Text

Technical Specification Bases
	ML043150420
Person / Time
Site:	Indian Point
Issue date:	10/23/2004
From:	; Entergy Nuclear Indian Point 3
To:	; Office of Nuclear Reactor Regulation
References
	IP-SNM-AD-103, Rev 0
	Download: ML043150420 (73)
	v • d • e

23-OCT-04 Page: 87 DISTRIBUTION CONTROL LIST v ument Name: ITS/BASES/TRM CCNAME NAME DEPT LOCATION 1 OPS PROCEDURE GROUP SUPV. OPS PROCEDURE GROUP IP2 3 PLANT MANAGER'S OFFICE UNIT 3(UNIT 3/IPEC ONLY) IP2 5 CONTROL ROOM & MASTER OPS(3PT-D001/6(U3/IPEC) IP3(ONLY) 9 MYERS VALERIE SYSTEM ENG. 72'ELEVATION IP2 4329 11 RES DEPARTMENT MANAGER RES (UNIT 3/IPEC ONLY) 45-4-A 16 BOCCIO JOHN I&C OFFICE (SUPERVISOR) 45-2-A 19 STEWART ANN LICENSING GSB-2D 20 CHEMISTRY SUPERVISOR CHEMISTRY DEPARTMENT 45-4-A 21 TSC(IP3) EEC BUILDING IP2 22 SHIFT MGR.(LUB-001-GEN) OPS (UNIT 3/IPEC ONLY) IP3 23 LIS LICENSING & INFO SERV OFFSITE 25 SIMULATOR TRAIN(UNIT 3/IPEC ONLY) 48-2-A 28 RESIDENT INSPECTOR US NRC 88' ELEVATION. IP2 32 EOF E-PLAN (ALL EP'S) EOF 47 CHAPMAN N BECHTEL OFFSITE 50 TADEMY L. SHARON WESTINGHOUSE ELECTRIC OFFSITE 54 NUC ENGINEERING LIBRARY DOC (UNIT 3/IPEC ONLY) WPO/7A.

55 GSB TECHNICAL LIBRARY A MCCALLION/IPEC,IP2/IP3 GSB-4B 61 SIMULATOR TRAIN(UNIT 3/IPEC ONLY) 48-2-A 69 CONROY PAT LICENSING/ROOM 205 GSB-2D 99 BARANSKI J (ALL) ST. EMERG. MGMT. OFFICE OFFSITE 102 BILYOU CHRISTINE I&C ONLY I&C SUPPORT 45-2-D 106 SIMULATOR INSTRUCT AREA TRG/3PT-D001-D006 ONLY) #48 164 CONTROL ROOM & MASTER OPS(3PT-D0O1/6(U3/IPEC) IP3.(ONLY) 207 TROY M PROCUREMENT ENG. 1A 273 FAISON CHARLENE NUCLEAR LICENSING WPO-12 319 L.GRANT (LRQ-OPS TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 354 L.GRANT(LRQ-OPS/TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 357 L.GRANT(ITS/INFO ONLY) TRAINING - ILO CLASSES 48-2-A 424 GRANT LEAH (9 COPIES) (UNIT 3/IPEC ONLY) #48 474 OUELLETTE P ENG., PLAN & MGMT INC OFFSITE 483 SCHMITT RICHIE MAINTENANCE ENG/SUPV 45-1-A 484 HANSLER ROBERT REACTOR ENGINEERING 72'UNIT 2 485 DRAKE RICH DESIGN ENG/GSB/3RD FLOOR GSB-3B 489 CLOUGHNESSY PAT PLANT SUPPORT TEAM GSB-3B 491 ORLANDO TOM (MANAGER) PROGRAMS/COMPONENTS ENG 45-3-G 492 FSS UNIT 3 OPERATIONS K-IP-I210 493 OPERATIONS FIN TEAM 33 TURBIN DECK 45-1-A 494 AEOF/A.GROSJEAN(ALL EP'S) E-PLAN (EOP'S ONLY) WPO-12D 495 JOINT NEWS CENTER EMER PLN (ALL EP'S) EOF 496 L.GRANT(LRQ-OPS/TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 497 L.GRANT(LRQ-OPS/TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 500 L.GRANT (LRQ-OPS TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 501 L.GRANT (LRQ-OPS TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 512 L.GRANT (LRQ-OPS TRAIN) LRQ (UNIT 3/IPEC ONLY) #48 513 L.GRANT (LRQ-OPS TRAIN) LRQ (UNIT 3/IPEC ONLY) #48

'518 DOCUMENT CONTROL DESK NRC (ALL EP'S) OFFSITE 527 MILIANO PATRICK NRC/SR. PROJECT MANAGER OFFSITE 529 FIELDS DEBBIE OUTAGE PLANNING IP3/OSB

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

~-Enfergy 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 I Enterv CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES TO: DISTRIBUTION DATE: 1012712004 TRANSMITTAL NO:

(Cirde one) I FROM: IPEC DOCUMENT CONTROL: EEC or IP2 53'EL PHONE NUMBER: 271-7057 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/BASESITRM DOC # REV# I TITLE INSTRUCTIONS THE FOLLOWING IS A TECHNICAL SPECIFICATION BASES UPDATE.

INCORPORATE INTO YOUR FILES:

" "*PLEASE NOTE EFFECTIVE DATE***********

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

NAME (PRINT) SIGNATURE DATE CC#

a.

INDIAN POINT 3 TECHNICAL SPECIFICATION BASES INSTRUCTIONS FOR UPDATE: 10-10/27/04 REMOVE INSERT a) List of Effective Sections; a) List of Effective Sections; 3 pages (Rev. 9) 3 pages (Rev. 10) b) Section 3.1.3; Rev. 0 b) Section 3.1.3; Rev. 1 7 pages 7 pages c) Section 3.3.5, Rev. 0 c) Section 3.3.5, Rev. 1 6 pages 6 pages d) Section 3.4.3, Rev. 0 d) Section 3.4.3, Rev. 1 9 pages 9 pages e) Section 3.4.12, Rev. 0 e) Section 3.4.12, Rev. 1 20 pages 20 pages f) Section B 3.5.1; Rev. 0 f) Section B 3.5.1; Rev. 1 10 pages 10 pages j

TECHNICAL SPECIFICATION BASES LIST OF EFFECTIVE SECTIONS BSECTOS E NUMBER IEFFECTIVE BASES SECTION REVI I NUMBER OF PAGES IEFFECTIVE SECTION REV OF PAGES DATE DATE Tb! of Cnt 1 4 05/18/2001 .,-:_---_ B 3.6 CONTAINMENT.

v cB 2;0 SAFETY LIMITS ;K<-'

. B 3.6.1 0 5 03/19/2001 B 2.1.1 l 0 l 5 03119/2001 B 3.6.2 0 9 03/19/2001 B 2.1.2 l 0 l 4 l03/19/2001 B 3.6.3 0 17 03/19/2001

':- I ;' B 3.0 LCOAND SRAPPLICABILITY-. '- B 3.6.4 0 3 03/19/2001 B 3.0 l 1 l 15 09/30/2002 B 3.6.5 1 5 06/20/2003 B 3.1 REACTIVITY-CONTROL, B 3.6.6 1 13 12/04/2002 B 3.1.1 0 6 03/19/2001 B 3.6.7 0 6 03/19/2001 B 3.1.2 0 7 03/19/2001 B 3.6.8 0 6 03/19/2001 B 3.1.3 1 7 1012712004 B 3.6.9 0 8 03/19/2001 B 3.1.4 0 13 03/19/2001 B 3.6.10 0 12 03/19/2001 B 3.1.5 0 5 03/19/2001 - B 3.7 PLANT SYSTEMS': -

B 3.1.6 0 6 03/19/2001 B3.7.1 1 6 12/04/2002 B 3.1.7 0 8 03/19/2001 B 3.7.2 0 10 03/19/2001 B 3.1.8 0 7 03/19/2001 B 3.7.3 1 7 05/18/2001

- B 3.2 POWER DISTRIBUTION LIMITS ,- f. B 3.7.4 0 5 03/19/2001 B 3.2.1 0 7 03/19/2001 B 3.7.5 0 11 03/19/2001 B 3.2.2 0 7 03/19/2001 B 3.7.6 1 4 12/04/2002 B 3.2.3 0 9 03/19/2001 B 3.7.7 0 4 03/19/2001 B 3.2.4 0 7 03/19/2001 B 3.7.8 0 7 03/1912001

___, B-3.3 INSTRUMENTATION B 3.7.9 1 9 09/30/2002 B 3.3.1 1 59 09/30/2002 B 3.7.10 0 3 03/19/2001 B 3.3.2 3 45 12/04/2002 B 3.7.11 2 9 06/20/2003 B 3.3.3 2 19 09/30/2002 B 3.7.12 0 4 03/19/2001 B 3.3.4 0 7 03/19/2001 B 3.7.13 2 7 06/2012003 B 3.3.5 1 6 10/27/2004 I B 3.7.14 0 3 03/19/2001 B 3.3.6 0 10 03/19/2001 B 3.7.15 0 5 03/1912001 B 3.3.7 0 6 03/19/2001 B 3.7.16 0 6 03/19/2001 B 3.3.8 1 4 03/17/2003 B 3.7.17 0 4 03/19/2001

' ~ 3.4 REACTOR COOLANT.SYSTEM.'.-: ..

-B . ,- B 3.8 ELECTRICAL POWER-B 3.4.1 0 6 03/19/2001 B 3.8.1 1 32 01/22/2002 B 3.4.2 0 3 03/19/2001 B 3.8.2 0 7 03/19/2001 B 3.4.3 1 9 10127/2004 B 3.8.3 0 13 03/19/2001 B 3.4.4 0 4 03/19/2001 B 3.8.4 1 11 01/22/2002 B 3.4.5 0 6 03/19/2001 B 3.8.5 0 4 03/19/2001 B 3.4.6 0 6 03/19/2001 B 3.8.6 0 8 03/19/2001 B 3.4.7 0 7 03/19/2001 B 3.8.7 1 8 06/20/2003 B 3.4.8 0 4 03/19/2001 B 3.8.8 1 4 06/20/2003 B 3.4.9 2 5 06/20/2003 B 3.8.9 2 14 06/20/2003 B 3.4.10 0 5 03/19/2001 B 3.8.10 0 4 03/19/2001 B 3.4.11 0 8 03/19/2001 B 3.9 REFUELING OPERATIONS' -va B 3.4.12 1 20 10/27/2004 B 3.9.1 0 4 03/19/2001 B 3.4.13 2 6 11/19/2001 B 3.9.2 0 4 l 03/19/2001 B 3.4.14 0 10 03/19/2001 B 3.9.3 1 8 03/17/2003 B 3.4.15 2 7 11/19/2001 B 3.9.4 0 4 03/19/2001 B 3.4.16 0 7 03/19/2001 B 3.9.5 0 4 l 03/19/2001

__ _ _' B 3.5 ECCS -'_-_'_ B 3.9.6 0 4 03/19/2001 B 3.5.1 J 1 l 10 10/27/2004 B 3.5.2 0 13 03/19/2001 B 3.5.3 0 4 03/19/2001 B 3.5.4 1 0 l 9 103/19/2001 INDIAN POINT 3 Page 1 of 3 Revision 10

TECNICL PECFICTIN BSE 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 PACKAGE 01-031901 '-jf- i u g.;:

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 02-051801:-- .'~-; '

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-03-11 1901 Correction to statement regarding applicability of Function B 3.3.2 1 11/19/01 5, to be consistent with the Technical Specification.

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

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

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

,BASES UPDATE PACKAGE:04-012202,- a- --.

B 3.3.2 2 01/22/02 Clarify starting logic of 32 ABFP per EVL-01-3-078 MULTI, Rev 0.

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

B 3.8.4 1 01/22/02 Revision of battery design description per plant

_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 UPDATE PACK GE 05-093002-4'r 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 3 Revision 10

TECHNICAL SPECIFICATION BASES REVISION HISTORY DESCRIPTION UTE PACKAGE 06-120402 ','i !!,. i Changes to reflect Tech Spec Amendment 213 regarding 1.4% power uprate.

t_-_________ ______- ________ BASES.UPDATESPACKAGE 07-031703 :tI-;

_ r <t ;%i ;`

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

XBASES UPDATE-PACKAGE:09-062003tf- - i B 3.4.9 2 06120/2003 Changes to reflect commitment for a dedicated operator per Tech Spec Amendment 216.

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

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

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

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

B 3.8.9 2 06/2012003 K.,s:!-;

d; '".;I,:

-, BASES UPDATE.PACKAGE1 0-1002704*.-l; t:,,,7,-I B 3.1.3 1 10127/2004 Clarification of the surveillance requirements for TS 3.1.3 per 50.59 screen.

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

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

B 3.5.1 1 1012712004 Changes to reflect Tech Spec Amendment 222 regarding extension of completion time for Accumulators.

INDIAN POINT 3 Page -3 of 3 Revision 10

MTC B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND 'According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that'reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a" negative MTC over the largest possible range of fuel cycle operation. Therefore,'a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.

MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by measurements. Both initial and reload cores are designed so that the beginning of life (BOL) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, suchlas fuel loading-and reactor coolant soluble boron concentration; The'core design may require additional fixed distributed'poisons to yield' an MTC at BOL within the range analyzed in-the'plant accident analysis. The end of life (EOL)

MTC is also limited'by the requirements of the accident analysis.

Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOL limit.

The limitations on MTC are provided to ensure that the value of this coefficient remains'within the limiting conditions assumed in the FSAR accident and transient analyses.

(continued)

INDIAN POINT 3 B. 3.1.3 - 1 Revision I

- - l MTC B 3.1.3 BASES BACKGROUND If the LCO limits are not met, the unit response during (continued) transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity.

The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.

APPLICABLE SAFETY ANALYSES The acceptance criteria for the specified MTC are:

a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2): and

b. The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

The FSAR, Chapter 14 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core.

MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref.

3).

The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 2) or RTP, loss of main feedwater flow, and loss of forced reactor coolant flow.

(continued)

INDIAN POINT 3 B. 3.1.3 - 2 Revision 1

MTC B 3.1.3 BASES APPLICABLE SAFETY ANALYSES (continued)

The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature. -

In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions

- appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether.-it is the BOL or EOL. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2).

MTC values are bounded in reload safety evaluations assuming steady state conditions at BOL and EOL.- An EOL measurement is

'conducted at conditions when the RCS boron concentration reaches approximately 300 ppm. The measured value may be extrapolated to project the EOL value, in order to confirmreload design

predictions.

MTC satisfies Criterion 2 of 10 CFR 50.36. Even though it is not directly observed'andcontrolled from the control room, MTC is considered an initial'condition process variable because of its

' . dependence on boron'concentration.

LCO LCO 3.1.3 requires the MTC to be within'specified limits of the COLR to ensurethat the core operates within the assumptions of the accident analysis., During the reload core safety evaluation, the MTC is analyzed 'to determine that its values remain within the bounds of the original accident analysis during operation.

Assumptions made insafety analyses require that the MTC be less positive than .agiven upper bound and-more positive than a given lower bound. The MTC is most positive near BOL; this upper bound must not be exceeded. This'maximum upper limit occurs at BOL, all rods out (ARD).,hot zero power-conditions.

(continued)

INDIAN POINT 3 B. 3.1.3 - 3 .Revision I

MTC B 3.1.3 BASES LCO At EOL the MTC takes on its most negative value, when (continued) the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.

During operation, therefore, the conditions of the LCO can only be ensured through measurement. The Surveillance checks at BOL and EOL on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met.

The LCO establishes a maximum positive value that cannot be exceeded. The BOL positive limit and the EOL negative limit are established in the COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule.

APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above.

In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled CONTROL ROD assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3. in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6. this LCO is not applicable, since no Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES.

ACTIONS A.1 If the BOL MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits. The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides enough time for evaluating the MTC measurement and computing the required bank withdrawal limits.

(continued)

INDIAN POINT 3 B. 3.1.3 - 4 Revision I

MTC B 3.1.3 BASES ACTIONS A.1 (continued)

As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. 'Using-physics-calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists.- The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect.

B.1 If the required administrative withdrawal limits at BOL are not established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the unit must be brought to MODE 2 with kff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses.'

The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experiencefor reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

C.1 Exceeding the EOL MTC limit means that the safety analysis assumptions for the'EOL accidents that use a bounding negative MTC value may be invalid. If the EOL MTC limit is exceeded, the plant must be brought to a MODE or condition in which the LCO requirements are not'applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed.Completion Time is reasonable, based on operating experience, for reaching'the required MODE from full power conditions in an orderly manner and-without challenging plant systems.

(continued)

INDIAN POINT 3 B. 3.1.3 - 5 Revision I

MTC B 3.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.1.3.1 This SR requires measurement of the MTC at BOL prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Meeting the limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.

The BOL MTC value for ARO will be inferred from isothermal temperature coefficient measurements obtained during the physics tests after refueling. The ARO value can be directly compared to the BOL MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for control banks.

SR 3.1.3.2 In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOL full power conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value.

Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOL LCO limit. The 300 ppm SR value is sufficiently less negative than the EOL LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.

SR 3.1.3.2 is modified by three Notes that include the following requirements:

1. This SR is not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm. This note alters the FREQUENCY to once each cycle within 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP ARO boron concentration of 300 ppm.

(continued)

INDIAN POINT 3 B. 3.1.3 - 6 Revision 1

MTC B 3.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.1.3.2 (continued)

2. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOL limit on MTC could be reached before the planned EOL. Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOL limit. This note establishes a new required action and completion time. The required action, verify the MTC is within the COLR lower limit (which is a repeat of the surveillance), occurs when the existing surveillance requirement (i.e., to verify the MTC is more positive than the limit specified in the COLR for a 300 ppm boron concentration) fails. The frequency is 14 EFPD after the initial surveillance test fails and every 14 EFPD thereafter.

3. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is more positive than the 60 ppm Surveillance limit, the EOL limit will not be exceeded because of the gradual manner in which MTC changes with core burnup. This note acts to limit the action requirement in Note 2. It allows the action to repeat the surveillance to be terminated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of < 60 ppm is less negative than the 60 ppm surveillance limit specified in the COLR.

REFERENCES 1. 10 CFR 50. Appendix A.

2. FSAR, Chapter 14.

3. WCAP 9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

INDIAN POINT 3 B. 3.1.3 - 7 Revision I

RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS'are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients,' and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

LCO 3.4.3, Figure 3.4.3-1, Heatup Limitations for the Reactor Coolant System, Figurei'3.4.3-2, Cooldown Limitations for the Reactor Coolant System, a'nd Figure 3.4.3-3, Hydrostatic and Inservice Leak Testing Limitations for the'Reactor Coolant System.

contain P/T limit curves for heatup, cooldown, and inservice leak and hydrostatic (ISLH) testing, respectively (Ref. 1).

Each P/T limit curve defines an'acceptable region for normal operation. The usual use of the curves is operational guidance

'during heatup or cooldown maneuvering, when pressure and temperature Iindications are monitored and compared to the applicable curve to determine'that operation is within the allowable region. The happy face icon'shown~on Figure 3.4.3-1, Figure, 3.4.3-2, and Figure 3.4.3-3. indicates the side of the curve in which operation'is-permissible. Conversely, the sad face icon indicates the side of the curve in which operation is prohibited.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits do not apply to the pressurizer, which has different design characteristics and operating functions.

(continued)

INDIAN POINT 3 B 3.4.3 - 1 Revision 1

RCS P/T Limits B 3.4.3 BASES BACKGROUND 10 CFR 50, Appendix G (Ref. 2), requires the (continued) establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 2 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section III, Appendix G (Ref.

3).

The neutron embrittlement effect on the material toughness is reflected by increasing the nil ductility reference temperature (RTNDT) as exposure to neutron fluence increases.

The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 4) and Appendix H of 10 CFR 50 (Ref. 5). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Regulatory Guide 1.99 (Ref. 6).

The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.

(continued)

INDIAN POINT 3 B 3.4.3 - 2 Revision 1

RCS P/T Limits B 3.4.3 BASES BACKGROUND The consequence of violating the LCO limits is that the (continued) RCS has been'o'perated under conditions that can result in brittle failure'of the RCPB. possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to'determine the effect on the structural integrity of the RCPB components. The ASME Code,Section XI. Appendix E (Ref. 7). provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.

APPLICABLE SAFETY ANALYSES, The P/T limits are not derived from Design Basis Accident (DBA) analyses. 'They are prescribed during normal 'operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB. an 'unanalyzed condition.

Reference 1 establishes the methodology for determining the P/T limits. Although.the.P/T limits are not derived from any DBA, the P/T limits are acceptance'limits since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy;Criterion'2 of 30 CFR 50.36.

LCO The two elements of this LCO are:

a. The limit curves for heatup, cooldown, and ISLH testing; and

b. Limits on the rate of change of temperature.

Figure 3.4.3-1, Heatup Limitations for the Reactor Coolant System, Figure. 3.4.3-2. Cooldown Limitations fort the Reactor Coolant System, and Figure 31.4.3-3. Hydrostatic and Inservice Leak Testing Limitations for the Reactor Coolant System, contain P/T limit curves for heatup, cooldown. and inservice leak and hydrostatic (ISLH) testing, respectively. These figures specify the maximum RCS pressure for various heatup and cooldown rates at (continued)

INDIAN POINT 3 B 3.4.3 - 3 Revision I

RCS P/T Limits B 3.4.3 BASES LCO any given reactor coolant temperature. The figures (continued) provide the limiting RCS pressure and reactor coolant temperature combination for reactor coolant temperature heatup rates up to 600F/hr and reactor coolant temperature cooldown rates up to 1000F/hr. Therefore, heatup rates that exceed 60°F/hr and cooldown rates that exceed 100 0F/hr are considered not within the limits of this LCO.

The LCO limits apply to all components of the RCS pressure boundary, except the pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.

The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves. Heatup and cooldown limits are specified in hourly increments (i.e., the heatup and cooldown limits are based on the temperature change averaged over a one hour period). Limit lines for cooldown rates between those presented may be obtained by interpolation.

Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and

c. The existence, size, and orientation of flaws in the vessel material.

(continued)

INDIAN POINT 3 B 3.4.3 - 4 Revision I

RCS P/T Limits B 3.4.3 BASES APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for preventionof fnonductile failure in accordance with 10 CFR 50, Appendix G (Ref. 2). Although the P/T limits were developed to provide guidance for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, their Applicability is at all times in keeping with the concern for nonductile failure. The limits do not apply to the pressurizer.

'During MODES I'and 2. other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits. iLCO 3.4.1.'"RCS Pressure.

Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits"; LCO 3.4.2, "RCS Minimum Temperature for Criticality"; and Safety Limit 2.1,'"Safety Limits," also provide operational restrictions for pressure and temperature and maximum pressure.

Furthermore, MODES Sand 2 are above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.

ACTIONS A.1 and A.2 Operation outside the P/T limits during MODE 1. 2, 3. or 4 must be corrected so that'the RCPB is returned to a condition that has been verified by stress analyses.

-The 30 minute Completion'Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe.'and the'activity can be accomplished in this time in a controlled-manner.'

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity'remains acceptable and must be completed before'continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.

(continued)

INDIAN POINT 3 B 3.4.3 - 5 Revision I

RCS P/T Limits B 3.4.3 BASES ACTIONS A.1 and A.2 (continued)

ASME Code,Section XI, Appendix E (Ref. 7). may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable to accomplish the evaluation. The evaluation for a mild violation is possible within this time, but more severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed before continuing to operate.

Condition A is modified by a Note requiring Required Action A.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. In reduced pressure and temperature conditions, the possibility of propagation with undetected flaws is decreased.

If the required restoration activity cannot be accomplished within 30 minutes, Required Action B.1 and Required Action B.2 must be implemented to reduce pressure and temperature.

If the required evaluation for continued operation cannot be accomplished within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the results are indeterminate or unfavorable, action must proceed to reduce pressure and temperature as specified in Required Action B.1 and Required (continued)

INDIAN POINT 3 B 3.4.3 - 6 Revision I

RCS P/T Limits B 3.4.3 BASES ACTIONS B.1 and B.2 (continued)

Action B.2. A favorable evaluation must be completed and documented before returning to operating pressure and temperature conditions.-

Pressure and temperature are reduced by bringing the plant to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 with RCS pressure < 500 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .. Note that LCO 3.4.12, Low Temperature Overpressure Protection (LTOP), will also apply and may require limits for

operation that are more restrictive than or supplement this limit.

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

C.1 and C.2 Actions must be initiated immediately to correct operation outside of the P/T limits at times other than when in MODE 1, 2, 3. or 4, so that the RCPB is returned to a condition that has been verified by stress analysis.

The immediate Completion Time reflects the urgency of initiating action to restore'the'parameters to within the analyzed range.

Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.

Besides'restoring operation'within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and completed prior into MODE 4. Several methods may be used. including comparison with pre-analyzed transients in the stress analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. 7), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

(continued)

INDIAN POINT 3 B 3.4.3 - 7 Revision I

- a-RCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE REQUIREMENTS SR 3.4.3.1 Verification that operation is within the PTLR limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Heatup and cooldown limits are specified in hourly increments (i.e., the heatup and cooldown limits are based on the temperature change averaged over a one hour period). Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

(continued)

INDIAN POINT 3 B 3.4.3 - 8 Revision I

RCS P/T Limits B 3.4.3 BASES REFERENCES 1. WCAP-7924-A, July 1972.

2. 10 CFR 50. Appendix G.

3. ASME, Boiler and Pressure Vessel Code.Section III, Appendix G.

4. ASTM E 185-70.

5. 10 CFR 50, Appendix H.

6. Regulatory Guide 1.99, Revision 2. May 1988.

7. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

8. WCAP-16037. Revision 1. "Final Report on Pressure-Temperature Limits for Indian Point Unit 3 NPP", Westinghouse Electric Company, May 2003.

INDIAN POINT 3 B 3.4.3 - 9 Revision I

LTOP B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP)

BASES "

BACKGROUND LTOP is established to limit RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50. Appendix G (Ref. 1). The reactor vessel is the limiting RCPB component for demonstrating such protection.

LCO 3.4.12, Figure 3.4.12-1 provides the maximum allowable nominal actuation logic setpoints for the power operated relief valves (PORVs) and the maximum.RCS, pressure for the coldest existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the LTOP MODES.

The reactor vessel material is less tough at low temperatures than at normal operating temperature. As'the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low~at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown because a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LCO 3.4.3, 'RCS Pressure and Temperature (P/T) Limits,"

requires administrative control of RCSpressure and temperature during heatup'and cooldown to prevent exceeding the limits in Figure 3.4.12-1.'

When the RHR System is isolated from the RCS, the RHR System is protected from overpressure by two'spring'loaded relief valves (SI-733A and SI-733B). When the RHR-System is not isolated from the RCS, the RHR System is protected from overpressure by spring loaded relief valve '(i.e., AC-1836) which has sufficient capacity to accommodate all 3 charging pumps.', However, this relief (continued)

INDIAN POINT 3 B 3.4.12 -I Revision 1

LTOP B 3.4.12 BASES BACKGROUND valve does not have sufficient capacity to ensure that (continued) the RHR system does not exceed design pressure limits during a mass addition resulting from an inadvertent injection of one or more high head safety injection (HHSI) pumps. Therefore, LTOP requirements are used to protect the RHR System whenever the RHR System is not isolated from the RCS.

This LCO provides RCS overpressure protection by limiting maximum coolant input capability and having adequate pressure relief capacity. Limiting coolant input capability is achieved by not permitting any High Head Safety Injection (HHSI) pumps to be capable of injection into the RCS and isolating the accumulators.

The pressure relief capacity requires either two redundant power operated relief valves (PORVs) or a depressurized RCS and an RCS vent of sufficient size. One PORV or the open RCS vent is sufficient to provide overpressure protection to terminate an increasing pressure event. Alternately, if redundant PORVs are not Operable or an RCS vent cannot be established, LTOP protection may be established by limiting the pressurizer level to within limits specified in Figure 3.4.12-2 and Figure 3.4.12-3 consistent with the number of charging pumps and number of high head safety injection (HHSI) pumps capable of injecting into the RCS. This approach is acceptable because pressurizer level can be maintained such that it will either accommodate any anticipated pressure surge or allow operators time to react to any unanticipated pressure surge. When pressurizer level is used to satisfy LTOP requirements, operator action is assumed to terminate the unplanned HHSI pump injection within 10 minutes.

With high pressure coolant input capability limited, the ability to create an overpressure condition by coolant addition is restricted. The LCO does not require the makeup control system deactivated or the safety injection (SI) actuation circuits blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve. There is no restriction on the status of charging pumps when LTOP is established using either a PORV or an RCS vent. If conditions require the use of more than one HHSI pump for makeup in the event of loss of inventory, then pumps can be made available through (continued)

INDIAN POINT 3 B 3.4.12 - 2 Revision I

LTOP B 3.4.12 BASES BACKGROUND manual actions. Charging pumps and low pressure injection systems (continued) are available to provide makeup even when LTOP requirements are applicable.

When configured to provide low temperature overpressure protection, -the PORVs are part of the Overpressure Protection System (OPS). LTOP for'ppressure relief can consist of either the OPS (two PORVs with reduced lift settings), or a depressurized RCS and an RCS vent of sufficient size. Two PORVs are required for redundancy. One PORV has adequate relieving capability to keep from overpressurization for'the required coolant input capability.

PORV Requirements, '

The Overpressure Protection System (OPS):provides the low temperature overpressure protection by controlling the Power Operated Relief Valves (PORVs) and their associated block valves with pressure setpoints that vary with RCS cold leg temperature.

Specifically,'coldcleg'temperature signals from three RCS loops are supplied to three associated function generators that calculate the maximum RCS pressures allowed at those temperatures.

The maximum RCS pressure limits at any RCS temperature correspond to the 10 CFR 50. App'endix G,'limit curve maintained in the Pressure and Temperature Limits Report and are used as the OPS pressure setpoint. Having the setpoints of both valves within the limits in Figure 3.4.12-1 ensures that the Reference 1 limits will not be exceeded in any analyzed event.

In addition to generating the OPS pressure'setpoint, the same cold leg temperature signals .are used to "arm" the OPS when RCS temperature falls below ,the temperature at which low temperature overpressure protection is required (319 0F). This temperature includes an allowance'of 14.4 0F'for instrument uncertainty and margin. Each PORV opens when a two-out-of-two'(temperature and pressure) coincidence logic' is satisfied. OPS is "armed" when RCS temperature falls' below'the temperature that satisfies one half of the two-out-of-two (temperature-pressure)'coincidence logic. When OPS is enabled, the'PORVs will open if RCS pressure exceeds the calculated pressure setpoint that varies with RCS temperature.

(continued)

INDIAN POINT 3 B 3.4.12-3 Revision 1

LTOP B 3.4.12 BASES BACKGROUND The PORV block valves open when the RCS temperature (continued) falls below the OPS arming temperature. Note that the control switches for the PORV and PORV block valves must be in the AUTO position and the OPS states links closed for OPS signals to actuate the PORVs.

Three channels of RCS cold leg temperature are used in the two-out-of-three coincidence logic to satisfy the temperature portion of the two-out-of-two (temperature and pressure) coincidence logic for each PORV. Three channels of RCS pressure are used in a two-out-of-three coincidence logic to satisfy the pressure portion of the two-out-of-two (temperature-pressure) coincidence logic for each PORV. Use of a two-out-of-three coincidence logic for pressure and for temperature ensures that a single failure will not cause or prevent an OPS actuation. Use of two PORVs, each with adequate relieving capability to prevent overpressurization, ensures that a single failure will not prevent an OPS actuation.

When a PORV is opened in an increasing pressure transient, the release of coolant will cause the pressure increase to slow and reverse. As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the reset pressure as the valve closes.

RCS Vent Requirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient, and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.

Multiple methods exist for establishing the required RCS vent capacity including removing or blocking open a PORV and disabling its block valve in the open position. An RCS vent of > 2.00 square inches when no HHSI pump is capable of injecting into (continued)

INDIAN POINT 3 B 3.4.12 - 4 Revision I

LTOP B 3.4.12 BASES

'BACKGROUND the RCS; or, an RCS vent with opening greater than or equal to one (continued) pressurizer code safety valve flange and up to two HHSI pumps capable of injecting Jnto the RCS will satisfy LTOP requirements because either configuration ensures pressure limits are not exceed during a transient. Alternately. an RCS vent of >

2.00 square inches coupled with a pressurizer level < 0% and up to two HHSI pumps capable of injecting into the RCS will satisfy LTOP

-requirements because it ensures a minimum of 10 minutes for operator action before pressure limits are exceeded during a transient. The vent path(s) must be above the level of reactor coolant, so as not to drain the RCS when open.

APPLICABLE SAFETY ANALYSES Safety analyses (Ref. 3) demonstrate that the reactor vessel is adequately protected against exceeding the Reference 1 P/T limits.

In MODES 1. 2, and 3, with RCS cold leg temperature exceeding 411 0F, the pressurizer safety valves will prevent RCS pressure from exceeding the Reference 1 limits. At 319 OF and below, overpressure prevention falls to two OPERABLE PORVs in conjunction with the Overpressure.Protection System (OPS) or to a depressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure relief capability. Alternately, if redundant PORVs are not Operable, Low Temperature Overpressure protection may be maintained by limiting the pressurizer level to within limits.-specifiedin Figure 3.4.12-2 and Figure 3.4.12-3 consistent with the number of charging pumps and number of high head safety injection (HHSI) pumps capable of injecting into the RCS. This approach is acceptable because pressurizer level can be established to either-accommodate any anticipated pressure surge or allow operators time to react to any unanticipated pressure surge.

When the RCS temperature is greater than the LTOP arming temperature (i.e., > 319 0F) but below the minimum temperature at which the pressurizer safety valves lift prior to violation of the 10 CFR 50, Appendix G. limits (i.e., < 380 0F), administrative controls in the Technical Requirements Manual (TRM) (Ref. 4) are I used to limit the potential for exceeding 10 CFR 50, Appendix G.

(continued)

INDIAN POINT 3 B 3.4.12 - 5 Revision I

LTOP B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued) limits. These administrative controls may include operating with a bubble in the pressurizer and/or otherwise limiting plant time or activities when the RCS temperature is in the specified range.

The use of administrative controls to govern operation above the LTOP arming temperature but below the minimum temperature at which the pressurizer safety valves lift prior to violation of the 10 CFR 50, Appendix G. limits is consistent with the guidance provided in Generic Letter 88-011, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations (Ref.2). GL 88-011 states that automatic, or passive, protection of the P-T limits will not be required but administratively controlled when in the upper end of the 10 CFR 50, Appendix G. temperature range.

The actual temperature at which the pressure in the P/T limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each time the Figure 3.4.12-1 curves are revised, LTOP must be re-evaluated to ensure its functional requirements can still be met using the OPS (PORVs) method or the depressurized and vented RCS condition.

Figure 3.4.12-1 contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated against the Ref. 3 analyses to determine the impact of the change on the LTOP acceptance limits.

Transients that are capable of overpressurizing the RCS are categorized as either mass or heat input transients, examples of which follow:

Mass Input Type Transients

a. Inadvertent safety injection; or

b. Charging/letdown flow mismatch.

(continued)

INDIAN POINT 3 B 3.4.12 - 6 Revision 1

LTOP B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

Heat Input Type Transients

a. Inadvertent actuation of pressurizer heaters;

b. Loss of RHR cooling: or

c. Reactor coolant pump'(RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

The following are required during the LTOP MODES to ensure that mass and heat input transients do not occur.' This is accomplished by the following:

a. Rendering all HHSI pumps incapable of injection;

b. Deactivating the accumulator discharge isolation valves in their ~closed positions or maintaining accumulator pressure less than the maximum RCS pressure for the coldest existing RCS cold leg temperature allowed by the P/T limit curves provided in Figure 3.4.12-1: and

c. Disallowing start of an RCP unless conditions are established that ensure a RCP pump start will not cause a pressure excursion that will exceed LTOP limits. Required conditions for starting a RCP when LTOP is required include a combination of primary and secondary water temperature differences and Overpressure Protection System (OPS) status or pressurizer level. Meeting the LTOP RCP starting surveillancesensures that theses 'conditions are satisfied prior to a RCP pump start. -

The Ref. 3 analyses demonstrate that either one PORV or the depressurized RCS'and RCS vent can maintain RCS pressure below limits when no HHSI pump is capable of injecting into the RCS.

This assumes an RCS vent'of > 2.OO square inches. The same protection can be provided when up to two HHSI pumps are capable of injecting into the RCS assuming an RCS vent with opening greater than or equal. to one code pressurizer safety valve flange.

Alternately, LTOP requirements can be satisfied by various (continued)

INDIAN POINT 3 B 3.-4.12 -_7 Revision I

LTOP B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued) combinations of pressurizer level. RCS pressure, and RCS injection capability (i.e., maximum number of HHSI pumps and/or charging pumps) shown in Figure 3.4.12-2 and 3.4.12-3. These combinations of pressurizer level, RCS pressure, and RCS injection capability satisfy LTOP requirements by ensuring a minimum of 10 minutes for operator action to terminate an unplanned event prior to exceeding maximum allowable RCS pressure. None of the analyses addressed the pressure transient need from accumulator injection, therefore, when RCS temperature is low, the LCO also requires the accumulator isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the coldest existing RCS cold leg temperature allowed in Figure 3.4.12-1.

If the accumulators are isolated and not depressurized, then the accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions.

Fracture mechanics analyses established the temperature of LTOP Applicability at 319 OF.

The consequences of a loss of coolant accident (LOCA) in LTOP MODE 4 conform to 10 CFR 50.46 and 10 CFR 50, Appendix K (Refs. 5 and

6) requirements by having ECCS OPERABLE in accordance with requirements in LCO 3.5.3, ECCS-Shutdown.

PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limit shown in Figure 3.4.12-1. The setpoints are derived by analyses that model the performance of the LTOP System, assuming the limiting LTOP transient with HHSI not injecting into the RCS. These analyses consider pressure overshoot and undershoot beyond the PORV opening and closing, resulting from signal processing and valve stroke times. The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met. The OPS setpoint is based (continued)

INDIAN POINT 3 B 3.4.12 - 8 Revision I

LTOP B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued) on a comparative analysis of Reference 3, with allowances for metalifluid temperature differences, static head due to elevation differences, and dynamic head from the operation of the reactor coolant pumps and RHR pumps.

The PORV setpoints.in Figure 3.4.12-1 will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are periodically modified as the reactor vessel material toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence-projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"

discuss these examinations.

The PORVs are'considered active components. Thus, the failure of one PORV is assumed to-represent the worst case, single active failure.

RCS Vent Performance With the RCS depressurized, analyses show a vent size of 1.4 square inches is capable'of mitigating the-allowed LTOP

- overpressure transient assuming no HHSI pumpaand no accumulator injects into the RCS. The LCO limit for an'RCS vent is conservatively established at'2.00 square inches. The capacity of a vent this size is greater than the flow of the limiting transient for the LTOP configuration, maintaining RCS pressure less than'the maximum pressure oh the P/T limit curve. An RCS vent with opening greater than or equal to one pressurizer code safety valve flange and up to two HHSI pumps capable of injecting into the RCS will-satisfy LTOP requirements because it ensures pressure limits are' not exceed during a transient. An RCS vent of

> 2.00 square inches coupled with a pressurizer level < 0% and up to two HHSI pumps capable of injecting into the RCS will satisfy

, -(continued)

INDIAN POINT 3 B 3.4.12 I .- , -. 9 Revision I

LTOP B 3.4.12 BASES APPLICABLE SAFETY ANALYSES (continued)

LTOP requirements because it ensures a minimum of 10 minutes for operator action before pressure limits are exceeded during a transient.

The RCS vent size will be re-evaluated for compliance each time the P/T limit curves are revised based on the results of the vessel material surveillance.

The RCS vent is passive and is not subject to active failure.

LTOP satisfies Criterion 2 of 10 CFR 50.36.

LCO This LCO requires that LTOP is OPERABLE. LTOP is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant input capability, the LCO requires that no HHSI pumps be capable of injecting into the RCS and all accumulator discharge isolation valves closed and de-energized if accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in Figure 3.4.12-1, Maximum Allowable Nominal PORV Setpoint for LTOP (OPS).

The elements of the LCO that provide low temperature overpressure mitigation through pressure relief are:

a. Two OPERABLE PORVs configured as part of an OPERABLE Overpressure Protection System (OPS); or

b. A depressurized RCS and an RCS vent.

A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set to the limit required by Figure 3.4.12-1 and testing proves its ability to open at this setpoint, and motive power is available to the two valves and their control circuits.

(continued)

INDIAN POINT 3 B 3.4.12 - 10 Revision 1

LTOP B 3.4.12 BASES LCO The OPS is OPERABLE for LTOP when there are three OPERABLE RCS (continued) pressure channels and three OPERABLE RCS temperature channels.'

The OPS is still'OPERABLE when an inoperable RCS pressure or temperature channel'is in the tripped condition. OPS is considered OPERABLE for meeting LCO 3.4.12 requirements even if one or. two RCS cold leg temperatures is above the LTOP Applicability limit.

An RCS vent-is OPERABLE when open,with-an area of > 2.00 square inches.

Each of these methods 'fboverpressure prevention is capable of mitigating the limiting LTOP transient.

APPLICABILITY This LCO is applicable whenever the RHR System is not isolated from the.RCS to protect the RHR system.piping. When all RCS cold leg temperatures are >-319 0F, RHR system'piping is adequately protected-by making'the accumulators and'all'HHSI pumps incapable of injecting into the RCS. Therefore, a Note in the LCO specifies that requirements for the OPS System and/or an RCS vent are not Applicable when all RCS cold leg temperatures are > 319 "F.

This LCO is applicable to-provide protection -forthe RCS pressure boundary in MODE 4,when any RCS cold leg'temperature is < 319 OF, in MODE 5. and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above 319 OF. When the reactor vessel head'is-off, overpressurization cannot occur. Although LTOP is not Applicable when the RCS temperature is greater than the LTOP arming temperature (i.e., >,319 OF) but below the minimum temperature at which the pressurizer safety valves lift prior to violation of the'1 CFR 50, Appendix G. limits(i.e., < 380 OF).

administrative'controls in the Technical Requirements Manual (TRM)

(Ref. 4)'are used'to limit the potential for exceeding 10 CFR 50. I Appendix G,'limits.' LCO 3.4.3 provides the operational P/T limits (continued)

INDIAN POINT 3 B 3.4.12 - 1 1 Revision I

LTOP B 3.4.12 BASES APPLICABILITY for all MODES. LCO 3.4.10, "Pressurizer Safety Valves." requires (continued) the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3, and MODE 4 above 319 OF when the RHR system is isolated from the RCS.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

The Applicability is modified by three Notes, Note 1 states that accumulator isolation is only required when the accumulator pressure is more than the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves. This Note permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

Note 2 ensures that LCO 3.4.12 will not prohibit a HHSI pump being energized and aligned to the RCS as needed to support emergency boration or to respond to a loss of RHR cooling.

Note 3 specifies that one HHSI pump may be made capable of injecting into the RCS for a period not to exceed 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to perform pump testing. During testing, administrative controls are used to ensure that HHSI testing will not result in exceeding RCS or RHR system pressure limits.

ACTIONS A.1. A.2.1. A.2.2. A.2.3. A.3.1 and A.3.2 When one or more HHSI pumps are capable of injecting into the RCS, LTOP assumptions regarding limits on mass input capability may not be met. Therefore, immediate action is required to limit injection capability consistent with the LTOP analysis assumptions and the existing combination of pressurizer level and RCS venting capacity. Required Action A.1 requires restoration with LCO (continued)

INDIAN POINT 3 B 3.4.12 -12 Revision 1

LTOP B 3.4.12 BASES ACTIONS A.1. A.2.1. A.2.2. A.2.3. A.3.1 and A.3.2 (continued) requirements. Required Actions A.2 and A.'3 require verification and periodic re-verification that alternate LTOP configurations are met. The Completion Times of immediately reflects the urgency that one of the acceptable LTOP configurations is established as soon as possible.

B.1. C.1 and C.2 To be considered isolated, an accumulator must'have its discharge valves closed and the valve power supply breakers fixed in the open position. I An unisolated accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This is only required when the accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves...

If isolation is needed and cannot be accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Required Action C.1,and Required Action C.2 provide two options, either of which must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By increasing the RCS temperature to > 319 OF, an accumulator pressure of 700 psig cannot exceed the LTOP limits if the accumulators are injected. Isolating the RHR system from the RCS ensures that the RHR system is not subjected to accumulator pressure.

Depressurizing the accumulators below the LTOP limit from Figure 3.4.12-1 also gives'sthis protection. Additionally, the RHR System must be isolated fromnthe RCS to'protect RHR piping from a potential mass addition'event.

The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

(continued)

INDIAN POINT 3 B 3.4.12-13 Revision I

LTOP B 3.4.12 BASES ACTIONS D.1 (continued)

When any RCS cold leg temperature is < 319 OF, with one required PORV inoperable, the PORV must be restored to OPERABLE status within a Completion Time of 7 days. Two PORVs are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time considers the facts that only one of the PORVs is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.

E.1 When both required PORVs are inoperable or the Required Action and associated Completion Time of Condition C or D is not met, an alternate method of low temperature overpressure protection must be established within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The acceptable alternate methods of LTOP include the following:

a. Depressurize the RCS and establish an RCS vent path; or

b. Increase all RCS cold leg temperatures to > 3190F and isolate the RHR system from the RCS; or If the option selected is to depressurize the RCS and establish an RCS vent path, the vent must be sized > 2.00 square inches to ensure that the flow capacity is greater than that required for the worst case mass input transient reasonable during the applicable MODES. This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.

(continued)

INDIAN POINT 3 B 3.4.12 -14 Revision I

LTOP B 3.4.12 BASES ACTIONS F.1 (continued)

If LTOP requirements are not met for reasons other than Conditions A. B. C. D or E. LTOP requirements must be re-established by depressurizing the RCS and establishing an RCS vent of > 2.00 square inches within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS

SR 3.4.12.1 and SR 3.4.12.2 To minimize the potential for a low temperature overpressure event by limiting the mass input capability, all HHSI pumps are verified incapable of injecting into the RCS. Additionally, the accumulator discharge isolation valves are'verified closed and locked out or the accumulator pressure less than the maximum RCS pressure for the'existing RCS cold leg temperature allowed by the P/T limit curves provided in Figure 3.4.12-1.

The HHSI pumps are rendered incapable of injecting into the RCS through removing'the power'from the pumps by racking the breakers out under administrative control. Other methods may be employed using at least two'independent means to prevent a pump start such that a single failure or single action will not result in an injection into the RCS. This may be accomplished through the pump control switch'being placed in Trip Pullout and at least one valve in the discharge flow path being closed.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering other indications and alarms available to the operator in the control room, to verify the'required status of the equipment.

SR 3.4.12.3' The RCS vent of > 2.00 square inches is proven OPERABLE by verifying its'open 'condition either-'

a. Once every '2hodrsfor'a-valve that is not locked.

(continued)

INDIAN POINT 3 B 3.4.12 - 15 Revision I

LTOP B 3.4.12 BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.3 (continued)

b. Once every 31 days for a valve that is locked, sealed, or secured in position. A removed pressurizer safety valve, PORV, or Manway Cover fits this category.

The passive vent arrangement must only be open to be OPERABLE.

This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of the LCO 3.4.12.b.

SR 3.4.12.4 Performance of the CHANNEL CHECK of the Overpressure Protection System (OPS) RCS pressure and temperature channels every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 similar parameter 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 even 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.

The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal (continued)

INDIAN POINT 3 B 3.4.12- 16 Revision I

LTOP B 3.4.12 BASES SURVEILLANCE REOUIREMENTS SR 3.4.12.4 (continued) operational use of the displays associated with the LCO required channels. This SR is required only when LCO 3.4.12.a is used to establish LTOP protection.

SR 3.4.12.5 The PORV block valve opens automatically when RCS cold leg

-temperature is below the OPS arming temperature; however, the valves must be verified open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ' to provide the flow path for each required PORV to perform its function when actuated.

The valve may be remotely verified open in the control room. This Surveillance is performed only if the PORV is being used to satisfy LCO 3.4.12.a.

The block valve is a remotely controlled, motor operated valve.

The power to the-valve operator is not required removed, and the manual operator is not required locked in the inactive position.

Thus, the block valve-can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation. If closed, the block valve must be de-energized to prevent the valve from re-opening automatically.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is considered adequate because the PORV block valves are opened automatically by the OPS when below the OPS arming temperature if the valve control is positioned to auto and other administrative controls available to the operator in the control room, such as valve position indication, that verify that the PORV block valve remains open.

SR 3.4.12.6' Performance of a COT is required within 12.hours after decreasing all RCS teiperatures to < 319 'F and every'31'days on each required PORV to verify and, as necessary, adjust its'lift setpoint.

(continued)

INDIAN POINT 3 B 3.4.12 -17 Revision 1

- -. A-LTOP B 3.4.12 BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.6 (continued)

The COT will verify the setpoint is within the allowed maximum limits in Figure 3.4.12-1. PORV actuation could depressurize the RCS and is not required.

The 24 month Frequency considers the demonstrated reliability of the Overpressure Protection System and the PORVs.

A Note has been added indicating that this SR is required to be met 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to < 319 OF. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering the LTOP MODES.

SR 3.4.12.7 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required every 18 months. Performance of a CHANNEL CALIBRATION of RCS pressure and temperature instruments that support the Overpressure Protection System is required every 24 months. These calibrations verify both the OPS and PORV function and ensure the OPERABILITY of the whole channel so that it responds and the valve opens within the required range and accuracy to known input.

SR 3.4.12.8 and SR 3.4.12.9 The RCP starting prerequisites must be satisfied prior to starting or jogging any reactor coolant pump (RCP) when low temperature overpressure protection is required, The RCP starting prerequisites prevent an overpressure event due to thermal transients when an RCP is started. Plant conditions prior to the RCP start determines whether SR 3.4.12.8 or SR 3.4.12.9 must be satisfied prior to starting any RCP.

(continued)

INDIAN POINT 3 B 3.4.12 - 18 Revision I

LTOP B 3.4.12 BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.8 and SR 3.4.12.9 (continued)

The principal contributor to an RCP start induced thermal and pressure transient is the difference between RCS cold leg temperatures and secondary side water temperature of any SG prior o'the start 6f an RCP. The RCP starting'prerequisites vary depending on plant conditions but include the following: reactor coolant temperature'relative to the LTOP enable temperature; secondary side water temperature of the hottest SG relative to the temperature of the coldest RCS cold leg temperature; and, status of the Overpressure Protection System (OPS). When the OPS is inoperable, additional compensatory requirements are required including limits for the pressurizer level and RCS pressure and temperature. When a pressurizer level is specified as a requirement, the level specified is sufficient to prevent the RCS from going water solid for 10 minutes which is sufficient time for operator action to terminate the pressure transient.

SR 3.4.12.8 is used"if secondary side water temperature of the

'hottest steam generator (SG) is less than-or equal to the coldest RCS cold leg temperature. SR 3.4.12.9 is more restrictive and is used if the secondary side water temperature of the hottest steam generator is < 64 0 F-above the coldest RCS cold leg temperature.

RCP starting is prohibited if the hottest'steam generator is > 64 OF above RCS cold leg temperature or if neither of the RCP starting prerequisites SRs can be satisfied. The steam generator temperature may be measured using the Control Room instrumentation or. as a backup, from a contact reading off the steam generator's shells. Pressurizer-level may be determined using control room instrumentation or alternate methods.

The FREQUENCY of the RCP starting prerequisites SRs is within 15 minutes prior to starting any RCP. This means that each of the required verifications must be performed within 15 minutes prior to the pump start and must be met at the time of the pump start.

(continued)

INDIAN POINT 3 B 3.4.12-19 Revision 1.

- s LTOP B 3.4.12 BASES SURVEILLANCE REQUIREMENTS SR 3.4.12.8 and SR 3.4.12.9 (continued)

SR 3.4.12.8 and SR 3.4.12.9 are each modified by two Notes. Note 1 specifies that these SRs are required as a condition for pump starting only when the RCS is below the LTOP arming temperature.

Note 2 specifies that meeting either SR 3.4.12.8 or SR 3.4.12.9 ensures that pump starting prerequisites are met.

REFERENCES 1. 10 CFR 50, Appendix G.

2. Generic Letter 88-011, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations.

3. IP3 Low Temperature Overpressurization System Analysis Final Report, August 24, 1984, in conjunction with ASME Code Case N-514, Low Temperature Overpressure Protection, February 12, 1992.

4. IP3 Technical Requirements Manual.

5. 10 CFR 50, Section 50.46.

6. 10 CFR 50, Appendix K.

7. WCAP-16037, Revision 1, "Final Report on Pressure-Temperature Limits for Indian Point Unit 3 NPP", Westinghouse Electric Company, May 2003.

INDIAN POINT 3 B 3.4.12 -20 Revision 1

Accumulators B 3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Accumulators BASES BACKGROUND The functions of the' ECCS'accumulators are to supply water to the reactor vessel during the'blowdown phase of a loss of coolant accident (LOCA). to provide inventory'to'help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for any LOCA that reduces RCS pressure to below the accumulator pressure.

The blowdown phase of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heatfrom fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant.

The blowdown phas6'of the transient ends when the RCS pressure falls to a value 'approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam'flashing',and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available'to help fill voids in the lower plenum and reactor vessel downcomer so as to establish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water.

The accumulators 'are'pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive component's, since no operator-or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

Each accumulator is piped into an RCS cold leg via an accumulator line and is'is6lated from the RCS by a motor operated isolation valve and two check valves in series.,

(continued)

INDIAN POINT 3 B 3.5.1-1 Revision I

Accumul ators B 3.5.1 BASES BACKGROUND The accumulator size, water volume, and nitrogen cover pressure (continued) are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE SAFETY ANALYSES The accumulators are assumed OPERABLE in both the large and small break LOCA analyses at full power (Ref. 1). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS.

The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, (continued)

INDIAN POINT 3 B 3.5-1-2 Revision I

Accumul ators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued) the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

The worst case small break LOCA analyses also assume a time delay before pumped flow reaches-the core. For the larger range of small breaks,'the rateof blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then'.providing'continued cooling. As break size decreases, the accumulators and high head safety injection (HHSI) pumps both play a partlin.terminating'the rise in clad temperature. As break'size continues to decrease, the role of the accumulators continues to decrease until they are not required and the HHSI pumps become solely responsible for terminating the temperature increase.

This LCO helps'to ensure'that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 2) will be met following a LOCA:

a. Maximum fuel element cladding temperature is < 22000F:

b. Maximum cladding-oxidation is < 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is < 0.01 times the hypothetical amount that would be generated if all of the metal in'the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and

d. Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdown phase of a LOCA,'they do not'contribute to the long term cooling requirements of 10 CFR 50.46.

(continued)

INDIAN POINT 3 B 3.5.1-3 Revision I

Accumul ators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued)

For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged.

Accumulator tank size and accumulator water volume directly affect the volume of nitrogen cover gas whose expansion produces the passive injection and thus affects injection rate. The amount of water is also important since the accumulator water which has not been injected and bypassed during blowdown is primarily responsible for filling the lower plenum (refill) and downcomer.

The elevation head of the downcomer water provides the driving force for core reflooding (Ref. 3).

For large break LOCAs, changes in accumulator water volume can result in either improved or worsened analysis results; therefore, a nominal accumulator water volume of 795 cubic feet is modeled in the analysis (Ref. 3).

For small break LOCAs, changes in accumulator water volume are not significant because the clad temperature transient is terminated before the accumulators empty; therefore, a nominal accumulator water volume of 795 cubic feet is modeled in the analysis (Ref.

3).

The minimum boron concentration setpoint is used in the post LOCA boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH.

(continued)

INDIAN POINT 3 B 3.5.1 -4 Revision 1

Accumul ators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The large and small break'LOCA analyses are performed at the minimum nitrogen cover pressure, since sensitivity analyses have demonstrated that higher nitrogen cover pressure results in a computed peak clad temperature benefit. The maximum nitrogen cover pressure limit prevents injection of nitrogen into the RCS.

accumulator relief valve actuation, and ultimately preserves accumulator integrity.

The effects on'containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs.

3 and 4).

The accumulators satisfy Criterion 3 of-10 CFR 50.36.

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 2) could be violated.

For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 2000 psig. and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure > 1000 psig, the accumulator OPERABILITY requirements are based on full power' operation.. Although cooling requirements decrease as power decreases,the accumulators are still required to provide core cooling as long'as-elevated'RCS pressures'and temperatures exist.

This LCO is only'applicable at pressures > 1000 psig. At pressures .'1000 'psig, the'rate of RCS blowdown is such that the (continued)

INDIAN POINT 3 B .3.5.1 - 5 Revision I

i -

Accumulators B 3.5.1 BASES APPLICABILITY ECCS pumps can provide adequate injection to ensure that peak clad (continued) temperature remains below the 10 CFR 50.46 (Ref. 2) limit of 2200 0F.

In MODE 3, with RCS pressure < 1000 psig, and in MODES 4, 5, and 6, the accumulator motor operated discharge isolation valves are closed to isolate the accumulators from the RCS. This allows RCS cooldown and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.

Note 1 provides an exception to SR 3.5.1.1 and SR 3.5.1.5 and specifies that all accumulator discharge isolation valves may be closed and energized for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during the performance of reactor coolant system hydrostatic testing. This allowance is necessary because limits imposed by the Pressure/Temperature Limits for a hydrostatic leak test, could, in some instances, require reactor coolant system hydrostatic testing above 3500 F (Mode 3). This allowance is acceptable because hydrostatic testing is performed in MODE 3 when the need for the accumulators is reduced and Note 1 limits the duration to the time needed to perform required testing.

Note 2 also provides an exception to SR 3.5.1.1 and SR 3.5.1.5 and specifies that one accumulator discharge isolation valve may be closed and energized in MODE 3 for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for accumulator check valve leakage testing. This allowance is acceptable because testing is limited to MODE 3 when the need for the accumulators is reduced and Note 2 limits the duration to the time needed to perform required testing.

ACTIONS A.1 If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of (continued)

INDIAN POINT 3 B 3.5.1 -6 Revision I

Accumul ators B 3.5.1 BASES ACTIONS A.1 (continued) the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core-subcriticality during reflood. Boiling'of ECCS water in the'core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large'main steam line break. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to return the boron concentration to within limits.

B.1 If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In this Condition, the required contents of three accumulators cannot be assumed to reach the core during a LOCA. Due to the severity of the consequences should a LOCA occur in these conditions, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable'accumulator to OPERABLE status.

The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable accumulator to OPERABLE'status is justified in WCAP-15049-A, Rev. 1 (Ref. 4).

C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reactor coolant pressure reduced to'<'1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable,!based on operating experience, to reach'the required plant-conditibns fr6m full power conditions in an orderly manner and without'challenging plant systems.

(continued)

INDIAN POINT 3 B 3.5.1 -7 Revision 1

Accumulators B 3.5.1 BASES ACTIONS D.1 (continued)

If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered imrnediately.

SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator valve should be verified to be fully open every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If a discharge isolation valve is not fully open, the rate of injection to the RCS would be reduced.

Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions. This Frequency is considered reasonable in view of other administrative controls that ensure a mispositioned isolation valve is unlikely.

SR 3.5.1.2 and SR 3.5.1.3 Every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , borated water volume and nitrogen cover pressure are verified for each accumulator. This Frequency is sufficient to ensure adequate injection during a LOCA. Because of the static design of the accumulator, a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency usually allows the operator to identify changes before limits are reached. Operating experience has shown this Frequency to be appropriate for early detection and correction of off normal trends.

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator every 31 days since the static design of the accumulators limits the ways in which the concentration can be changed. The 31 day Frequency is adequate to identify changes that could occur from mechanisms such as stratification or (continued)

Revision 1 INDIAN POINT 3 INDIAN POINT 3 B 3.5.1-8 B 3.5.1 -8 Revision 1

Accumulators B 3.5.1 BASES SURVEILLANCE REOUIREMENTS SR 3.5.1.4 (continued) inleakage. Sampling the affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after an increase of 8.4 cubic feet will identify whether inleakage has caused a reduction in boron concentration to below the required limit. Considering the nominal accumulator volume of 795 cubic feet of water, inleakage of 8.4 cubic feet of pure water would result in a boron concentration reduction of approximately 1%. An increase in the accumulator volume of 8.4 cubic feet causes a change of approximately 1O% in the indicated accumulator level.

It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST),

because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 5).

SR 3.5.1.5 Verification every 31 days that power is removed from each accumulator discharge isolation valve operator when the reactor coolant system pressure is > 2000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Since power is removed under administrative control, the 31 day Frequency will provide adequate assurance that power is removed.

This SR allows power to be supplied to the motor operated discharge isolation valves when reactor coolant system pressure is

< 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Should closure of a valve occur, the SI signal provided to the valves would open a closed valve in the event of a LOCA.

(continued)

INDIAN POINT 3 B 3.5.1-9 Revision I

Accumulators B 3.5.1 BASES REFERENCES 1. FSAR, Chapter 6.

2. 10 CFR 50.46.

3. FSAR, Chapter 14.

4. WCAP-15049-A, Rev. 1. April 1999.

5. NUREG-1366, February 1990.

INDIAN POINT 3 B 3.5.1 -10 Revision 1

LOP DG Start Instrumentation B 3.3.5 B 3.3 INSTRUMENTATION B 3.3.5 Loss of Power (LOP) Diesel Generator CMG) Start Instrumentation BASES ' ' - ' '

BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or is insufficiently stable to allow safe unit operation. Undervoltage protection will generate a DG start if a loss of voltage or degraded voltage condition occurs on a 480 V bus.

Two unde'rvoltage relays are provided on each 480 V bus for

'detecting a bus undervoltage. Either'of the two relays is sufficient to satisfy requirements for the 480 V bus undervoltage Function even though the failure of the one remaining undervoltage relay could result in the failure of one DG to start because there is redundancy in the number of EDGs available. The two undervoltage relays are combined in a one'-out-of-two logic per bus to generate an undervdltage signal.' The allowable value and trip setpoint for this function is established in accordance with Reference 3. Actuation of these relays will trip the bus supply breaker, initiate load shedding, start the DG. and initiate load sequencing. There is no explicit time delay for this function because the undervoltage protection devices are induction type disc relays. Therefore, the time to actual trip will decrease as a function of voltage decrease below the'setpoint.

Two degraded voltage relays are provided 'on each 480 V bus for detecting degraded bus voltage. The relays are combined in a two-out-of-two logic per bus (to prevent spurious actuation). The allowable value and trip'setpoint for'this'function is established in accordance wiith Reference 3. Function actuation includes a time delay 'of '10 seconds if a coincident SI signal indicates accident conditions'exist and a'time delay of 45 seconds if no SI signal is generated (i.e., non-accident condition). These time delays ensure proper coordination with plant electrical transients (e.g. large motor starts, fast transfers, etc.). Actuation of these relays will trip the bus supply breaker, which will in turn actuate the undervoltage relays.

(continued)

INDIAN POINT 3 B 3.3.5 -1 Revision I

LOP DG Start Instrumentation B 3.3.5 BASES BACKGROUND (continued) The LOP start actuation is described in FSAR, Section 8.2 (Ref.

1).

Trip Setpoints and Allowable Values Technical Specification Allowable Values are determined based on the relationship between an analytical limit and a calculated trip setpoint. A detailed discussion of the relative position of the safety limit, analytical limit, allowable value and the trip setpoint with respect to the normal plant operation point is presented in the Bases of LCO 3.3.1, Reactor Protection System (RPS) Instrumentation.

A detailed description of the methodology used to calculate the channel Allowable and bistable device, including their explicit uncertainties, is provided in Engineering Standards Manual IES-3 and IES-3B, Instrument Loop Accuracy and Setpoint Calculation Methodology (IP3) (Ref. 3).

APPLICABLE SAFETY ANALYSES The LOP DG start instrumentation is required for the Engineered Safety Features (ESF) Systems to function in any accident with a loss of offsite power. Its design basis is that of the ESF Actuation System (ESFAS).

Accident analyses credit the loading of the DG based on the loss of offsite power during a loss of coolant accident (LOCA). The actual DG start has historically been associated with the ESFAS actuation. The DG loading has been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.

The required channels of LOP DG start instrumentation, in conjunction with the ESF systems powered from the DGs, provide unit protection in the event of any of the analyzed accidents discussed in Reference 2, in which a loss of offsite power is assumed.

(continued)

INDIAN POINT 3 B 3.3.5 -2 Revision I

LOP DG Start Instrumentation B 3.3.5 BASES APPLICABLE SAFETY ANALYSES (continued)

The delay times assumed in the' safety analysis for the ESF equipment include the 10 'second DG start delay, and the appropriate sequencing delay. The response times for ESFAS actuated equipment in LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," include the appropriate DG loading and sequencing delay.

The LOP DG start instrumentation channels satisfy Criterion 3 of 10 CFR 50.36.

LCO The LCO for LOP DG start instrumentation requires that 1 channel per bus of the undervoltage (480 V b6s) Function and two channels per bus of the Degraded Voltage (480 V'bus) Function must be OPERABLE in MODES 1. 2, 3 and 4 when the LOP DG'start instrumentation supports safety systems associated with the ESFAS.

In MODES 5 and 6, 1 channel per bus of the unde'rvoltage (480 V bus) Function and two channels per'bus' of the Degraded Voltage (480 V bus) Function must be OPERABLE whenever the associated DG is required to be OPERABLE to ensure that the automatic start of the DG is available vwhen'needed.

APPLICABILITY The LOP DG Start Instrumentation Functions are required in MODES 1, 2, 3, and 4 because ESF Functions are designed to provide protection in these MODES. Actuation in MODE,5 or 6 is required whenever the required DG must be OPERABLE so that it can perform its function on an LOP or.degradedipowerto the vital bus.

ACTIONS In the event a channel's Trip Setpoint is found nonconservative with respect to the Allowable Value, or the channel is found

-inoperable, then the function that channel provides must be declared inoperable,and-the. LCO Condition entered for the particular protection function affected.

(continued)

INDIAN POINT 3 B 3.3.5 -3 Revision I

LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS Because the required channels are specified on a per bus (continued) basis, the Condition may be entered separately for each bus as appropriate. A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in the LCO. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A.1 Condition A applies to the LOP DG start Function with one required channel of the undervoltage function inoperable. Note that LCO 3.3.5 requires that only one of the two undervoltage (480 V bus) channels must be OPERABLE. Therefore, Condition A applies when there is no OPERABLE undervoltage (480 V bus) channel on one or more 480 volt vital bus(es).

If one required channel is inoperable or one or more 480 V buses, Required Action A.1 requires that channel to be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The specified Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore an undervoltage (480 V bus) channels to OPERABLE status is needed because this Condition represents a loss of the undervoltage DG starting Function for the associated DG. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months delay in declaring the DG inoperable is acceptable because of the low probability of an event occurring during this interval.

B.1 Condition B applies when one of the two required degraded voltage channels is inoperable on one or more 480 V bus. Required Action B.1 requires placing the inoperable channel in trip so that trip capability is restored to the 2 out of 2 logic used to initiate this Function. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time takes into account the low probability of an event requiring an LOP start occurring during this interval.

(continued)

INDIAN POINT 3 B 3.3.5 -4 Revision 1

LOP DG Start Instrumentation B 3.3.5 K>y BASES ACTIONS C.1 (continued)

Condition C applies to each of the LOP DG start Functions'when the Required Action and associated Completion Time for Condition A or B are not met. Condition C also applies when two channels of Degraded Voltage Function inoperable in one or more buses. In this Condition, Function trip capability is lost even if one of the channels is placed in trip as specified in Required Action B.1.

In these circumstances the Conditions :specified in LCO 3.8.1, "AC Sources-Operating." or LCO 3.8.2, "AC Sources-Shutdown." for the

-DG made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately. The-actions of those LCOs provide for adequate compensatory actions to-assure unit safety.

SURVEILLANCE REOUIREMENTS SR 3.3.5.1

'- 'SR 3.3.5.1 is the performance of a TADOT. This test is performed every 31 days. The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment. The Frequency is based on the known reliability of the relays and controls and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

This SR excludes verification of setpoints from the TADOT. Since this TADOT applies to 480 V degraded voltage and undervoltage, setpoint verification requires bench calibration and is accomplished during CHANNEL CALIBRATION. Although the SR is not modified by a note, this&is' a non-conservative SR whose intent was never to. require pulling relays for bench testing. The 480 Volt Bus degraded voltage is sensed by two' (2) undervoltage relays per bus. A trip signal requires both relays to sense the degraded voltage condition so pulling a relay makes EDGs inoperable. NRC Administrative Letter 98-10 requires non-conservative Technical Specification requirements to be treated as a nonconforming (continued)

INDIAN POINT 3 B 3.3.5 -5 Revision I

LOP OG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continuted) condition under Generic Letter 91-18 with administrative controls (i.e., the clarification in this Basis) in place until a change to the Technical Specification is processed.

SR 3.3.5.2 SR 3.3.5.2 is the performance of a CHANNEL CALIBRATION.

The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay, as applicable.

A CHANNEL CALIBRATION is performed every 24 months for the undervoltage relay and every 18 months for the degraded voltage relay. 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.

The Frequency is based on operating experience and is justified by the assumption of the calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis (Ref.

3).

REFERENCES 1. FSAR, Section 8.2.

2. FSAR, Chapter 14.2.

3. Engineering Standards Manual IES-3 and IES-3B, Instrument Loop Accuracy and Setpoint Calculation Methodology (1P3).

INDIAN POINT 3 B 3.3.5 -6 Revision I

^ IPEC SITE QUALITY RELATED IP-SMM-AD-103 Revision 0 Z-Enteergy 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 Pace I of I

-z-Entergy CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES TO:.DISTRIBUTION DATE: 1012412004 TRANSMITTAL NO:

(Circle one)

FROM: IPEC DOCUMENT CONTROL: EEC or IP2 53'EL PHONE NUMBER: 271-7057 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.

THE FOLLOWING IS A TMI UPDATE. INCORPORATE INTO YOUR FILES:

DATED 10/28/04

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

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

NAME (PRINT) SIGNATURE DATE CC# 67 b<

UPDATE FOR IP3 TECHNICAL REQUIREMENTS MANUAL AFFECTED SECTION REMOVE INSERT List of Effective Sections Page 1 of 1 with Page 1 of 1 with Effective date 08/24/2004 Effective date 10/28/2004 Section 3.3-B Revision 2 Revision 3 Pages 3.3.B-1 through Pages 3.3.B-1 through 3.3.B-12 3.3.B-12

LIST OF EFFECTIVE SECTIONS TRM _ EFFECTIVE TRM EFFECTIVE SECTION ; Page(s) DATE SECTION _ Page(s) DATE Table of 2 i through iii 12/04/2002 3.8.C 1 3.8.C-1 through 10 08/24/2004 Contents _ 3.8.D 0 3.8.D-1 through 2 03/19/2001 1.1 2 1.1-1 through 5 02/23/2004 3.9 0 3.9-1 03/19/2001 1.2 0 1.2-1 through 3 03/19/2001 4.0 0 4.0-1 03/19/2001 1.3 0 1.3-1 through 8 03/19/2001 5.0 4 5.0-1 through 7 08/24/2004 1.4 0 1.4-1 through 4 03/19/2001 2.0 0 2.0-1 03/1912001 3.0 1 3.0-1 through 15 07/06/2001 3.1.A 1 3.1.A-1 through 8 07/06/2001 3.1.8 0 3.1.B-1 03/19/2001 3.1.C.1 1 3.1.C.1-1 through 8 03/06/2003 3.1.C.2 1 3.1.C.2-1 through 6 03/06/2003 3.2.A 0 3.2.A-1 03/19/2001 3.3.A 1 3.3.A-1 through 3 08/24/2004 3.3.8 3 3.3.B-1 through 12 10/28/2004 I 3.3.C 0 3.3.C-1 through 5 03/19/2001 3.3.D 2 3.3.D-1 through 20 09/03/2003 3.3.E 1 3.3.E-1 through 3 08/24/2004 3.3.F 1 3.3.F-1 through 3 08/24/2004 3.3.G 0 3.3.G-1 through 2 03/19/2001 3.3.H 1 3.3.H-1 through 2 08/24/2004 3.3.1 ----- NOT USED -----

3.3.J 1 3.3.J.1 through 5 04/16/2003 3.4.A 0 3.4.A-1 through 2 03/19/2001 3.4.B 0 3.4.8-1 through 3 _03/19/2001 3.4.C 0 3.4.C-1 through 2 03/19/2001 3.4.D 0 3.4.D-1 through 2 03/19/2001 3.5.A 0 3.5.A-1 through 2 03/19/2001 3.6 0 3.6-1 03/19/2001 3.7.A.1 1 3.7.A.1-1 through 5 08/24/2004 3.7.A.2 2 3.7.A.2-1 through 3 08/24/2004 3.7.A.3 5 3.7.A.3-1 through 6 08/24/2004 3.7.A.4 3 3.7.A.4-1 through 3 08/24/2004 3.7.A.5 1 3.7.A.5-1 through 3 08/24/2004 3.7.A.6 1 3.7.A.6-1 through 2 08/24/2004 3.7.A.7 2 3.7.A.7-1 through 4 08/24/2004 3.7.B 2 3.7.B-1 through 17 08/24/2004 3.7.C 0 3.7.C-1 through 8 03/19/2001 3.7.D 0 3.7.D-1 through 2 03/19/2001 3.7.E 0 3.7.E-1 through 2 03/19/2001 3.8.A 0 3.8.A-1 through 5 03/19/2001 3.8.B 0 3.8.8-1 through 7 03/19/2001 Technical Requirements Manual (TRM)

Indian Point 3 1 of 1 EFFECTIVE DATE 10/28/2004

Meteorological Monitoring Instrumentation 3.3.8 3.3 INSTRUMENTATION 3.3.B Meteorological Monitoring Instrumentation TRO 3.3.B The Meteorological Monitoring Instrument Channel per Table 3.3.8-1 shall be OPERABLE.

APPLICABILITY: At all times.

__ ----------------- NOT E - ------------------ - ---- - --- -- - - -----------------------

1. TRO 3.0.C is not applicable.

2. TRO 3.0.D is not applicable.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The A.1 DEMONSTRATE the ability to obtain 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Meteorological meteorological data, using IP-EP-510, Monitoring Instrument AND Channel is inoperable. ----------------------NOTE-------------------

Action A.2 is NOT required when IP3 control room meteorological display and/or strip chart recorder are the only inoperable equipment.

A.2 Notify IP2 of system inoperability, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND A.3 Restore the inoperable Meteorological 7 days Instrument Channel to OPERABLE status.

B. Required Actions 8.1 Prepare and submit a Special Report to the 10 days and associated On-Site Safety Review Committee outlining Completion the actions taken, the cause of the Times of inoperability and the plans for restoring the Condition A.3 not meteorological monitoring instrumentation met. channel(s) to OPERABLE status.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.B-1 Revision 3

Meteorological Monitoring Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE---------

Control Room display on the back of the Flight Panel and the Meteorological Strip Chart Recorder are not required to meet the TRO.

Perform CHANNEL CHECK.

TRS 3.3.B.1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

NOTE-------------.-----------

This surveillance is not required to be performed to meet the TRO.

TRS 3.3.B.2 Perform calibration of meteorological strip chart . 24 months recorder. .

NOTE-------------------------

This surveillance is not required to be performed to meet the TRO when primary power source is available.

TRS 3.3.B.3 - ------ 31 days DEMONSTRATE Meteorological Diesel Generator OPERABILITY by starting and running for 15 minutes.

NOTE--------------------------

This surveillance is not required to be performed to meet the TRO when primary power source is available.

TRS 3.3.8.4 --------- - 12 months DEMONSTRATE Diesel Generator Automatic Power Transfer by simulating power loss.

TRS 3.3.B.5 Perform CHANNEL CALIBRATION. 184 days TRS 3.3.B.6 Perform CHANNEL OPERATIONAL TEST. 184 days Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.13-2 Revision 3

Meteorological Monitoring Instrumentation 3.3.B TABLE 3.3.B-1 Meteorological Monitoring Instrumentation Channels Instrument Channels Instrument Channel Minimum Operable Minimum Accuracies Channels

1. WIND SPEEDl A. 10m + 0.5 mph 1

2. WIND DIRECTION 1 A. 1m +50 1

3. ATMOSPHERIC STABILITY (PASQUILL CATEGORY) 2 A. 60- 10m + 0.1°C 1 for temperature inputs Note 1 The 60m and 122m level instruments are not required to meet the TRO but are maintained to support Indian Point 2 requirements.

Note 2 The 122-1Om delta temperature instruments are not required to meet the TRO but are maintained to support Indian Point 2 requirements.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.13-3 Revision 3

Meteorological Monitoring Instrumentation 3.3.8 BASES BACKGROUND The meteorological monitoring instrumentation system was installed to meet the requirements, in part, of 10 CFR 50 Appendix A (Reference 1),

10 CFR 50 Appendix E (Reference 2), and 10 CFR 50.47(b)(9)

(Reference 3). These sections require that adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency be available.

Guidance on the meteorological monitoring requirements is provided in NUREG-0737 (Reference 4), NUREG-0654 (Reference 5), Regulatory Guide 1.23 (Reference 6), and Regulatory Guide 1.97 (Reference 7).

NUREG-0737 required that each nuclear facility "upgrade its emergency plans to provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Specific criteria to meet this requirement is delineated in NUREG-0654."

NUREG-0737 also provided a schedule of implementation milestones to be met in order to address the introduction of NUREG-0654, Appendix 2.

Letter IPN-80-117 (Reference 8) addressed each item of NUREG-0737 that was applicable to Indian Point 3 (IP3) and which had not been previously identified as complete. IP3 agreed to the staged implementation schedule required by the NUREG in this letter.

NUREG-0654 was issued, in part, to provide a basis for the development of radiological emergency plans and the improvement of emergency preparedness. Appendix 2 of NUREG-0654 states that "the emergency facilities and equipment as stated in Appendix E to 10 CFR Part 50 shall include '(E)quipment for determining the magnitude of and for continuously assessing the impact of the release of radioactive materials to the environment.' To address this requirement, in part, the nuclear power plant operator shall have meteorological measurements from primary and backup systems. Each site ... shall have a primary meteorological measurements system. The primary system shall produce current and record historical local meteorological data ... The acceptance criteria for meteorological measurements are described in the proposed Revision 1 to U.S. NRC Regulatory Guide 1.23."

Regulatory Guide (RG) 1.23 provides information on meteorological instrument accuracy and meteorological instrument maintenance and servicing schedules. The meteorological instrument accuracies are listed in Table 3.3.B-1. The guidance from RG 1.23 section C.4 and C.5 on meteorological maintenance and servicing schedules is reflected in the "Surveillance Requirements" section of this Technical Requirement.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.B-4 Revision 3

Meteorological Monitoring Instrumentation 3.3.B RG 1.97 describes a method for complying with the NRC's regulations to provide instrumentation to monitor, display and record plant variables and systems during and following an accident. Table 3 of the RG lists meteorological variables and the minimum ranges these variables should operate within. In addition, RG 1.97 stated that information gathered by these parameters "may be continually updated, stored in computer memory, and displayed on demand. Intermittent displays such as data loggers and scanning recorders may be used if no significant transient response information is likely to be lost by such a device."

The NRC issued a Confirmatory Order (Reference 9), requiring that IP3 perform certain additional actions to increase the margin of public health and safety. Included in the Order were a number of interim measures that pertained to the meteorological program and to Control Room instrumentation. Annex 1 to the Order laid out the meteorological acceptance criteria for emergency preparedness.. The Annex essentially described the meteorological program as found in NUREG-0654 and added additional acceptance criteria from NUREG-75/087 section 2.3.3 (Reference 10).

NUREG-75/087, section 2.3.3 states that "Generally, the onsite meteorological programs must produce data which can be summarized to provide an adequate meteorological description of the site and its vicinity for the purpose of making atmospheric diffusion estimates for accidental and routine airborne releases of effluents. Guidance on an adequate program is given in Regulatory Guide 1.23."

IP3's response to the Confirmatory Order, letter IPN-80-77 (Reference 11), was to perform a detailed review of the meteorological program. The results of the review were that IP3 and IP2 complied with the Annex 1 meteorological criteria.

The NRC issued Generic Letter (GL) 82-33 (Reference 12) as a supplement to NUREG-0737. One purpose of the letter was to provide additional clarification regarding the application of RG 1.97 to emergency response facilities. In addition, the letter required licensees to evaluate how their post-accident monitoring instrumentation in the Control Room met the content of RG 1.97. Letter IPN-86-05 (Reference 13) outlined the status of IP3's compliance with RG 1.97 (e.g., the actual ranges that the meteorological variables should operate in and IP3's compliance with the requirements for data recording). The letter indicated that IP3 met the data recording requirements and also included the actual variable ranges used by the plant.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.8-5 Revision 3

Meteorological Monitoring Instrumentation 3.3.8 The meteorological variable ranges required by the RG are as follows:

Wind Direction required: 0 to 3600 Wind Speed required: 0 to 50 mph Atmospheric *required: -5 to 10C Stability (for Temperature inputs)

Note: The actual range (-4.44 to I 10C) was deemed acceptable.

NRC Inspection Report 85-17 (Reference 14) documented a conversation between the NRC and IP3. During the conversation, the NRC stated that "Unit 2 technical specifications require that meteorological monitoring instrumentation channels be operable at all times with indication of the tabulated parameters available in the control room." As a result, the Authority stated that a method would be instituted to verify the readouts in the control room as well as at the meteorological tower. NRC Inspection Report No. 87-23 (Reference 15) closed this unresolved item. In this Inspection Report, the NRC stated, 'The licensee has installed a meteorological tower display in the control room demand metering panel.

The panel displays wind speed, wind direction, Pasquill category and the time of the last data update. The inspector reviewed Nuclear Safety Evaluation 87-03-049 INST, Rev. 0 for the modification."

In 1991, the NRC issued a Safety Evaluation (Reference 16) which re-evaluated IP3's conformance to RG 1.97. The evaluation was performed as a follow-up to determine if and how we were conforming to the contents of GL 82-33. Contained in this evaluation was the NRC's conclusion that "... the licensee (IP-3) has provided an explicit commitment on conformance to RG 1.97."

NRC Inspection Report 92-17 (Reference 17) documented an inspection involving IP3's Radiological Environmental Monitoring Program. The purpose of the inspection, in part, was to review the "meteorological monitoring program to determine whether the instrumentation and equipment were operable, calibrated and maintained in accord with licensee's requirements ... Based on the review of the program and discussions with the licensee's representatives, the inspector determined that overall the licensee has implemented an effective Meteorological Monitoring Program."

In addition to the above NRC commitments, IP3 will comply with the requirements of other outside agencies. These agencies include the Federal Aviation Administration, Environmental Protection Agency, etc.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.1-6 Revision 3

Meteorological Monitoring Instrumentation 3.3.B APPLICABLE The meteorological system is described in FSAR chapter 2.6 (Reference SAFETY 18), Emergency Plan Procedure, IP-EP-l10, "Meteorological, ANALYSES Radiological & Plant Data Acquisition Systems (Reference 19), and Nuclear Safety Evaluation 87-03-049 INST (Reference 20). The meteorological measurements program consists of primary and backup systems. The primary system consists of a 122m instrumented tower which provides measurements for wind speed and wind direction at a minimum of two levels, one of which is representative of the 10 meter level. Data obtained from the 1Om elevation of the meteorological tower is transmitted through a computer system to a meteorological LED display panel in the Control Room. IP3 maintains responsibility of the Meteorological Monitoring Program, except for the Meteorological Computer System, which is the responsibility of IP2. The meteorological tower display indicates wind speed, wind direction, Pasquill Category and the time of the last update. The output to the LED display panel is the result of a fifteen minute average of computated data from the Meteorological Computer System. The LEDs are updated every fifteen minutes. Also located in the control room is a two-pen variable trend recorder (strip chart) which is used to tre'nd wind speed and wind direction. The data displayed represents a 15-minute average.

In the event of a power outage, a diesel generator has been installed to provide immediate power to the meteorological tower system.

In the event of a failure of the primary meteorological measurement system, a backup meteorological system is used. Changeover from the primary system to the backup system occurs automatically.

This system is independent of the primary system and consists of two instrumented meteorological towers, a primary backup tower and a standby backup tower. The backup meteorological tower records wind direction and speed measurements at the 1Om level. The backup system provides information in the real-time mode. In the event of primary power failure, power is supplied for six days by a battery located adjacent to the tower. In the event of a failure of the backup meteorological measurement system, changeover from the backup system to the standby system is accomplished manually.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.1B-7 Revision 3

Meteorological Monitoring Instrumentation 3.3.B TRO The Meteorological Monitoring Instrument Channel must be OPERABLE to allow adequate assessing, monitoring and recording of actual or potential offsite consequences of a radiological emergency.

An OPERABLE Meteorological Monitoring Instrument Channel constitutes the following:

1.. Instrumentation on the primary meteorological tower for providing -

wind direction and speed measurement, representative of the 1Om level per Table 3.3.B-1, shall be OPERABLE.

2. The Meteorological Computer System shall be OPERABLE.

3. Power supply is available. A power supply must be available from the normal power supply or the meteorological diesel generator.

APPLICABILITY The Meteorological Monitoring Instrumentation Channel are required to be OPERABLE at all times.

ACTIONS A.1 The meteorological monitoring instrumentation was installed to meet the requirements of NUREG-0737 Section III.A.2.2. The operation of this equipment is also described in the IPEC Emergency Plan, stating that the Meteorological Monitoring Instrumentation Channel meets the requirements for indication and remote access. The channel is required in order to comply with the requirements of RG 1.97 which requires 'the instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand. Signals from meteorology monitors should be recorded. For recording, it may be continuously updated, stored in computer memory and displayed on demand."

A Meteorological Monitoring Instrument Channel would be required for determining the magnitude if and for continuously assessing the impact of the release of radioactive materials to the environment.

With the meteorological monitoring instrumentation channel inoperable, the backup meteorological monitoring instrumentation channel(s) must be DEMONSTRATED OPERABLE within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . DEMONSTRATION shall be achieved using Emergency Plan Procedure IP-EP-510, which describes the means to obtain meteorological data.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.B-8 Revision 3

Meteorological Monitoring Instrumentation 3.3.B A.2 With the meteorological monitoring instrumentation channel inoperable, IP2 shall be notified within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This notification is not required for IP3 control room display and/or recorder inoperability as this equipment does not directly impact IP2.

A.3 With the meteorological monitoring instrumentation channel inoperable, the channel must be restored to OPERABLE status within 7 days. The meteorological monitoring instrumentation channel(s) would be required in the event of a radiological emergency.

The allowable outage time (AOT) of 7 days, which is specified by this Action, was developed, in part, by taking into consideration former Westinghouse Standard Technical Specifications section 3.3.3.4 (Reference 21) which specified a 7 day time frame. In addition, consideration was given to IP2's Technical Requirements Manual section 3.3.A (Reference 22) which also specifies an AOT of 7 days.

B.1 This Action shall be taken if the Required Actions and associated Completion Times of Condition A have not been met. A Special Report shall be prepared and submitted to the On-Site Safety Review Committee outlining the cause of the malfunction and the plans for restoring the meteorological monitoring instrumentation channel(s) to OPERABLE status. This reporting is necessary to ensure oversight for restoring the OPERABILITY of the Meteorological Monitoring Instrument Channel and the collection of meteorological data at the plant site. This data is used for estimating potential radiation doses to the public resulting from routine or accidental releases of radioactive materials to the atmosphere.

A meteorological data collection program, as described in this technical requirement, is necessary to meet the requirements of 10 CFR 50.36a(a)(2), Appendix E to 10 CFR 50 and 10 CFR 51.

The ten-day period for preparing and submitting the Special Report was developed by taking into consideration IP2 Technical Requirements Manual section 3.3.A. This section requires that with one or more of the required meteorological monitoring channels inoperable for more than seven (7) days, prepare and submit to the On-Site Safety Review Committee within the next 10 days ... a Corrective Action Report ...

outlining the cause of the malfunction(s) and the plans for restoring the channel(s) to operable status.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.8-9 Revision 3

Meteorological Monitoring Instrumentation 3.3.B SURVEILLANCE TRS 3.3.8.1 REQUIREMENTS The performance of daily CHANNEL CHECKs is required to meet a commitment to the NRC. IP3 committed to daily CHANNEL CHECKs via a telephone conversation with the NRC (on August 12, 1985). The NRC acknowledged this verbal commitment in Inspection Report 85-17.

Inspection Report 85-17 documented the conversation in which the NRC stated that Indian Point Unit 2 Technical Specifications (now Technical Requirements Manual) contain the requirement that "meteorological monitoring instrumentation channels be operable at all times with indication of the tabulated parameters available in the control room.

Furthermore, the IP2 Technical Specifications also require a daily CHANNEL CHECK of the meteorological monitoring instrumentation and states that 'each meteorological monitoring channel shall be demonstrated operable' (T.S. 4.19.A)." As a result, IP3 agreed that the IP3 control room instrumentation should be DEMONSTRATED OPERABLE by a daily CHANNEL CHECK.

TRS 3.3.B.2 Based on engineering judgement, IP3 has concluded that the 24 month calibration interval of the meteorological strip chart recorder is adequate.

TRS 3.3.B.3 Based on engineering judgement, 1P3 has concluded that monthly testing is adequate to demonstrate the OPERABILITY of the meteorological diesel generator.

TRS 3.3.84 Based on engineering judgement, IP3 has concluded that annual testing is adequate to DEMONSTRATE diesel generator automatic power transfer.

TRS 3.3.8.5 The performance of semiannual instrument CHANNEL CALIBRATION is required to satisfy RG 1.23 section C.5. Compliance with RG 1.23 section C.5 is required per the NRC's February 11, 1980 Confirmatory Order.

Section C.5 stated that meteorological "instruments should be calibrated at least semiannually." In addition, this calibration frequency is consistent with TRS 3.3.A.1 and TRS 3.3.A.2 of IP2's Technical Requirements Manual.

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.1-10 Revision 3

Meteorological Monitoring Instrumentation 3.3.B TRS 3.3.B.6 The performance of semiannual instrument CHANNEL OPERATIONAL TEST ensures the signal is being delivered through the instrument channel. The frequency is chosen to be consistent with the frequency for instrument CHANNEL CALIBRATION.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50 Appendix A, Criterion 64, "Monitoring Radioactivity Releases."

2. Title 10, Code of Federal Regulations, Part 50 Appendix E, Section E, "Emergency Facilities and Equipment."

3. Title 10, Code of Federal Regulations, Part 50.47, "Emergency Plans."

4. NUREG-0737, "Clarification of TMI Action Plans Requirements."

5. NUREG-0654/FEMA, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Appendix 2, "Meteorological Criteria for Emergency Preparedness at Operating Nuclear Power Plants."

6. Regulatory Guide 1.23, "Onsite Meteorological Programs."

7. NRC Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."

8. NYPA Letter IPN-80-117, J. P. Bayne to D. G. Eisenhut, dated December 30,1980, "Post TMI Requirements."

9. NRC Confirmatory Order, H. R. Denton to E. R. Weiss, dated February 11, 1980.

10. NUREG-75/087, "Standard Review Plan."

11. NYPA Letter IPN-80-77, G. M. Wilverding to S. A. Varga, dated August 11, 1980, "Confirmatory Order (Interim Actions) Six Month Responses."

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.B-11 Revision 3

Meteorological Monitoring Instrumentation 3.3.B

12. Generic Letter 82-33, dated December 17, 1982, "Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability."

13. NYPA Letter IPN-86-05, J. C. Brons to S. A. Varga, dated January 7,1986, "Regulatory Guide 1.97 Implementation Program."

14. NRC Inspection Report No. 50-286/85-17, Section 7.0, T. T. Martin to W. Josiger, dated August 22,1985, "Implementation of the Meteorological Monitoring Program."

15. NRC Inspection Report No. 50-286/87-23, E. C. Wenzinger to W.

Josiger, dated October 15,1987.

16. NRC Safety Evaluation, J. D. Neighbors to R. E. Beedle, dated April 3, 1991, "Emergency Response Capability - Conformance to Regulatory Guide 1.97, Revision 3, for Indian Point 3."

17. NRC Inspection Report No. 50-286/92-17, J. H. Joyner to J. E.

Russell, dated July 18, 1992.

18. Indian Point 3 FSAR, Section 2.6.5, "Onsite MeteorologicaI Measurements Program."

19. Emergency Plan Procedure, IP-EP-510, "Meteorological, Radiological & Plant Data Acquisition System."

20. Nuclear Safety Evaluation NSE 87-03-049 INST, "Control Room Meteorological Display Upgrade."

21. NUREG-1431, Westinghouse Standard Technical Specifications section 3.3.3.4, "Meteorological Instrumentation."

22. Unit 2 Technical Requirements Manual Section 3.3.B

'Meteorological Monitoring."

Technical Requirements Manual (TRM) (continued)

Indian Point 3 3.3.B-12 Revision 3

ML043150420

Text

Navigation menu

Search