ML050960059

Distribution of IP3 Technical Specification Amendment 224 and 225

ML050960059
Person / Time
Site:	Indian Point
Issue date:	03/25/2005
From:	Entergy Nuclear Operations
To:	Office of Nuclear Reactor Regulation
References
Download: ML050960059 (40)

Text

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

=Enteigy MANAGEMENT ADMINISTRATIVE PROCEDURE MANUAL _ _ _ _ _ _ _ _ _ _

INFORMATIONAL USE Page 13 of 21 ATTACHMENT 10.1 SMM CONTROLLED DOCUMENT TRANSMITTAL FORM SITE MANAGEMENT MANUAL CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES Page 1 of 1 Entew CONTROLLED DOCUMENT TRANSMITTAL FORM - PROCEDURES TO: DISTRIBUTION DATE: 3125105 PHONE NUMBER: 271-7057 FROM: IPEC DOCUMENT CONTROL The Document(s) identified below are forwarded for use. In accordance with IP-SMM-AD-103, please review to verify receipt, incorporate the document(s) into your controlled document file, properly disposition superseded, void, or inactive documents). Sign and return the receipt acknowledgement below within fifteen (15) working days.

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• • • • • • • • • • • PLEASE NOTE EFFECTIVE DATE***********

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24-MAR-05 Page: 1 DISTRIBUTION CONTROL LIST Document 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-DO01/6(U3/IPEC) IP3(ONLY) 11 RES DEPARTMENT MANAGER RES (UNIT 3/IPEC ONLY) 45-4-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 55 GSB TECHNICAL LIBRARY A MCCALLION/IPEC & IP3 GSB-3B 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 106 SIMULATOR INSTRUCT AREA TRG/3PT-DO01-D006 ONLY) #48 164 CONTROL ROOM & MASTER OPS(3PT-D001/6(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 2 489 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

Page 1 of2 Distribution of IP3 Technical Specification Amendment 224 and 225 (Approved by NRC March 22, 2005 for Amend 224 and March 24 for Amend 225)

Pages are to be inserted into your controlled copy of the IP3 Technical Specifications following the instructions listed below. The TAB notation indicates which section the pages are located.

REMOVE PAGES INSERT PAGES TAB - Facility Operating License Remove all 7 pages Insert 8 new pages TAB - List of Effective Pages Pages 1 through 3, Pages 1 through 3, (Amendment 223) (Amendment 225)

TAB - List of Amendments Page 13 Page 13 TAB 1.0 - Use and Application Page 1.1-3 (Amendment 205) Page 1.1-3 (Amendment 224)

Page 1.1-5 (Amendment 213) Page 1.1-5 (Amendment 225)

TAB 2.0 - Safety Limits Page 2.0-1 (Amendment 205) Page 2.0-1 (Amendment 225)

Page 2.0-1 (Amendment 213) N/A ; page is deleted TAB 3.3 - Instrumentation Page 3.3.1-13 (Amendment 205) Page 3.3.1-13 (Amendment 225)

Page 3.3.1-15 (Amendment 205) Page 3.3.1-15 (Amendment 225)

Page 3.3.1-19 (Amendment 205) Page 3.3.1-19 (Amendment 225)

Page 3.3.1-20 (Amendment 205) Page 3.3.1-20 (Amendment 225)

Page 3.3.2-8 (Amendment 213) Page 3.3.2-8 (Amendment 225)

Page 3.3.2-11 (Amendment 213) Page 3.3.2-11 (Amendment 225)

Page 3.3.7-1 (Amendment 205) Page 3.3.7-1 (Amendment 224)

Page 2 of 2 TAB 3.4 - Reactor Coolant System Page 3.4.1-1 (Amendment 205) Page 3.4.1-1 (Amendment 225)

Page 3.4.1-2 (Amendment 205) Page 3.4.1-2 (Amendment 225)

Page 3.4.9-1 (Amendment 216) Page 3.4.9-1 (Amendment 225)

Page 3.4.9-2 (Amendment 216) Page 3.4.9-2 (Amendment 225)

TAB 3.7 - Plant Systems Page 3.7.1-3 (Amendment 213) Page 3.7.1-3 (Amendment 225)

Page 3.7.11-2 (Amendment 205) Page 3.7.11-2 (Amendment 224)

TAB 5.0 - Administrative Controls Page 5.0-21 (Amendment 205) Page 5.0-21 (Amendment 224)

Page 5.0-22 (Amendment 205) Page 5.0-22 (Amendment 224)

Page 5.0-23 (Amendment 205) Page 5.0-23 (Amendment 224)

Page 5.0-24 (Amendment 219) Page 5.0-24 (Amendment 224)

Page 5.0-25 (Amendment 205) Page 5.0-25 (Amendment 224)

Page 5.0-31 (Amendment 206) Page 5.0-31 (Amendment 225)

Page 5.0-34 (Amendment 205) Page 5.0-34 (Amendment 225)

Page 5.0-35 (Amendment 217) Page 5.0-35 (Amendment 225)

ENTERGY NUCLEAR INDIAN POINT 3. LLC AND ENTERGY NUCLEAR OPERATIONS. INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDED FACILITY OPERATING LICENSE Amendment No. 203 License No. DPR-64

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the Power Authority of the Amdt. 203 State of New York (PASNY) and Entergy Nuclear Indian Point 3, 11/27/00 LLC (ENIP3) and Entergy Nuclear Operations, Inc. (ENO),

submitted under cover letters dated May 11 and May 12, 2000, as supplemented on June 13, June 16, July 14, September 21, October 26, and November 3, 2000, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. ENIP3 and ENO are financially and technically qualified to Amdt. 203 engage in the activities authorized by this amendment; 11/27/00 E. ENIP3 and ENO have satisfied the applicable provisions of Amdt. 203 10 CFR Part 140, "Financial Protection Requirements and 11127/00 Indemnity Agreements" of the Commission's regulations; F. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; Amendment No. 225

G. The receipt, possession and use of source, byproduct and special nuclear material as authorized by this amendment will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70 including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31; and H. The issuance of this amendment is in accordance with 10 CFR Part 51 of the.

Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, Facility Operating License No. DPR-64 (previously Amdt. 203 issued to Consolidated Edison Company of New York, Inc., and 11/27/00 the Power Authority of the State of New York) is hereby amended in its entirety and transferred to ENIP3 and ENO on November 21, 2000, to read as follows:

A. This amended license applies to the Indian Point Nuclear Amdt. 203 Generating Unit No. 3, a pressurized water nuclear reactor 11/27/00 and associated equipment (the facility), owned by ENIP3 and operated by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the "Final Facility Description and Safety Analysis Report" as supplemented and amended, and the Environmental Report, as amended.

B. Subject to the conditions and requirements incorporated herein, the Commission licenses:

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, Amdt. 203 "Licensing of Production and Utilization Facilities," 11/27/00 (a) ENIP3 to possess and use, and (b) ENO to possess, use and operate, the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this amended license; (2) ENO pursuant to the Act and 10 CFR Part 70, to receive, Amdt. 203 possess, and use, at any time, special nuclear material as 11/27/00 reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Facility Description and Safety Analysis Report, as supplemented and amended; (3) ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, Amdt. 203 to receive, possess, and use, at any time, any byproduct 11/27/00 source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; Amendment No. 225

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Amdt. 203 to receive, possess, and use in amounts as required any 11/27/00 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Amdt. 203 possess, but not separate, such byproduct and special 11/27100 nuclear materials as may be produced by the operation of the facility.

C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 225 are hereby incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications.

(3) (DELETED - Amdt. 205 2-27-01 (4) (DELETED) Amdt. 205 27-01 D. (DELETED) Amdt.46 2-16-83 E. (DELETED) Amdt.37 5-14-81 F. This amended license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter of May 2, 1975, to Consolidated Edison Company of New York, Inc., granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.

Amendment No. 225

G. ENO shall fully implement and maintain in effect all provisions Letter of of the Commission-approved physical security, training and 10-28-04 qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and to the authority of 10 CFR 50.90 and CFR 50.54(p). The combined set of plans' for the Indian Point Energy Center, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

"Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0," and was submitted by letter dated October 14, 2004.

H. ENO shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for Indian Point Nuclear Generating Unit No. 3 and as approved in NRC fire protection safety evaluations (SEs) dated September 21, 1973, March 6. 1979, May 2, 1980, November 18, 1982, December 30, 1982, February 2, 1984, April 16, 1984, January 7, 1987, September 9, 1988, October 21, 1991, April 20, 1994, January 5, 1995, and supplements thereto, subject to the following provision:

ENO may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

1. (DELETED) Amdt. 205 2/27101 J. (DELETED) Amdt. 205 2/27/01 K. (DELETED) Amdt.49 5-25-84 L. (DELETED) Amdt. 205 2127101 M. (DELETED) Amdt. 205 2127/01

14. (DELETED) Amdt. 49 5-25-84 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Amendment No. 225

0. Evaluation, status and schedule for completion Amdt. 47 of balance of plant modifications as outlined in letter 5-27-83 dated February 12,1983, shall be forwarded to the NRC by January 1, 1984.

P. Entergy Nuclear IP3 and ENO shall take no Amdt. 203 action to cause Entergy Global Investments, Inc. 11/21100 or Entergy International Ltd. LLC, or their parent companies to void, cancel, or modify the $70 million contingency commitment to provide funding for the facility as represented in the application for approval of the transfer of the license from PASNY to ENIP3 and ENO, without the prior written consent of the Director, Office of Nuclear Reactor Regulation.

Q. The decommissioning trust agreement shall Amdt. 203 provide that the use of assets in the 11/27/00 decommissioning trust fund, in the first instance, shall be limited to the expenses related to decommissioning of the facility as defined by the NRC in its regulations and issuances, and as provided in this license and any amendments thereto.

R. The decommissioning trust agreement shall Amdt. 203 provide that no contribution to the 11/27/00 decommissioning trust fund that consists of property other than liquid assets shall be permitted.

S. With respect to the decommissioning trust fund, Amdt. 203 investments in the securities or other obligations 11)27100 of PASNY, Entergy Corporation, ENIP3, Entergy Nuclear FitzPatrick, LLC, ENO, or affiliates thereof, or their successors or assigns, shall be prohibited. Except for investments that replicate the composition of market indices or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear plants is prohibited.

T. The decommissioning trust agreement shall Amdt. 203 provide that no disbursements or payments from 11127/00 the trust, other than for ordinary administrative expenses, shall be made by the trustee until the Amendment No. 225

trustee has first given the NRC 30 days prior written notice of the payment. In addition, the trust agreement shall state that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the Director, Office of Nuclear Reactor Regulation.

(21) The decommissioning trust agreement shall Amdt. 203 provide that the trust agreement shall not be 11/27/00 modified in any material respect without the prior written consent of the Director, Office of Nuclear Reactor Regulation.

V. The decommissioning trust agreement shall state Amdt. 203 that the trustee, investment advisor, or anyone 11127/00 else directing the investments made in the trust shall adhere to a "prudent investment" standard, as specified in 18 CFR 35.32(a)(3) of the Federal Energy Regulatory Commission's regulations.

W. For purposes of ensuring public health and Amdt. 203 safety, ENIP3, upon the transfer of this license to 11/27/00 it, shall provide decommissioning funding assurance for the facility by the prepayment or equivalent method, to be held in a decommissioning trust fund for the facility, of no less than the amount required under NRC regulations at 10 CFR 50.75.

Any amount held in any decommissioning trust maintained by PASNY for the facility after the transfer of the facility license to ENIP3 may be credited towards the amount required under this paragraph.

X. ENIP3 shall take all necessary steps to ensure Amdt. 203 that the decommissioning trust is maintained 11(27100 in accordance with the application for the transfer of this license to ENIP3 and ENO and the requirements of the order approving the transfer, and consistent with the safety evaluation supporting such order.

AA. The following conditions relate to the amendment approving Amdt. 205 the conversion to Improved Standard Technical Specifications: 2/27/01

1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documents as described in Table R, "Relocated Technical Specifications Amendment No. 225

from the CTS," and Table LA, 'Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staff's Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the.

implementation of this amendment.

2. The following is a schedule for implementing surveillance requirements (SRs):

For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of this amendment.

For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after the date of implementation of this amendment.

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the date of implementation of this amendment.

For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the date of implementation of this amendment.

AB. With the reactor critical, Entergy shall maintain the reactor coolant system cold leg at a temperature (TId) greater than or equal to 525 F. Entergy shall maintain a record of the cumulative time that the plant is operated with the reactor critical while Tcold is below 525 F. Upon determination by Entergy that the cumulative time of plant operation with the reactor critical while Tho is below 525 F has exceeded one (1) year, Entergy must:

(a) within one (1) month, inform the NRC, in writing, and (b) within six (6) months submit the results of an analysis of the impact of the operation with Tcm below 525 F on the pressurized thermal shock reference temperature (RTpTs).

Amendment No. 225

3. This amended license is effective at 12:01 a.m., November 21, 2000, and shall expire at midnight December 12, 2015.

Original signed by Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: March 8, 1978 Amendment No. 225

INDIAN POINT 3 TECHNICAL SPECIFICATIONS - APPENDIX A LIST OF EFFECTIVE PAGES Page 1 of 3 Page Amn d Section 3.0 2 205 8 225 Table of Contents 1 205 3 J 205 9 213 205 2 205 Section 3.2.3 10 205 ii 205 3 205 1 205 11 225 iii 207 4 212 2 205 12 205 iv 210 5 205 3 205 13 205 V 207 Section 3.1.1 4 205 Section 3.3.3 Section 1.1 I 1 205 Section 3.2.4 1 211 1 205 Section 3.1.2 1 205 2 211 205 1 205 2 205 3 205 3 224 2 205 3 205 4 ~ 211 4 ~ 205 Section 3.1.3 4 205 5 211 5 225 1 205 Section 3.3.1 Section 3.3.4 205 2 205 1 205 1 l 205 7 205 Section 3.1.4 2 205 2f 205 8 205 1 205 3 205 Section 3.3.5 Section 1.2 2 205 4 205 1 205 1 205 3 205 5 205 2 1 205 2 f 205 4 205 6 205 Section 3.3.6 3 205 5 205 7 205 1 205 Section 1.3 Section 3.1.5 8 205 2 205 1 205 1 205 9 205 3 205 2 205 2 205 10 205 4 205 3 205 Section 3.1.6 11 205 5 205 Seto ..

4F 205 1 f205 12 205 Section 3.3.7 5 205 2 205 13 225 1 1 205 6 205 3 205 14 205 2 205 7 205 Section 3.1.7 15 225 3 205 8- 205 1 205 16 205 Section 3.3.8 9 205 2 205 17 205 1 215 10 205 3 205 18 205 2 205 11 205 4 205 19 225 Section 3.4.1 12 205 Section 3.1.8 20 225 1 225 13 205 1 205 Section 3.3.2 2 225 Section 1.4 2 205 1 205 Section 3.4.2 1 205 Section 3.2.1 2 205 1 1 205 2 205 11 205 3 205 Section 3.4.3 3 205 2 205 4 205 1 1 205 4 205 3 205 5 205 2 1 205 Section 2.0 Section 3.2.2 6 205 3 1 220 4 220 1 l 22:5 1 i 205 7 205 14 1 220 The latest amendment reflected in this list is: Amendment 225

INDIAN POINT 3 TECHNICAL SPECIFICATIONS - APPENDIX A LIST OF EFFECTIVE PAGES Page 2 of 3 5 1 220 Section 3.4.13 4 205 Section 3.7.3 Section 3.4.4 S1 205 5 205 1 207 1 1 _205 2 205 Section 3.6.3 2 F 207 Section 3.4.5 Section 3.4.14 1, 205 3 [ 207 1 205 1 1 205 2 205 Section 3.7.4 2 205 2 - 205 3 205 1 205 3 205 3 205 4 205 2 [ 205 Section 3.4.6 4 205 5 205 Section 3.7.5 1 205 5 205 6 205 1 205 2 205 Section 3.4.15 Section 3.6.4 2 205 3 205 1 205 1 1 205 3 205 Section 3.4.7 2 205 Section 3.6.5 4 205 1 205 3 205 1i r 205 Section 3.7.6 2 205 4 205 Section 3.6.6 1 .205 3 205 Section 3.4.16 1 _205 2 205 Section 3.4.8 1 205 2 205 Section 3.7.7 1 205 2 205 3 205 1 1 205 2 J205 3 205 4 205 2 218 Section 3.4.9 4 205 Section 3.6.7 Section 3.7.8 Si3225 Section 3.5.1 1 205 1 1 205 2 225 1 222 2 205 2 1 205 Section 3.4 .10 2 205 Section 3.6.8 Section 3.7.9 I 1 205 3 205 1 F 205 1 l 205 2 1 205 Section 3.5.2 2 [ 205 2 205 Section 3.4.11 1 205 Section 3.6.9 3 205 1 205 2 205 1 1 205 Section 3.7.10 2 205 3 205 2 1 205 1 1 205 3 205 4 205 Section 3.6.10 Section 3.7.11 Section 3.4.12 Section 3.5.3 1 205 1 223 1 205 1 205 2 205 2 205 2 205 2 ]205 3 205 Section 3.7.12 3 205 Section 3.5.4 4 205 1 [ 205 4 205 1 205 Section 3.7.1 Section 3.7.13 5 205 2 205 1 205 I 1 215 6 205 Section 3.6.1 2 205 2 1 205 7 205 1 205 3 225 Section 3.7.14 8 205 2 205 4 205 1 l 205 9 220 Section 3.6.2 Section 3.7.2 Section 3.7.15 10 220 1 205 1 205 1 205 11 220 2 205 2 205 2 205 12 220 3 205 3 205 Section 3.7.16 -

The latest amendment reflected in this list is: Amendment 225

INDIAN POINT 3 TECHNICAL SPECIFICATIONS - APPENDIX A LIST OF EFFECTIVE PAGES Page 3 of 3 1 205 2 205 9 210 2 205 3 205 10 205 3 205 Section 3.8.8 11 221 Section 3.7.17 1 205 12 205 1 . I205 2 205 13 205 Section 3.8.1 Section 3.8.9 14 205 1 205 1 1205 15 205 2 205 2 205 16 205 3 205 Section 3.8.10 17 205 4 205 1 205 18 205 5 205 2 205 19 205 6 205 Section 3.9.1 20 205 7 205 1 ] 205 21 224 8 205 Section 3;9.2 22 224 9 205 1 205 23 224 10 205 2 205 24 224 Section 3.8.2 Section 3.9.3 25 224 1 205 1 . 215 26 205 2 205 2 215 27 205 3 205 3 215 28 205 4 205 Section 3.9.4 29 205 Section 3.8.3 1 205 30 206 1 205 2 205 31 225 2 205 Section 3.9.5 32 205 3 205 1 205 .33 - 205 205 2 205 34 225 5 205 Section 3.9.6 35 225 Section 3.8.4 1 [ 205 36 205 1 216 Section 4.0 37 205 2 216 1 205 38 205 3 . 216 2 205 Section 3.8.5 3 205 1 1 205 Section 5.0 2 J 205 1 205 Section 3.8.6 2 205 1 205 3 205 2 T 205 4 1205 3 X_205 5_____ 205 4 205 6 205 Section 3.8.7 7 1205 I I 205 8 1205 The latest amendment reflected in this list is: Amendment 225

Entergy Nuclear Operations, Inc Indian Point 3 Nuclear Power Plant License Amendments Page 13 AMENDMENT SUBJECT LETTER DATE 217 Use of Best-Estimate Large-Break Loss of 05/06/2003 Coolant Accident analysis methodology (WCAP 12945) 218 Revise City Water surveillance to reflect 08/04/2003 addition of (backflow preventer) valves 219 Revise Ventilation Filter Testing Program 10/30/2003 to adopt ASTM D3803 charcoal filter testing requirements per GL 99-02.

220 Extension of the RCS pressure/temperature 12/03/2003 limits and corresponding OPS limits from 16.17 to 20 EFPY.

221 Extension of RCP flywheel inspection interval 07/02/2004 (from 10 years to 20 years) per TSTF 421.

222 Inoperable accumulator time extended from 08 /18/2004 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per TSFT-370.

223 Extension of the allowed outage time to 01/19/2005 support the placement of the CRVS in an alternate configuration for tracer gas testing.

224 Full-scope adoption of alternate source term 03/22/2005 for dose consequence analysis of DBAs.

225 Stretch Power Uprate (4.85%) from 3067.4 MWt 03/24/2005 to 3216 MWt, and adoption of TSTF-339.

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT I-131 DOSE EQUIVALENT 1-131 shall be that amount of 1-131 (curies) that alone would produce the same committed effective dose equivalent (CEDE) dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and 1-135 actually present. The CEDE dose conversion factors used for this calculation shall be those listed in Table 2.1 of EPA Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

E - AVERAGE E shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 10 minutes, making up at least 95% of the total noniodine activity in the coolant.

La The maximum allowable primary containment leakage rate, La, shall be 0.1% of primary containment air weight per day at the calculated peak containment pressure (Pa).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except for leakage into closed systems and reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (Leakage into closed systems is leakage that can be accounted for and contained by a (continued)

INDIAN POINT 3 1.1 - 3 Amendment 224

Definitions 1.1 1.1 Definitions MODE (continued) vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE -OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it-is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in FSAR Chapter 13, Initial Tests and Operations;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper RATIO (QPTR) excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 3216 MWt.

I (continued)

INDIAN POINT 3 1.1 - 5 Amendment 225

SLs 2.0 2.1 SLs 2.1.1 Reactor Core SLs.

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Vessel inlet temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.17 for the WRB-l DNB correlations.

2.1.1.2 The peak fuel centerline temperature shall be 0

maintained < 5080F, decreasing by 58 F per 10,000 MWD/MTU of burnup.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, 5, and in MODE 6 when the reactor vessel head is on, the RCS pressure shall be maintained < 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, 5, or 6, restore compliance within 5 minutes.

INDIAN POINT 3 2.0- 1 Amendment 225

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 8)

Reactor Protection System Instrumentation APPLICABLE MODES OR FUNCTION OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE -

1. Manual Reactor 1,2 2 B SR 3.3.1.14 NA Trip 3 (a) 4 (a) 5(a) 2 C SR 3.3.1.14 NA

2. Power Range Neutron Flux

a. High 1,2 40) D SR 3.3.1.1 *111% RTP SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.1 1

b. Low 40) E SR 3.3.1.1 <25% RTP SR 3.3.1.8 SR 3.3.1.11

3. Intermediate 1 (b) 2 (c) 1 F SR 3.3.1.1 NA Range Neutron SR 3.3.1.8 Flux SR 3.3.1.11 (continued)

(a) With Rod Control System capable of rod withdrawal and one or more rods not fully inserted.

(b) Below the P-10 (Power Range Neutron Flux) interlocks.

(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(j) Only 3 channels required during Mode 2 Physics Tests, LCO 3.1.8 INDIAN POINT 3 3.3.1-13 Amendment 225

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 8)

Reactor Protection System Instrumentation APPLICABLE FUNCTION MODES OR SURVEILLANCE OTHER SPECIFIED REQUIRED CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

7. Pressurizer Pressure

a. Low 1(e) 4 H SR 3.3.1.1 21900 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.1

b. High 1,2 3 E SR 3.3.1.7 <2400 psig SR 3.3.1.10

8. Pressurizer 1(e) 3 H SR 3.3.1.1 *97%

Water Level - SR 3.3.1.7 High SR 3.3.1.10

9. Reactor Coolant 1(e) 3 per loop H SR 3.3.1.1 Ž90%

Flow- Low SR 3.3.1.7 SR 3.3.1.10 (continued)

(e) Above the P-7 (Low Power Reactor Trips Block) interlock.

INDIAN POINT 3 3.3.1-1 5 Amendment 225

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 8)

Reactor Protection System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 2.8% of AT span:

AT<ATo K,-K2 ) [T-T']+K,(P- P)- f,( I)}

',+

Where: AT is measured RCS AT, OF.

ATo is the indicated AT at RTP, OF.

s is the Laplace transform operator, sec' 1.

T is the measured RCS average temperature, OF.

T' is the nominal Tavg at RTP, < [ * ]OF.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, _ [ * ] psig K, <[*] K 2> *]/OF K 3 Ž[*I/psig

'r, 2 [*]sec T2 * [*]sec fl(Al) = [ *f ] + (qt - qb)) when qt - qb J% RTP 0%of RTP when -[ * ]% RTP < qt - qb < [% RTP

-[* (qt - qb) - [ }when qt - qb > [ * ]% RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

The values denoted with (* I are specified in the COLR.

INDIAN POINT 3 3.3.1-19 Amendment 225

RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 8)

Reactor Protection System Instrumentation' Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 1.8% of AT span:

A T <A To {Kt - Ki (7 3 ) T -K 6 (T - T")-f 2 (A I)}

Where: AT is measured RCS AT, "F.

ATo is the indicated AT at RTP, OF.

s is the Laplace transform operator, sec-'.

T is the measured RCS average temperature, 0F.

T" is the nominal Tavg at RTP, * [* ]OF.

K4* [*] K(5Ž P*JcFforincreasingTzvg K.6 t*1/OFwhenT>T

[*1/OF for decreasing Tavg [/*'1F when T

• T r3 * [*]sec f2 (AI) =1*]

• The values denoted with [*] are specified in the COLR.

INDIAN POINT 3 ,3.3.1-20 Amendment 225

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 6)

Engineered Safety Feature Actuation System Instrumentation APIPLICABLE MODES OR FUNCTION OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

1. Safety Injection

a. Manual Initiation 1,2,3,4 2 B SR 3.3.2.6 NA

b. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays

c. Containment 1,2,3 3 D SR 3.3.2.1 <4.80 psig Pressure-Hi SR 3.3.2.4 SR 3.3.2.7

d. Pressurizer 1,2,3(b) 3 D SR 3.3.2.1 Ž1710 psig I Pressure-Low SR 3.3.2.4 SR 3.3.2.7

e. High Differential 1,2,3 3 per D SR 3.3.2.1 NA Pressure steam line SR 3.3.2.4 Between Steam SR 3.3.2.7 Lines

f. High Steam 1 ,2 (d),3(d) 2 per D SR 3.3.2.1 (c)

Flow in Two steam line SR 3.3.2.4 Steam Lines SR 3.3.2.7 Coincident with 1 ,2 (d),

3 (d) 1 per loop D SR 3.3.2.1 >540.50F Tavg- Low SR 3.3.2.4 SR 3.3.2.7 (continued)

(a) Not used (b) Above the Pressurizer Pressure interlock.

(c) Less than or equal to turbine first stage pressure corresponding to 54% full steam flow below 20% load, and increasing linearly from 54% full steam flow at 20% load to 120% full steam flow at 100% load, and corresponding to 120% full steam flow above 100% load. Time delay for SI <6 seconds.

(d) Except when all MSIVs are closed.

INDIAN POINT 3 3.3.2-8 Amendment 225

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR FUNCTION OTHER SPECIFIED REQUIRED SURVEILLANCE CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

4. Steam Line Isolation

a. Manual Initiation 1 ,2 (d)3(d) 2 per F SR 3.3.2.6 NA steam line

b. Automatic 1,2 (d), 3(d) 2 trains G SR 3.3.2.2 NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays

c. Containment 1,2(d) 2 sets of 3 E SR 3.3.2.1 *24 Pressure (Hi-Hi) 3(d) SR 3.3.2.4 psig SR 3.3.2.7

d. High Steam Flow 1,2(d) 2 per D SR 3.3.2.1 (c) in Two Steam 3(d) steam SR 3.3.2.4 Lines line SR 3.3.2.7 Coincident with 1,2(d), 1 per loop D SR 3.3.2.1 >540.50 F Tavg-LOw 3(d) SR 3.3.2.4 SR 3.3.2.7

e. High Steam Flow 1,2(d), 2 per steam D SR 3.3.2.1 (c) in Two Steam 3(d) line SR 3.3.2.4 Lines SR 3.3.2.7 Coincident with 1 ,2(d), 1 per steam D SR 3.3.2.1 >500 psig Steam Line 3(d) line SR 3.3.2.4 Pressure-Low SR 3.3.2.7 (c) Less than or equal to turbine first stage pressure corresponding to 54% full steam flow below 20% load, and increasing linearly from 54% full steam flow at 20% load to 120% full steam flow at 100% load, and corresponding to 120% full steam flow above 100% load. Time delay for SI *6 seconds.

(d) Except when all MSIVs are closed.

- INDIAN POINT 3 3.3.2-1 1 Amendment 225

CRVS Actuation Instrumentation 3.3.7 3.3 INSTRUMENTATION 3.3.7 Control Room Ventilation System (CRVS) Actuation Instrumentation LCO 3.3.7 The CRVS actuation instrumentation for each Function in Table 3.3.7-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4 ACTIONS S t i n s e e--NOTE-.

Separate Condition entry is allowed for each Function.

COMPLETION REQUIRED ACTION TIME CONDITION A. One or more Functions A.1 Place CRVS in CRVS 7 days with one channel or train Mode 3.

inoperable.

B. One or more Functions B.1.1 Place CRVS in CRVS 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with two channels or two Mode 3.

trains inoperable.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time for Condition A or B AND not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> INDIAN POINT 3 3.3.7-1 Amendment 224

RCS Pressure, Temperature, and Flow Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is greater than or equal to the limit specified in the COLR;

b. RCS average loop temperature is less than or equal to the limit specified in the COLR; and

c. RCS total flow rate 2 354,400 gpm and greater than or equal to the limit specified in the COLR.

APPLICABILITY: MODE 1.

---

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

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS A.I Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> DNB parameters not parameters to within limits. within limits.

B. Required action and B.l Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

INDIAN POINT 3 3.-4 .1-1 Amendment 225

RCS Pressure, Temperature, and Flow Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> than or equal to the limit specified in the COLR.

SR 3.4.1.2 Verify RCS average loop temperature is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> less than or equal to the limit specified in the COLR.

SR 3.4.1.3 Verify RCS total flow rate is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2 354,400 gpm and greater than or equal to the limit specified in the COLR.

SR 3.4.1.4 ------------------- NOTE----------------

Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 2 90% RTP.

Verify by precision heat balance that 24 months RCS total flow rate is 2 354,400 gpm and greater than or equal to the limit specified in the COLR.

INDIAN POINT 3 3. 4. 1-2 Amendment 225

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level < 54.3% in MODES 1 and 2 or < 90% in MODE 3; and

b. Two groups of pressurizer heaters OPERABLE with the capacity of each-group 2 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water A.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> level not within reactor trip limit. breakers open.

AND A.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. One required group B.1 Restore required 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of pressurizer group of heaters inoperable. pressurizer heaters to OPERABLE status.

C. Required Action and C.l Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of AND Condition B not met. C.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> INDIAN POINT 3 3 .4 .9- 1 Amendment 225

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

< 54.3% in MODES 1 and 2 OR < 90% in MODE 3.

SR 3.4.9.2 Verify capacity of each required group 24 months of pressurizer heaters is 2 150 kW.

INDIAN POINT 3 3 .4 .9- 2 Amendment 225

MSSVs 3.7.1 Table 3.7.1-1 (page 1 of 1)

OPERABLE Main Steam Safety Valves versus Applicable Neutron Flux Trip Setpoint in Percent of RATED THERMAL POWER MINIMUM NUMBER OF MSSVS APPLICABLE Neutron Flux Trip PER STEAM GENERATOR Setpoint REQUIRED OPERABLE (% RTP) 4 5- 57 3 5- 38 2

• 20 INDIAN POINT 3 3. 7. 1-3 Amendment 225

CRVS 3.7.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Operate each CRVS train for 2 15 minutes. 31 days SR 3.7.11.2 Perform required CRVS filter testing in In accordance with the Ventilation Filter accordance Testing Program (VFTP). with VFTP SR 3.7.11.3 Verify each CRVS train actuates on an 24 months actual or simulated actuation signal.

SR 3.7.11.4 Verify one CRVS train can maintain a 24 months on slight positive pressure relative to a STAGGERED the adjacent enclosed area during CRVS TEST BASIS Mode 3 operation at a makeup flow rate of > 1500 cfm.

INDIAN POINT 3 3. 7.11-2 Amendment 224

Programs and Manuals 5.5 5.5 Programs and Manuals

5. 5.10 Ventilation Filter Testing Program (VFTP)

This program provides controls for implementation of required testing of the ventilation filter function for the Control Room Ventilation System and Containment Fan Cooler Units.

Applicable tests described in Specifications 5.5.10.a, 5.5.10.b, 5.5.10.c and 5.5.10.d shall be performed:

1) After 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber use since the last test; and,

2) Every 24 months for the Control Room Ventilation System, and Containment Fan Cooler Units; and,

3) After each complete or partial replacement of the HEPA filter train or charcoal adsorber filter; and,

4) After any structural maintenance on the system housing that could alter system integrity; and,

5) After significant painting, fire, or chemical release in any ventilation zone communicating with the system while it is in operation.

SR 3.0.2 is applicable to the Ventilation Filter Testing Program.

(continued)

INDIAN POINT 3 5.0 - 21 Amendment 224

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

a. Demonstrate for each system that an inplace test of the high efficiency particulate air (HEPA) filters shows the specified penetration and system bypass leakage when tested in accordance with the referenced standard at the flowrate specified below.

Ventilation Removal Flowrate Sys Efficiency (cfm) Reference Standard 80% to 120%

Control Room of design Ventilation > 99% accident Regulatory Guide 1.52, Rev 2, System rate Sections C.5.a and C.5.c 80% to 120%

of design Containment Fan > 99% accident Regulatory Guide 1.52, Rev 2, Cooler Units rate Sections C.5.a and C.5.c (continued)

INDIAN POINT 3 5. 0 - 22 Amendment 224

Programs and Manuals 5.5 5.5 Programs and Manuals.

5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

b. Demonstrate for each system that an inplace test of the charcoal adsorber shows the specified penetration and system bypass leakage when tested in accordance with the referenced standard at the flowrate specified below.

Ventilation Removal Syst Efficiency Flowrate (cfm) Reference Standard 80% to 120% of Control Room design Ventilation accident rate Regulatory Guide 1.52, Rev 2, System > 99% Sections C.5.& and C.5.d 80% to 120% of Containment Fan design Regulatory Guide 1.52, Rev 2, Cooler Units > 99% accident rate Sections C.5.a and C.5.d (continued)

INDIAN POINT 3 5. 0 - 23 Amendment 224

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

c. Demonstrate for each system that a laboratory test of a sample of the charcoal adsorber shows the methyl iodide removal efficiency specified below when tested in accordance with ASTM D3803-1989, subject to clarification 0

below, at a temperature of 86 F and a relative humidity of 95%.

Methyl iodide ASTM D3803-1989 Ventilation System removal Clarification efficiency (0:

Control Room > 95.5 78 ft/min face Ventilation System velocity Containment Fan > 85 59 ft/min face Cooler Units velocity Note: For the 1" beds, the Control Room Ventilation System methyl iodide removal efficiency is verified greater than or equal to 93% rather than 95.5% at a face velocity of 50 ft/min under the above requirements.

This is done prior to fuel movement in Refuel Outage 12 and every 6 months after Refuel Outage 12 until the end of Refuel Outage 13 or the 2" beds are installed.

(continued)

INDIAN POINT 3 5. 0 - 24 Amendment 224

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP) (continued)

d. Demonstrate for each system that the pressure drop across the combined HEPA filters, the demisters and prefilters (if installed), and the charcoal adsorbers is less than the value specified below when tested at the flowrate specified below.

Ventilation System elta P Flowrate (cfm):

(inches wg)

I Control Room Ventilation System > 90% of design accident 6 rate Containment Fan Cooler Units > 90% of design accident 6 rate (continued)

INDIAN POINT 3 5. 0 - 2 5 Amendment 224

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Containment Leakage Rate Testing Program (continued) cooler unit when pressurized at > 1.1 Pa. This limit protects the internal recirculation pumps from flooding during the 12-month period of post accident recirculation.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10CFR50, Appendix J.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 42.0 psig. The containment design pressure is 47 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of primary containment air weight per day.

INDIAN POINT 3 5. 0 - 31 Amendment 225

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. Specification 2.1, Safety Limits (SL);

2. Specification 3.1.1, Shutdown Margin;

3. Specification 3.1.3, Moderator Temperature Coefficient;

4. Specification 3.1.5, Shutdown Bank Insertion Limits;

5. Specification 3.1.6, Control Bank Insertion Limits;

6. Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

7. Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor;

8. Specification 3.2.3, AXIAL FLUX DIFFERENCE (AFD);

9. Specification 3.3.1, Reactor Protection System Instrumentation;

10. Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and

11. Specification 3.9.1, Boron Concentration.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Specifications 3.1.5, Shutdown Bank Insertion Limits, 3.1.6, Control Bank Insertion Limits, and 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor);

2a. WCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES, TOPICAL REPORT," September 1974 (W Proprietary). (Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control);

2b. T. M. Anderson to K. Kneil (Chief of Core Performance Branch, NRC) January 31, 1980 --

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package. (Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control));

2c. NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission,Section 4.3,Nuclear Design,July 1981. Branch INDIAN POINT 3 5 .0 - 34 Amendment 225

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981. (Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control));

3a. WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), "Code Qualification Document for Best-Estimate Loss-of-Coolant-Accident Analysis,"

March 1998 (Westinghouse Proprietary);

3b. WCAP-11397-P-A, "Revised Thermal Design Procedure,"

April 1989 (Specification 2.1, Safety Limits (SL)) and Specification 3.4.1, (RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits);

3c. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and Thermal Overtemperature AT Trip Functions,"

September 1986 (Specification 2.1, Safety Limits (SL));

3d. WCAP-10054-P-A, "SMALL BREAK ECCS EVALUATION MODEL USING NOTRUMP CODE," (W Proprietary). (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

3e. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code; Safety Injection into the Broken Loop and Cosi Condensation Model," July 1997 (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)));

3f. WCAP-10079-P-A, "NOTRUMP NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE," (W Proprietary). (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))); and 3g. WCAP-12610, "VANTAGE+ Fuel Assembly Report,"

(W Proprietary). (Specification 3.2.1, Heat-Flux Hot Channel Factor).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided for each reload cycle to the NRC.

5.6.6 NOT USED (continued)

INDIAN POINT 3 5. 35 Amendment 225