months while performing routine surveillance testing of other channels. The Note also allows placing the inoperable channel in the bypass condition to allow setpoint adjustments of other channels when required to reduce the setpoint in accordance with other Technical Specifications.

E.1 and E.2 Condition E applies to the following reactor trip Functions:

Power Range Neutron Flux-Low; Overtemperature AT; Overpower AT; Pressurizer Pressure-High;

.

SG Water Level -Low Low; and SG Water Level -Low.coincident with Steam Flow/Feedwater Flow Mismatch.

(continued)

INDIAN POINT 3 B 3.3.1--41 Revision 3

RPS Instrumentation B 3.3.1 BASES ACTIONS E.1 and E.2 (continued)

A known inoperable channel must be placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Placing the channel in the.tripped condition results in a partial trip condition requiring only one-out-of-two logic for actuation of the two-out-of-three trips and one-out-of-three logic'for actuation of the two-out-of-four trips.

The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to place the inoperable channelin the tripped condition is-justified inhReference7.

If the operable channel cannot be placed in the trip condition within the specified Completion Time, the unit must be placed in a MODE where these Functions are not required OPERABLE. An additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed to place the unit in MODE 3. Six hours is-a.a reasonable time, based on operating experience, to place the unit in MODE 3 from full power in an orderly manner and without challenging unit systems.

The Required-Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months while performing routine surveillance testing of the other

channels.

F.1 and F.2 Condition F applies when there are no Intermediate Range Neutron, Flux trip channels OPERABLE in MODE 2 when THERMAL POWER is above the P-6 setpoint and below the P-1O setpoint. Required Actions -

specified in-this Condition are only applicable when channel

.. fi.lr...d.not result in reactor .trip. Above the P-6 setpoint and below the P-10 setpoint, the NIS intermediate range detector%

performs the monitoring Functions. With no intermediate range .

channels OPERABLE, .the Required Actions are to suspend operations involving positive reactivity additions immediately. This will preclude any power level increase since there are no OPERABLE .

Intermediate Range Neutron Flux channels.- The operator must also reduce THERMAL POWER below the. P-6 setpoint within two hours. Below P-6, one or both Source Range Neutron Flux channels will be able to-monitor the core power level. The Completion-Time of.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months will .

allow a slow and controlled power reduction to less than the P-6 setpoint and takes into accountthe low probability of occurrence of an event during this period that may require the protection afforded by the NIS Intermediate Range Neutron Flux trip.- " -

(continued)

INDIAN POINT 3 B 3.3.1-42 Revision 3

RPS Instrumentation B 3.3.1 BASES ACTIONS (continued)

G.1 Condition G applies when there are no Source Range Neutron Flux trip channels.OPERABLE when in-MODE 2, below the P-6 setpoint, and in MODE 3,.4, or,.5 with the Rod Control capable of rod withdrawal or one or more rods not rods fully inserted. With the unit in this Condition, below P-6, the NIS source range performs the monitoring and protection functions. With both source range channels noperable, the RTBs must be opened, immediately. With-the RTB's open, the-core is in a more stable condition.

H.1 and H.2 Condition H applies to the following reactor trip Functions:

Pressurizer Pressure-Low;

-Pressurizer Water. Level- High;

Reactor Coolant Flow-Low; RCP Breaker Position (Two Loops);

Undervoltage RCPs; and

Underfrequency RCPs.

.. ........ -W.i-th- one-ehannel--inoperable-- the AinoperabIe--channeV-must-beý-placed-in the tripped condition within-6 hours. -Placing the channel in the:

tripped condition results in a partial trip condition requiring only one additional channel' to initiate a reactor trip above the P-7 .

setpoint for the two loop function and above the P-8 setpointfor

.the single loop'function.

These Functions do not have to be OPERABLE below the P-7 setpoint

because there are no loss of flow trips'below the P-7 setpoint. The 6 :hours allowed to place the channel'in the tripped condition is justified in Reference 7.. An additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed to reduce THERMAL POWER to below P-7 if the inoperable channel cannot

- (continued)

INDIAN POINT 3 1B 3.3.1-43 Revi sion 3

RPS Instrumentation B 3.3.1 BASES ACTIONS (continued)

H.1 and H.2 (continued) be restored to OPERABLE'status or placed in trip within the specified Completion Time. The Reactor Coolant Flow-Low (Single Loop) reactor trip does not have to be OPERABLE below the P-8 setpoint; however, the Required Action must take the plant below the P-7 setpoint if the inoperable channel is not tripped within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months because of the shared components between this function and the Reactor Coolant Flow-Low (Two.Loop) reactor trip function.

Allowance of this time interval takes into consideration the redundant capability provided by the remaining redundant OPERABLE channel, and the low probability of occurrenrceof an event during this period that may require the protection afforded by the

.Functions associated with Condition H.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months while performing routine surveillance testing of the other channels.

1.1 and 1.2 Condition I applies to the RCP Breaker Position (Single Loop) reactor trip Function. There is one breaker position device per RCP breaker. With one channel inoperable, the inoperable channel must be restored to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If the channel cannot be restored to OPERABLE status within the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , then THERMAL, POWER must be reduced below the P-8 setpoint within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .-

This places the unit in a MODE where the LCO is no longer applicable. This Function doesnot have to be OPERABLE below the P-8 setpoint because other RPS Functions provide core protection below the. P-8 setpoint. The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to restore the channel to OPERABLE status and the 4 additional hours allowed to reduce THERMAL POWER to below the P-8 setpoint are justified in Reference 7.

(continued)

INDIAN POINT 3 B 3.3.1-44 Revision 3

RPS Instrumentation B 3.3.1 BASES ACTIONS I.1 and 1.2. (continued)

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months while performing routine, surveillance testing of the other channels.

J.1 and J.2 Condition J appliesto Turbine Trip on Low Auto-Stop Oil Pressure.

With one channel. inoperable, the inoperable channel, must be placed in the trip condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If placed in the tripped condition, this results in a: partial trip condition requiring only

-one additional channel to initiate a reactor trip. If the channel cannot be restored to OPERABLE status or placed in the trip

-condition, then power must be reduced below the P-8 setpoint within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed to place the inoperable.

channel in the tripped condition and the 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months allowed for reducing power are justified in Reference 7.

The Required Actions have:been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months while performing routine surveillance testing of the other channels.

K.1 and K.2 Condition K applies to the SI Input from ESFAS reactor trip and the RPS Automatic. Trip Logic in MODES 1 and 2. These actions address the train orientation of the RPS for these Functions. With one

- train inoperable, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months are allowed to restore the train to OPERABLE status (Required Action K.1) or theunit must be placed in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Required Action K.1) is reasonable considering that in this

,Condition, the remaining OPERABLE train is adequate to perform the safety function .and~given the low probability of an event during this interval. The Completion Time'of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (Required Action K.2) is reasonable, based:on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging unit systems.

The Required Actions have been modified by a Note that allows bypassing one train up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for surveillance testing, provided the other train is OPERABLE.

(continued)

INDIAN POINT 3 B 3.3.1-45 Revision 3

RPS Instrumentation B 3.3.1 BASES ACTIONS (continued)

L.1 and L.2 Condition L applies to the RTBs in MODES 1 and 2. These actions address the train orientation of the RPS for the RTBs. With one train inoperable, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to restore the train to OPERABLE status or the unit must be placed in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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 in an orderly manner and without challenging unit systems. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Times are equal to the time allowed by LCO 3.0.3 for shutdown actions in the event of a complete loss of RPS Function. Placing the unit inMODE 3 results in ACTION C entry while RTB(s) are inoperable.

The Required Actions have been modified by two Notes. Note 1 allows one channel to be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance.

testing, provided the other channel is OPERABLE. Note 2 allows one RTB to be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for maintenance on undervoltage or shunt trip mechanisms if the other RTB train is OPERABLE. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time'limit is justified in Reference 7.

As noted in Reference 9, the allowance of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for test and maintenance of reactor trip breakers provided in Condition L, Note 1, is less than the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowable out of service time and the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allowance for testing of RPS train A and train B. In practice, if the reactor trip breaker is being tested at the same time as the associated logic train, the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allowance for testing of RPS train A and train B applies to both the logic train and the reactor trip breaker. This is acceptable based on the Safety Evaluation Report for Reference 7.

M.1 and M.2' Condition M applies to theTP46 and P-10 interlocks. With one or more channels inoperable for one-out-of-two or two-out-of-four.

coincidence logic, the associated interlock must be verified to be in its required state for the existing unit condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or the unit must be placedin MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Verifying the interlock status manually accomplishes the interlock's Function. The Completion' Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is based on operating experience and the minimum amount of time allowed for manual (continued)

INDIAN POINT 3 B 3.3.1-46 Revision 3

RPS Instrumentation B 3.3.1 BASES ACTIONS M.1 and M.2 (continued) operator actions. 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 in an orderly manner and without challenging unit systems. The I hour and 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion Times are equal to the time allowed by LCO 3.0.3 for shutdown actions in the event of a complete loss of RPS Function.

N.1 and N.2 Condition ,N applies to the P-7 and P-8 interlocks and the turbine first stage pressure'input to P-7. With one or more channels inoperable for one-out-of-two or two-out-of-four coincidence logic, the associated interlock must be. verified to be in its required state for the existing unit condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or the unit must be placed in MODE 2 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These actions are conservative for the case where power level is being raised.

Verifying the interlock status manually accomplishes the interlock's Function. The Completion Time of: 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is based on operating experience and the minimum amount of time allowed for manual operator actions: 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 2 from full power in an orderly manner and without challenging unit systems.

0.1 and 0.2 Condition 0 applies to the RTB Undervoltage and Shunt Trip Mechanisms, or diverse trip features, in MODES 1 and2.1 With one of the. diverse trip features inoperable, it must be restored to an OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or the unit must be placed in a MODE where the requirement does.not apply, ,This is accomplished.by placing the unit in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months total time).. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging unit systems.

'.ith the unit in MODE 3, ACTION C applies to any inoperable RTB trip mechanism. The affected RTB shall not be bypassed while one of the diverse features is inoperable except for the time required to perform maintenance to one of the diverse features. The allowable time.for performing maintenance of the diverse features is2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months forýthe reasons stated under Condition L.

(continued)

INDIAN POINT 3 .B 3.3.1-47 Revision 3-.

RPS Instrumentation B 3.3.1 BASES ACTIONS 0.1 and 0.2 (continued)

The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for Required Action 0.1 is reasonable considering that in this Condition there is one remaining diverse feature for the affected RTB, and one OPERABLE RTB capable of performing the safety function and given the low probability of an event occurring during this interval.

SURVEILLANCE REQUIREMENTS The SRs for each RPS Function are identified by the SRs column of Table 3.3.1-1 for that Function..

A Note has been added to the SR Table stating that Table 3.3.1-1 determines which SRs apply to which RPS Functions.

Note that each channel of process protection supplies both train A and train B of the RPS. When testing an individual channel, the SR is not met until both train A and train B logic are tested. The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similarparameter.on other channels. It is based on the.

assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff based on a combination-of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

(continued)

INDIAN POINT 3 B 3.3.1-48 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.1 continued)

The Frequency is based on operating experience that demonstrates channel failure is.rare. The CHANNEL CHECK supplements less formal checks of channels during normal 'operational use of the displays associated with the.,LCO.required channels.

SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the calorimetric exceeds.'the.

NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted. If the NIS channel output cannot be properly adjusted, the channel is declared inoperable.

Two Notes modify SR 3.3.1.2., The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute'difference between the NIS channel output and-the calorimetric :is > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is Ž 15% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.

The Frequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate.. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Together these...___

factors demonstrate the change in the absolute difference between NIS and heat balance-calculated powers rarely exceeds 2% in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

In addition, control:room operators periodically monitor redundant "indications and alarms to detect deviations in channel outputs.

SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output every 31 EFPD. If the absolute difference is Ž 3%, the NIS channel is ýstill OPERABLE, but must be readjusted.

(continued)

INDIAN POINT 3 B 3.3.1-49 Revision .3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.3 (continued)

If the NIS channel cannot be properly readjusted, the channel is declared inoperable. This Surveillance is performed to verify the f( I) input to the overtemperature T Function.

Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is Ž3%. SR 3.3.1.3 is performed to ensure that the AFD input to the Overtemperature Delta T and the system used to monitor LCO 3.2.3, AFD, are within acceptable limits. The limiting AFD is established to provide the required margin when operating at the highest power level. As power level decreases, the thermal limit becomes less sensitive to AFD because the overall margin to the'thermal limit increases. Note 2 clarifies that the Surveillance is required only if reactor power is ý 90% because the requirements of LCO 3.2.3, Axial Flux Difference (AFD),. are relaxed significantly below 90% RTP.

The Frequency of every 31 EFPD is adequate. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Also, the slow changes in neutron flux during the fuel cyclecan be detected during this interval.

. .... .SR -- 3.3.1.4 1... --

SR 3.3.1.4 is the performance. of a TADOT every 31 days on a STAGGERED TEST BASIS. This test shall verify OPERABILITY by actuation of the end devices.

The RTB test shall include separate verification of thelundervoltage and shunt trip mechanisms. Independent verification of RTB undervoltage and shunt trip' Function is not required for the bypass breakers. No capability is provided for performing such a test at power. The independent test of'the undervoltage and shunt trip function for bypass breakers is included in SR '3.3.1.14. The bypass (continued)

INDIAN POINT 3 B 3.3.1-50 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.4 (continued) breaker test shall include a local shunt trip. A Note, has been added to indicate that this test must be performed on the bypass breaker prior to placing it in~service.

The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering instrument reliability, and operating history data.

SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The RPS

-relay logic is tested every 31 days on a STAGGERED TEST BASIS. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. All possible logic combinations, with and.without applicable permissives, are tested for each protection function required by Table 3.31-1. The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering instrument reliability and operating. history data.

SR 3.3.1.6 SR 3.3.1.6 is a calibration of the excore channels to the incore channels. If the measurements do not agree, the excore channels are

. t-_dl-ared

. i i6o-preabl e'bu-must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared 'inoperable. This Surveillance is performed to verifythe f(AI) input to the overtemperature AT Function.

A Note modifies SR 3.3.1.6. The Note states that this Surveillance

'is required only if reactor power is > 90% because the requirements.

of LCO 3.2.3, Axial Flux Difference (AFD), are relaxed significantly below .90%. RTP. SR 3.3.1.6 is performed to ensure that the AFD input to the Overtemperature Delta T and the, system used to monitor LCO (continued)

INDIAN POINT 3B B 3.3.1-51 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.6 (continued) 3.2.3 AFD are within acceptable limits. The limiting AFD is established to provide theirequired margin when operating at the highest power level. As power level decreases, the thermal limi.t becomes less sensitive to AFD because the overall margin to the thermal limit increases.

The Frequency of 92 EFPD is adequate based on operating experience, considering instrument reliability and operating history data for instrument drift.

SR 3.3.1..7 SR 3.3.1.7 is the performance of a COT every 92 days.

A COT is performed on each required channel to ensure the entire channel will perform the intended Function.

Setpoints must be within-the Allowable Values specified in Table 3.3.1-1.

The "as found" and "as left" values must also be recorded and reviewed. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift

.. ..... allowance used in the setpoint methodology. The setp oint.shall.be left set consistent with the assumptions of Reference 6 which incorporates. the requirements of Reference 7.

SR 3.3.1.7 is modified by a Note that provides a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months delay in the

requirement to perform this Surveillance for source range

instrumentation when entering-MODE 3 from MODE.2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in MODE 3 until the RTBs are open and SR 3.3.1.7 is no longer required to be performed. If theunitis to be in MODE 3 with the RTBs closed for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months this Surveillance must be performed prior to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months deferral is needed because the testing required by SR 3.3.1.7 and SR (continued)

INDIAN POINT 3 B 3.3.1-52 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.7 (continued) 3.3.1.8 cannot be performed on the Source Range, Intermediate Range and Power Range Instruments until in the Applicable. Mode and the proximity of these instruments prevents working on more than one instrument at any onetime.

The Frequency of 92 days is justified in Reference 7.

SR 3.3.1.8:

SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, except it is modified by a Note that this test shall include verification that the P-6 and P-10 interlocks are in their required state for the existing unit condition. The Frequency is.modified by a Note that allows this'surveillance to be satisfied if it has been performed within 92 days of the Frequencies prior to reactor startup and 12'hours after reducing power below P-IO and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing~power below P-6".

The Frequency of "prior to startup" and the note on surveillance intervals ensures this surveillance is performed prior to critical operations or within the prior 92 days and applies to the source, intermediate and power range low instrument channels. :The Frequency of "12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-IO" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after

.reducing power below P-6" (a4ppicable to source range channels) allows a normal shutdown to be completed- and the unit removed from the MODE of Applicability for this' surveillance without a delay to perform the testing required by this surveillance.

The Frequency of every 92 days thereafter applies if the plant remains in the MODE ofApplicability 'after the initial performances "of prior to reactor~startup.. Additionally, this SR must be completed for theintermediate and power range low channels within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below-the P-1O setpoint.and must be' completed for the source range low channel within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below the'P-6 setpoint'. The MODE of Applicability for this -

(continued)

INDIAN POINT 3 B 3.3.1ý-53 IRevi~sion 3

RPS Instrumentation

B 3.3.1 4

BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.8 (continued) surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels., Once the unit is in MODE 3, this surveillance is no longer required. If power is to bemaintained < P-I0 for. more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or < P-6 for more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , then the testing required by this surveillance must be performed prior to the expiration of the time limit unless removed from~the mode of applicability. The specified Frequency

.provides a reasonable time to complete the required testing or place the unit inma MODE where this surveillance is no longer required.

This test ensures that the. NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and within a reasonable time after reducing power into theapplicable.

MODE (< P-1O or< P-6), based on startup testing or testing within the prior. 92 days (e.g., during the shutdown). The deferral of the requirement to perform this test until 12 and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering the Applicable condition is needed because the testing required by SR 3.3.1.7 and SR 3.3.1.8 cannot be performed on the Source Range,

Intermediate Range, and Power Range instruments until in the Applicable Mode and the proximity of these instruments prevents working on more than one instrument at any one time.

SR 3.3.1.9 SR 3.3.1.9 is the performance of a'TADOT and is performed every 92 days, as justified in Reference 7.

The SR is modified by a Note that excludes verification of setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint Verification requires elaborate.

bench calibration and is accomplished during theCHANNEL CALIBRATION.

(continued)

INDIAN POINT 3 B 3.3.1-54 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.10 A CHANNEL CALIBRATION is performed at every refueling and every 18 months for function 11.. CHANNEL CALIBRATION is a complete check of the* instrument loop, including the sensor. The test verifies that.

the channel responds to a measured parameter within the necessary range and accuracy. .

CHANNEL CALIBRATIONS must be performed consistent with the assumptions used in Reference 6. The difference between the current as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.

The Frequency is based on the calibration interval used for the determination of the magnitude of equipment drift in the setpoint methodology.

SR 3.3.1.10 is modified by a Note stating that this test shall include verification that the time constants are adjusted to the prescribed values where applicable.

SR 3.3.1.11 SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10, every 24 months. This.SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL

.-7:..'-"--.CALIBRATIONY-Th*STs ine d because the CHANNELCALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric and flux map performed above 15% RTP. The CHANNEL CALIBRATION for, the source range and

.intermediate range neutron detectors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data.

(continued)

INDIAN POINT 33 B 3.3.1-55 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS SR 3.3.1.11 (continued)

This Surveillance is not required for the NIS-power range detectors for entry into MODE 2 or,1, and is not required for the NIS intermediate:*range detectors for entry into MODE 2, because the unit must be in atleast MODE 2 to perform'the test for the intermediate range detectors and MODE I for the power range detectors. The.

24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown:

these components usually pass the Surveillance when performed on the 24 month Frequency.

SR 3.3.1.12 SR 3.3.1.12 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10, every 24 months. This SR is modified by a Note stating that this test shall include verification of the rate lag compensation for flow from the core to the RTDs. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of resistance temperature detectors (RTD) sensors, which may consist of an.inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel, is accomplished byan-inp-lace--crossca.li.bration. that,.-compares the other-sensing elements with the recently installed element.

-The Frequency is justified 'by the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.1.13 SR 3.3.1.13:is the performance of a COT of RPS interlocks every 24 months.

The Frequency is based on the known reliability of the interlocks and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

(continued)

INDIAN POINT 3 B 3.3.1-56 Revision 3

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.1.14 SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip, RCP Breaker Position, Turbine-Trip, and the SI Input from ESFAS. This TADOT is performed every 24 months. The test'shal1 independently verify~the.OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function for the Reactor Trip Breakers and Reactor Trip Bypass Breakers. The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip.

The Frequency is based on the known-reliability of the Functions and the multichannel redundancy available, and has been shown to be acceptable through operating experience. The SR is modified by a Note that excludes verification of setpoints from the TADOT. The

'Functions affected have no setpoints associated with them.

REFERENCES 1. FSAR, Chapter 7.

2. FSAR, Chapter 6.

3. FSAR, Chapter 14.

4. IEEE-279-1968

5. 10 CFR 50.49.

6. Engineering Standards Manual IES,3 and IES-3B, Instrument Loop Accuracy and Setpoint Calculation Methodology (IP3).

7. WCAP-10271-P-A,.Supplement 2, Rev. 1, June 1990.

8. Consolidated Edison Company of New York, Inc. Indian Point Nuclear Generating Station Unit No. 3 Plant Manual Volume VI:

Precautions, Limitations, and Setpoints, March 1975.

9. WCAP-14384, Implementation of RPS Technical Specification Relaxation Programs, Rev. 0, January 1996.

INDIAN POINT 3 B 3.3.1-57 Revision 3

AC Sources.- Shutdown B 3.8.2 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.2 AC Sources-Shutdown BASES BACKGROUND A description of the AC sources is provided in the Bases for LCO 3.8.1, "AC Sources-Operating."

APPLICABLE SAFETY ANALYSES The OPERABILITY of the minimum AC'sources during MODES 5 and 6 and

during movement of irradiated fuel assemblies ensures that:

a. The unit can be maintained in.the shutdown or refueling condition for extended periods;

b. Sufficient instrumentation and control capability is available for monitoring and maintaining the unit status; and

c. Adequate AC. electrical power is provided to mitigate events postulated during shutdown, such as a fuel handling accident..

In general, when the unit is shut down, the Technical Specifications requirements ensure that the unit has the capability to mitigate the consequences of postulated accidents. However, assuming a single failure and concurrent loss of all offsite or all onsite power is

- not--requiu-re1-w-- The--ra-t-i-onai-l-e-4or -t-his::i-s,-ýbased_-on._ t-he7-.acthamn--

Design Basis Accidents (DBAs) that are-analyzed in MODES1, 2, 3, and 4 have no specific analyses in MODES 5 and 6. Worst case bounding eVents are deemed not credible in MODES 5 and 6 because the energy contained within the reactor pressure boundary, reactor coolant temperature and pressure, and the corresponding stresses result in the probabilities of occurrence being significantly reduced or eliminated, and in minimal consequences. These deviations from DBA analysis assumptions and design requirements (continued)

INDIAN POINT 3 B 3. 8. 2 - I Revision I

AC Sources- Shutdown B 3.8.2 BASES APPLICABLE SAFETY ANALYSES (continued) during shutdown conditions are.allowed by the:LCO for required systems. During MODES 1, 2, 3, and 4, various deviations from the analysis assumptions and design requirements are allowed within. the Required Actions. This. allowance is in recognition that certain testing and maintenance activities must be conducted provided an acceptable level of risk..is not exceeded. During MODES 5 and 6, performance~of: a significant ,number ofrequired testing and maintenance activities is also required. In MODES 5.and 6, the.

activities are generally planned and-administratively controlled.

Relaxations from MODE 1, 2, 3, and.4 LCO.requirements are acceptable during shutdown modes based on:

a. The fact that time in an outage is limited. This is a risk, prudent goal as well as a utility economic consideration.

b. Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical p design requirements applied to-systems credited in operating MODE analyses, or both.

c. Prudent utility consideration of the risk associated with multiple activities that could affect multiplesystems.

d. Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODE 1, 2, 3, and 4 OPERABILITVYrequt-rements.)witthsystems-assumed...to.Junction-d~uinganevent.

In the event of an accident during shutdown, this LCO ensures the capability to support systems necessary to avoid immediate difficulty, assuming either a loss,.of all offsite power or aloss of all onsite diesel generator (DG) power.

The AC sources'satisfy Criterion.3 of 10 CFR 50.36.

(continued)

INDIAN POINT 3 B 3.8.2 - 2 Revision 1

AC Sources - Shutdown B 3.8.2 BASES LCO One offsite circuit capable of supplying the onsite power distribution subsystem(s) of LCO 3.8.10, "Distribution Systems-Shutdown," ensures that all required loads are powered from offsite power. Two OPERABLE DGs, associated with the distribution system train required to be OPERABLE by. LCO 3.8.10, ensures a diverse power source is available to provide electrical power support, assuming a loss of the offsite circuit.. Together, OPERABILITY of the required offsite circuit and, DGs ensures 'the availability of sufficient AC sources to operate the unit in a safe manner and to mitigate. the consequences of' postulated events during shutdown (e.g., fuel handling accidents).- Under specific plant conditions the number of required operable DGs may be reduced to one. The plant conditions described .by the LCO. ensures that ample time is available for operator actions in response to a loss of offsite power. When one residual heat removal (RHR) loop is required to be OPERABLE and in operation, and one RHR loop is required to be OPERABLE,- the RHR loop:

that is OPERABLE but not operating needs to be' capable. of being powered in the event the operating RHR loop fails. However, only the operating RHR loop needs to-'be> capable of being powered from an onsite AC electrical power distribution subsystem associated with an OPERABLE diesel.

The offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while.

connected to the Engineered Safety Feature (ESF)'bus(es). Offsite circuits are those that 'are described in the Bases of LCO 3.8.1, AC Sources - Operating, except that safeguards power trains may be cross connected when in MODES 5 and 6.

-.The DGs'must be capable of starting, accelerating to rated speed'and otgidnh tg'tterrespec tive' ESF bus on detection of bus undervoltage. This sequence must be accomplished within 10 seconds. The DG must be capable of accepting required loads within the assumed.loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses.

Proper sequencing of loads, including tripping of nonessential loads, is a required function.for.DG OPERABILITY.

It is acceptable for safeguards power trains to be cross tied during shutdown conditions, allowing a single offsite power circuit to supply all required trains. In this case, interlocks that disconnect the affected tie breakers before DGs are automatically connected to 'the bus must be OPERABLE.

(continued)

INDIAN POINT 3 B 3.8.2 - 3 Revision. 1

AC Sources - Shutdown B 3.8.2 BASES APPLICABILITY The AC sources required to be OPERABLE in MODES 5 and 6 and during movement of irradiated fuel assemblies provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core;

b. Systems 'needed to'mitigate-a fuel handling accident are available;'

c. Systems necessary to mitigate the effects of events that can lead to'core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit iný a 'cold shutdown condition or refueling condition.

The AC power requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.1.

ACTIONS A.1 An offsite circuit would be considered inoperable if it were not available to one required safeguards power train. Although two safeguards power trains may be required'by LCO 3.8.10, the one train with offsite power available may be capable of supporting sufficient required features to allow.continuation of CORE ALTERATIONS and fuel movement. By theallowance of the option to declare required

..... ....... f ea~tures=inoperabl*e~wi~hs.no-of~fs~terrpower-=ava~i:ab-lerappropr~iate*-=-

... . ... . ........... . .. . * ., . . , r o. a restrictions will be implemented in accordance with the affected required features LCO's.ACTIONS.

A.2.1, A.2.2, A.2.3 and A.2.4 With the offsite circuit not available to all required trains, the option would still exist to declare all required features.

inoperable. Since this option may involve undesired administrative efforts, the allowance for sufficiently (continued)

INDIAN POINT 3 B 3.8.2 - 4 Revision 1

AC Sources - Shutdown B'3.8.2 BASES ACTIONS A.2.1, A.2.2, A.2.3 and A.2.4 (continued) conservative actions is made. It is, therefore, required to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies, and operations involvingpositive reactivity additions. The Required Action to suspend positive reactivity additions does not preclude actions to maintain or increase reactor vessel inventory provided the required SDM is maintained.

Suspension of these activities doesnot preclude completion of actions to establish a safe conservative condition. These actions minimize the probability or the occurrence of postulated events. It is further required to immediately initiate action to restore the required AC sources and to continue this action until restoration is accomplished in order to provide the necessary AC power to the unit safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the unit safety systems may be without sufficient power.

Pursuant to LCO 3.0.6, the Distribution System's ACTIONS would not be entered even if all AC sources to it are inoperable, resulting in:

de-energization. 'Therefore, the Required Actions of Condition-A are modified by a.Note to indicate that when Condition A is entered with no AC power to any required ESF bus, the ACTIONS for LCO 3.8.10 must

. .be immediately e ntered Note all.ws .Condition.A to..p.r.vide requirements for the loss.of the offs'ite circuit, whether or not a train is de-energized. LCO 3.8.10 would provide the appropriate restrictions for the situation involving a de-energized bus..

(continued)

INDIAN POINT 3 B 3.8.2 - 5 Revision 1

AC Sources - Shutdown B 3.8.2 BASES ACTIONS B.1, B.2, B.3 and B.4 (continued)

Condition B is entered when any required DGS are inoperable. A DG would be considered inoperable if it could not support its associated safeguards power train. When LCO 3.8.2.b.1 applies, 2 DGs are required to be OPERABLE:. In this case, whether one or both of the required DGs is inoperable, the minimum required diversity of AC power sources is not available to. required features. Therefore, it is required to suspend CORE ALTERATIONS, movement of irradiated fuel assemblies, and operations involving positive reactive additions.

When specific limitations are satisfied, as stated in LCO 3.8.2.b.2, only one DG is required. The additional restrictions on plant conditions for requiring only one DG provides ample time for operator action, in the event of a loss of offsite power, to manually restore decay heat removal capability. The combination of.

subcritical duration, fuel location, and refueling cavity water level results in a time period of at least 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months for heatup of this water inventory from 140 OF to 200 OF.

With any required DGs inoperable, the Required Action to suspend positive reactivity additions does not preclude actions to maintain or increase reactor vessel inventory provided the required SDM is maintained. Additionally, Required Actions B.1, B.2, and B.3 do not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability or the occurrence

...... _pofpos-tu-ated-e-ents..- -7.777 Furthermore, when Required Actions B.1, B.2 and B.3 are implemented, it is required to immediately initiate action (B.4) to restore the required DG(s) and to continue this action until restoration is accomplished in order to provide the necessary AC power to the unit safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required AC electrical power sources should be completed as quickly as possible in order to minimize the time during which the unit safety systems may be .without sufficient power.

(continued)

INDIAN POINT 3 B 3.8.2 - 6 Revision 1

AC Sources.- Shutdown B 3.8.2 BASES SURVEILLANCE REQUIREMENTS SR 3.8.2.1 SR:3.8.2.1 requires the SRs from LCO.3.8.1 that are necessary for

.. ensuring the OPERABILITY of the AC sources in other than.MODES 1, 2,'

. A3, and 4. SR 3.8.1.8 isnot required to be met since only one offsite circuit is required to be OPERABLE.. SR.3.8.1.9 is not required.to be met because the DG.*automatic trips, are bypassed only on the safety injection start signal, not. on the loss of power..start

signal ... SR'3.8..1.13 is excepted because starting independence is.

not-required.with the DG(s) that.is'not required to be operable.

This SR is modified by two Notes. The reason for the' first Note is to-preclude requiring the OPERABLE DG(s) from being paralleled with the offsite power network or otherwise rendered inoperable during performance of SRs, and to preclude deenergizing a required 480 V ESF bus or disconnecting a required offsite circuit during performance of SRs. With limited AC sources available, a single event could compromise both the required circuit and the DG. It is the intent that these SRs must still be capable of being met, but actual performance is not required during periods when the DG and offsite circuit is required to be OPERABLE.. Refer to the corresponding Bases for LCO 3.8.1 for a discussion of each SR.

The reason for the second Note is that SR 3.8.1.12 includes testing.

with an actual or simulated ESF actuation signal. ESF actuation is.

not required in MODES 5 and 6 so that this portion of the surveillance is not required to be met.

REFERENCES' None.

INDIAN POINT 3 B 3.8.2 - 7 Revi sion.1

ML072910292

Text

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