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Official Exhibit - NRCR00118-00-BD01 - Indian Point Unit 3 Technical Specifications (Excerpt) (Revised)
	ML15335A291
Person / Time
Site:	Indian Point
Issue date:	08/10/2015
From:	; NRC/OGC
To:	; Atomic Safety and Licensing Board Panel
	SECY RAS
References
	RAS 28146, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
	Download: ML15335A291 (10)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 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 State of New York (PASNY) and Entergy Nuclear Indian Point 3, 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; Arndt. 203 11/27/00 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. E. F. ENIP3 and ENO are financially and technically qualified to engage in the activities authorized by this amendment; ENIP3 and ENO have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commission's regulations; Arndt. 203 11/27/00 Arndt. 203 11/27/00 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

2. 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 1 O CFR Sections 30.33, 40.32, 70.23, and 70.31; and H. The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Accordingly, Facility Operating License No. DPR-64 (previously issued to Consolidated Edison Company of New York, Inc., and 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 Generating Unit No. 3, a pressurized water nuclear reactor 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.

Arndt. 203 11/27/00 Arndt. 203 11/27/00 B. Subject to the conditions and requirements incorporated herein, the Commission licenses:

(1) (2) (3) Pursuant to Section 104b of the Act and 1 O CFR Part 50, "Licensing of Production and Utilization Facilities,*

(a) ENIP3 to possess and use, and (b) ENO to possess, use and operate, the f acitity at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this amended license; ENO pursuant to the Act and 10 CFR Part 70, to receive,

possess, and use, at any time, special nuclear material as 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; ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive,

possess, and use, at any time, any byproduct 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; Arndt. 203 11/27/00 Arndt. 203 11/27/00 Arndt. 203 11/27/00 Amendment No. 225 (4) (5) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive,

possess, and use in amounts as required any 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

ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. Arndt. 203 11/27/00 Arndt. 203 11/27/00 C. This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 1 O CFR D. E. Chapter I: 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 (3) (4) The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 257 are hereby incorporated in the License. ENO shall operate the facility in accordance with the Technical Specifications

. (DELETED)

Arndt. 205 2-27-01 (DELETED)

Amdt.46 2-16-83 (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. 257 APPENDIX A FACILITY OPERfJING LICENSE DPR-64 TECfiNICAL SPECIFICATIONS AND BA$ES FOR THE INDIAN POINT.3 NUCLEAR GENERATING STJ\TION UNIT NO. 3 WESTCHESTER COQNTY. NEW YORK ENTERGY NUCLEAR INDIAN POINT 3. LLC (ENIP3) AND EN'fERGY NUCLEAR OPERATIONS.

INC. (ENO) I DQCJ{ET NO. 50-286 Date of Issuance:

Apri 1 . 15, 1976 Amendment No. 203 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained

. SG Tube Integrity 3.4.17 All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program. APPLICABILITY:

MODES 1, 2, 3, and 4. ACTIONS -----------------------------------------

N 0 TE----------------------

S e pa rate Condition entry is allowed for each SG tube. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube(s) is plugging criteria and not maintained until the next plugged in accordance refueling outage or SG with the Steam tube inspection

. Generator Program. AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not met. AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SG tube integrity not maintained

. INDIAN POINT 3 3.4.17 -1 Amendment No. 257 SURVEILLANCE REQUIREMENTS SR 3.4.17.1 SR 3.4.17.2 INDIAN POINT 3 SURVEILLANCE Verify SG tube integrity in accordance with the Steam Generator Program. Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program.

3.4.17-2 SG Tube Integrity 3.4.17 FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following a SG tube inspection Amendment No. 257 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

Programs and Manuals 5.5 i. Limitations on the annual and quarterly doses to a member of the public from iodine-131,

tritium, and.all radionuclides in particulate form with half lives> 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Controls Program surveillance frequency.

5.5.5 Component Cyclic or Transient Limit This program provides controls to track the FSAR, Section 4.1.5, cyclic and . transient occurrences to ensure that components are maintained within the design limits. 5.5.6 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel.

The program shall include inspection frequencies and acceptance criteria.

The inspection frequency will ensure that each reactor coolant pump flywheel is surface and volumetrically inspected at 20-year intervals.

(continued)

INDIAN POINT 3 5.0 -11 Amendment 232 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained

. In addition, the Steam Generator Program shall include the following

a. Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met. b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE. INDIAN POINT 3 1. Structural integrity performance criterion

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification and design basis accidents

. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to

-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials

. Apart from the above requirements

, additional loading conditions associated with the design basis accidents

, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. 2. Accident induced leakage performance criterion

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the (continued) 5.0 -13 Amendment 257 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 0.3 gpm per SG and 1 gpm through all SGs. 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE." c. Provisions for SG tube plugging criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged. d. Provisions for SG tube inspections

. Periodic SG tube inspections shall be performed

. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection

. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment

, to determine which inspection methods need to be employed and at what locations

. INDIAN POINT 3 1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

2. After the first refueling outage following SG installation

, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections)

. In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determinat ion that a new (continued) 5.0 -14 Amendment 257 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage. a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; b) During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods. 3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections)

. If definitive information

, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-line indication is not associated with a crack(s), then the indication need not be treated as a crack. e. Provisions for monitoring operational primary to secondary LEAKAGE. INDIAN POINT 3 ---------------------NOTE-------------------

Pages 5.0-16 through 5.0-19 are deleted. Next page is 5.0-20. 5.0 -15 (continued) Amendment 257 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 3. LLC AND ENTERGY NUCLEAR OPERATIONS.